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| {{#Wiki_filter:OFFICIAL USE ONLY SECURITY RELATED INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 29, 2010 Mr. 1. Preston Gillespie Site Vice President Oconee Nuclear Station Duke Energy Carolinas, LLC 7800 Rochester Highway Seneca, SC 29672 | | {{#Wiki_filter:}} |
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| ==SUBJECT:==
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| OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3, ISSUANCE OF AMENDMENTS REGARDING TRANSITION TO A RISK-INFORMED, PERFORMANCE-BASED FIRE PROTECTION PROGRAM IN ACCORDANCE WITH 10 CFR 50.48(c) (TAC NOS. ME3844, ME3845, AND ME3846)
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| ==Dear Mr. Gillespie:==
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| The Nuclear Regulatory Commission has issued the enclosed Amendment Nos. 371, 373, and 372 to Renewed Facility Operating Licenses DPR-38, DPR-47, and DPR-55, for the Oconee Nuclear Station, Units 1, 2, and 3, respectively. The amendments consist of changes to the licenses and Technical Specifications (TSs) in response to your application dated application dated May 30,2008, as supplemented by letters dated October 31,2008, January 30,2009, February 9,2009, February 23,2009, May 31,2009, August 3,2009, September 29, 2009, and November 30, 2009. Duke Energy Carolinas, LLC (Duke, the licensee), submitted a license amendment request (LAR) to allow the licensee to maintain a fire protection program in accordance with 10 CFR 50.48(c) for the Oconee Nuclear Station, Units 1, 2, and 3 (ONS), and change the license and TSs accordingly.
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| By letter dated April 14, 2010, the licensee resubmitted the application and superseded the contents of the application submitted by letter dated May 30, 2008, as supplemented October 31,2008. This resubmitted LAR, however, does not supersede the supplements dated January 30,2009, February 9,2009, February 23,2009, May 31,2009, August 3,2009, September 29,2009, and November 30,2009. By letters dated September 13,2010, September 27,2010, October 14, 2010, November 19, 2010, and December 22,2010, the licensee supplemented the April 14, 2010, application.
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| A new fire protection license condition will replace the existing fire protection license condition in each unit's license. As a result of placing the new license condition in each unit's license, the NRC will be issuing license pages 2 through 11 in each unit's license because of pagination issues. The only changes to the licenses are the changes to the fire protection license condition.
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| Enslosure 4 transmitted herewith sontains sesurily related information. 'Nhen separated from Enslosuro 4, this dooument is deoontrolled.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION T. Gillespie -2 Pursuant to Section 2.390 of Title 10 of the Code of Federal Regulations (10 CFR), by letter dated December 6, 2010, the NRC sent the licensee the draft Safety Evaluation approving the proposed amendments for an opportunity for the licensee to comment on any proprietary or security-related aspects of the draft Safety Evaluation. By letter dated December 22, 2010, the licensee provided comments. The NRC reviewed and accepted all comments made by the licensee. Pursuant to 10 CFR 2.390 the NRC has redacted information as identified by blank space enclosed within double brackets as shown here [[ ]].
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| In addition, the December 6, 2010, letter also requested the licensee to provide comments on factual errors or clarity concerns contained in the draft Safety Evaluation. By letter dated December 22, 2010, the licensee provided comments. The NRC has considered each comment and changed the Safety Evaluation as appropriate.
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| A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
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| If you have any questions, please call me at 301-415-1345.
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| hn Stang, Senior Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-269, 50-270, and 50-287
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| ==Enclosures:==
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| : 1. Amendment No. 371 to DPR-38
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| : 2. Amendment No. 373 to DPR-47
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| : 3. Amendment No. 372 to DPR-55
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| : 4. Safety Evaluation contains official use only security related information cc: Distribution via Listserv Enclosure 4 tra~i~I~~6iUtl~¥se~i?~W¥n~DJ~~e1IjQ~ Enclosure 4, this document is decontrolled.
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| UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-269 OCONEE NUCLEAR STATION! UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 371 Renewed License No. DPR-38
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| : 1. The Nuclear Regulatory Commission (the Commission) has found that:
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| A. The application for amendment to the Oconee Nuclear Station, Unit 1 (the facility), Renewed Facility Operating License No. DPR-38 filed by the Duke Energy Carolinas, LLC (the licensee), dated April 14, 2010, and supplemented January 30,2009, February 9,2009, February 23,2009, May 31,2009, August 3, 2009, September 29, 2009, and November 30, 2009, September 13, 2010, September 27,2010, October 14, 2010, November 19, 2010, and December 22, 2010, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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 set forth in 10 CFR Chapter I; 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.
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| -2
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| : 2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.B of Renewed Facility Operating License No. DPR-38 is hereby amended to read as follows:
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| B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 371, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
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| : 3. Accordingly, the license is hereby amended by changing the Renewed Facility Operating License No. DPR-38 fire protection License Condition 3.D to read as follows:
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| D. Fire Protection Duke Energy Carolinas, LLC, 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 revised licensee's amendment request dated April 14,2010, supplemented by letters dated: January 30, 2009, February 9, 2009, February 23, 2009, May 31,2009, August 3,2009, September 29,2009, November 30,2009, September 13, 2010, September 27,2010, October 14, 2010, November 19, 2010, and December 22, 2010, approved in the NRC safety evaluation (SE) dated December 29, 2010. 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.
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| Risk-Informed Changes that May Be Made Without Prior NRC Approval:
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| 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.
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| -3 Due to the need for the licensee to have an industry full-scope peer review of its Fire PRA and to resolve the findings of that peer review, the licensee is not allowed to self-approve quantitative risk-informed fire protection program changes, except those implementation items needing a plant change evaluation as part of the Transition License Condition below. To enable self-approval of quantitative risk-informed fire protection program changes, the licensee will need to make a 10 CFR 50.90 submittal to the NRC requesting to change this license condition. The submittal should describe how the licensee has addressed each of the peer review findings and justify the adequacy of its Fire PRA for use in this application.
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| Other Changes that May Be Made Without Prior NRC Approval:
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| : 1) Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3 fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3 element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard.
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| 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:
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| * "Fire Alarm and Detection Systems" (Section 3.8);
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| * "Automatic and Manual Water-Based Fire Suppression Systems" (Section 3.9);
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| * "Gaseous Fire Suppression Systems" (Section 3.10); and
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| * "Passive Fire Protection Features" (Section 3.11)
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| This License Condition does not apply to any demonstration of equivalency under Section 1.7 of NFPA 805.
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| -4
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| : 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 SE dated December 29, 2010, to determine that certain fire protection program changes meet the minimal risk 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.
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| Transition License Conditions
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| : 1) The licensee shall complete the items described in Section 2.9, Table 2.9-1, "Implementation Items," in the NRC SE dated December 29,2010, prior to January 1, 2013. Implementation items that result in a risk increase, as part of a plant change evaluation, can be self-approved by the licensee, as long as the overall transition risk remains a decrease (Le., collective risk increases of transition and implementation are offset by the PSW modification risk decrease.)
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| : 2) To complete the transition to full compliance with 10 CFR 50.48(c), the licensee shall implement the modifications listed in Section 2.8, Table 2.8.1-1, "Committed Plant Modifications," in the NRC SE dated December 29, 2010.
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| : 3) The licensee shall maintain appropriate compensatory measures in place until completion of all modifications and implementation items delineated above.
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| : 4. This license amendment is effective as of its date of issuance and shall be fully implemented prior to January 1, 2013.
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| FOR THE NUCLEAR REGULATORY COMMISSION Gloria Kulesa, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
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| ==Attachment:==
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| Changes to Renewed Facility Operating License No. DPR-38 and the Technical Specifications Date of Issuance: December 29,2010
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| UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 373 Renewed License No. DPR-47
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| : 1. The Nuclear Regulatory Commission (the Commission) has found that:
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| A. The application for amendment to the Oconee Nuclear Station, Unit 2 (the facility), Renewed Facility Operating License No. DPR-47 filed by the Duke Energy Carolinas, LLC (the licensee), dated April 14, 2010, and supplemented January 30, 2009, February 9, 2009, February 23, 2009, May 31, 2009, August 3, 2009, September 29, 2009, and November 30, 2009, September 13, 2010, September 27,2010, October 14,2010, November 19, 2010, and December 22, 2010, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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 set forth in 10 CFR Chapter I; 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.
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| -2
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| : 2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.B of Renewed Facility Operating License No. DPR-47 is hereby amended to read as follows:
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| B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 373, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
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| : 3. Accordingly, the license is hereby amended by changing the Renewal Facility Operating License No. DPR-47 fire protection License Condition 3.0 to read as follows:
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| D. Fire Protection Duke Energy Carolinas, LLC, 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 revised licensee's amendment request dated April 14, 2010, supplemented by letters dated: January 30, 2009, February 9, 2009, February 23, 2009, May 31,2009, August 3, 2009, September 29,2009, November 30,2009, September 13,2010, September 27,2010, October 14,2010, November 19, 2010, and December 22, 2010, approved in the NRC safety evaluation (SE) dated December 29, 2010. 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.
| |
| Risk-Informed Changes that May Be Made Without Prior NRC Approval:
| |
| 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.
| |
| | |
| -3 Due to the need for the licensee to have an industry full-scope peer review of its Fire PRA and to resolve the findings of that peer review, the licensee is not allowed to self-approve quantitative risk-informed fire protection program changes, except those implementation items needing a plant change evaluation as part of the Transition License Condition below. To enable self-approval of quantitative risk-informed fire protection program changes, the licensee will need to make a 10 CFR 50.90 submittal to the NRC requesting to change this license condition. The submittal should describe how the licensee has addressed each of the peer review findings and justify the adequacy of its Fire PRA for use in this application.
| |
| Other Changes that May Be Made Without Prior NRC Approval:
| |
| : 1) Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3 fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3 element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard.
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| The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 80S, 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.
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| | |
| -4
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| : 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 SE dated December 29,2010, to determine that certain fire protection program changes meet the minimal risk 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.
| |
| Transition License Conditions
| |
| : 1) The licensee shall complete the items described in Section 2.9, Table 2.9-1, "Implementation Items," in the NRC SE dated December 29, 2010, prior to January 1, 2013. Implementation items that result in a risk increase, as part of a plant change evaluation, can be self-approved by the licensee, as long as the overall transition risk remains a decrease (i.e., collective risk increases of transition and implementation are offset by the PSW modification risk decrease.)
| |
| : 2) To complete the transition to full compliance with 10 CFR 50.48(c), the licensee shall implement the modifications listed in Section 2.8, Table 2.8.1-1, "Committed Plant Modifications," in the NRC SE dated December 29, 2010.
| |
| : 3) The licensee shall maintain appropriate compensatory measures in place until completion of all modifications and implementation items delineated above.
| |
| : 4. This license amendment is effective as of its date of issuance and shall be fully implemented prior to January 1, 2013.
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| FOR THE NUCLEAR REGULATORY COMMISSION Gloria Kulesa, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
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| | |
| ==Attachment:==
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| | |
| Changes to Renewed Facility Operating License No. DPR-47 and the Technical Specifications Date of Issuance: December 29, 2010
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| | |
| UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 372 Renewed License No. DPR-55
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| : 1. The Nuclear Regulatory Commission (the Commission) has found that:
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| A. The application for amendment to the Oconee Nuclear Station, Unit 3 (the facility), Renewed Facility Operating License No. DPR-55 filed by the Duke Energy Carolinas, LLC (the licensee), dated April 14, 2010, and supplemented January 30,2009, February 9,2009, February 23,2009, May 31,2009, August 3, 2009, September 29, 2009, and November 30, 2009, September 13, 2010, September 27,2010, October 14,2010, November 19,2010, and December 22, 2010, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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 set forth in 10 CFR Chapter I; 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.
| |
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| -2
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| : 2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.B of Renewed Facility Operating License No. DPR-55 is hereby amended to read as follows:
| |
| B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 372, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
| |
| : 3. Accordingly, the license is hereby amended by changing the Renewed Facility Operating License No. DPR-55 ONS fire protection License Condition 3.D to read as follows:
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| D. Fire Protection Duke Energy Carolinas, LLC, 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 revised licensee's amendment request dated April 14, 2010, supplemented by letters dated: January 30, 2009, February 9, 2009, February 23, 2009, May 31, 2009, August 3, 2009, September 29, 2009, November 30, 2009, September 13,2010, September 27,2010, October 14, 2010, November 19, 2010, and December 22, 2010, approved in the NRC safety evaluation (SE) dated December 29, 2010. 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.
| |
| Risk-Informed Changes that May Be Made Without Prior NRC Approval:
| |
| 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.
| |
| | |
| -3 Due to the need for the licensee to have an industry full-scope peer review of its Fire PRA and to resolve the findings of that peer review, the licensee is not allowed to self-approve quantitative risk-informed fire protection program changes, except those implementation items needing a plant change evaluation as part of the Transition License Condition below. To enable self-approval of quantitative risk-informed fire protection program changes, the licensee will need to make a 10 CFR 50.90 submittal to the NRC requesting to change this license condition. The submittal should describe how the licensee has addressed each of the peer review findings and justify the adequacy of its Fire PRA for use in this application.
| |
| Other Changes that May Be Made Without Prior NRC Approval:
| |
| : 1) Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3 fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3 element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard.
| |
| The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3 elements are acceptable because the alternative is "adequate for the hazard." Prior NRC review and approval would not be required for alternatives to four specific sections of NFPA 805, Chapter 3 for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard. The four specific sections of NFPA 805, Chapter 3 are as follows:
| |
| * "Fire Alarm and Detection Systems" (Section 3.8);
| |
| * "Automatic and Manual Water-Based Fire Suppression Systems" (Section 3.9);
| |
| * "Gaseous Fire Suppression Systems" (Section 3.10); 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.
| |
| | |
| -4
| |
| : 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 SE dated December 29, 2010, to determine that certain fire protection program changes meet the minimal risk 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.
| |
| Transition License Conditions
| |
| : 1) The licensee shall complete the items described in Section 2.9, Table 2.9-1, "Implementation Items," in the NRC SE dated December 29, 2010, prior to January 1, 2013. Implementation items that result in a risk increase, as part of a plant change evaluation, can be self-approved by the licensee, as long as the overall transition risk remains a decrease (Le., collective risk increases of transition and implementation are offset by the PSW modification risk decrease.)
| |
| : 2) To complete the transition to full compliance with 10 CFR 50.48(c), the licensee shall implement the modifications listed in Section 2.8, Table 2.8.1-1, "Committed Plant Modifications," in the NRC SE dated December 29, 2010.
| |
| : 3) The licensee shall maintain appropriate compensatory measures in place until completion of all modifications and implementation items delineated above.
| |
| : 4. This license amendment is effective as of its date of issuance and shall be fUlly implemented prior to January 1, 2013.
| |
| FOR THE NUCLEAR REGULATORY COMMISSION Gloria Kulesa, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
| |
| | |
| ==Attachment:==
| |
| | |
| Changes to Renewed Facility Operating License No. DPR-55 and the Technical Specifications Date of Issuance: December 29, 2010
| |
| | |
| ATTACHMENT TO LICENSE AMENDMENT NO. 371 RENEWED FACILITY OPERATING LICENSE NO. DPR-38 DOCKET NO. 50-269 AND TO LICENSE AMENDMENT NO. 373 RENEWED FACILITY OPERATING LICENSE NO. DPR-47 DOCKET NO. 50-270 AND TO LICENSE AMENDMENT NO. 372 RENEWED FACILITY OPERATING LICENSE NO. DPR-55 DOCKET NO. 50-287 Replace the following pages of the Licenses and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
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| Remove Pages Insert Pages Licenses Licenses License No. DPR-38, pages 2-9 License No. DPR-38, pages 2-11 License No. DPR-47, pages 2-9 License No. DPR-47, pages 2-11 License No. DPR-55, pages 2-9 License No. DPR-55, pages 2-11 5.0-6 5.0-6
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| | |
| -2 On the basis of the foregoing findings regarding this facility, Facility Operating License No. DPR-38, issued on February 6, 1973, is superseded by Renewed Facility Operating License No. DPR-38, which is hereby issued to Duke Energy Carolinas, LLC, t~ read as follows:
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| : 1. This license applies to the Oconee Nuclear Station, Unit 1, a pressurized water reactor and associated equipment (the facility) owned and operated by Duke Energy Carolinas, LLC. The facility is located in eastern Oconee County, about eight miles northeast of Seneca, South Carolina, and is described in the "Updated Final Safety Analysis Report" (UFSAR) as supplemented and amended and the Environmental Report as supplemented and amended.
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| : 2. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses Duke Energy Carolinas, LLC (the licensee):
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| A. Pursuant to Section 104b of the Act and 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," to possess, use, and operate the facility at the designated location on the Oconee Nuclear Station Site in accordance with the procedures and limitations set forth in this license; B. Pursuant to the Act and 10 CFR Part 70, to receive, possess, and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the UFSAR as supplemented and amended; C. Pursuant to the Act and 10 CFR Parts 30, 40 and 70 to receive, possess, and use at any time 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; D. Pursuant to the Act and 10 CFR Parts 30, 40 and 70 to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components; E. 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 Oconee Nuclear Station, Units 1, 2 and 3.
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| : 3. This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I, 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; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
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| Renewed License No. DPR-38 Amendment No. 371
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| -3 A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal.
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| B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 371, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
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| C. This license is subject to the following antitrust conditions:
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| Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity.
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| Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in ,-r1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.
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| : 1. As used herein:
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| (a) "Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub transmission voltage by one electric system to another.
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| (b) "Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and transmission of electricity which meets each of Renewed License No. DPR-38 Amendment No. 371 I
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| -4 the following criteria: (1) its existing or proposed facilities are economically and technically feasible of interconnection with those of the applicant and (2) with the exception of municipalities, cooperatives, governmental agencies or authorities, and associations, it is, or upon commencement of operations will be, a public utility and subject to regulation with respect to rates and service under the laws of North Carolina or South Carolina or under the Federal Power Act; provided, however, that as to associations, each member of such association is either a public utility as discussed in this clause (2) or a municipality, a cooperative or a governmental agency or authority.
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| (c) Where the phrase "neighboring entity" is intended to include entities engaging or proposing to engage only in the distribution of electricity, this is indicated by adding the phrase "including distribution systems".
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| (d) "Cost" means any appropriate operating and maintenance expenses, together with all other costs, including a reasonable return on applicant's investment, which are reasonably allocable to a transaction. However, no value shall be included for loss of revenues due to the loss of any wholesale or retail customer as a result of any transaction hereafter described.
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| : 2. (a) Applicant will interconnect and coordinate reserves by means of the sale and exchange of emergency and scheduled maintenance bulk power with any neighboring entity(ies), when there are net benefits to each party, on terms that will provide for all of applicant's properly assignable costs as may be determined by the Federal Energy Regulatory Commission and consistent with such cost assignment will allow the other party the fullest possible benefits of such coordination.
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| (b) Emergency service and/or scheduled maintenance service to be provided by each party will be furnished to the fullest extent available from the supplying party and desired by the party in need. Applicant and each party will provide to the other emergency service and/or scheduled maintenance service if and when available from its own generation and, in accordance with recognized industry practice, from generation of others to the extent it can do so without impairing service to its customers, including other electric systems to whom it has firm commitments.
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| Renewed License No. DPR-38 Amendment No. 371
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| -5 (c) Each party to a reserve coordination arrangement will establish its own reserve criteria, but in no event shall the minimum installed reserve on each system be less than 15%, calculated as a percentage of estimated peak load responsibility. Either party, if it has, or has firmly planned, installed reserves in excess of the amount called for by its own reserve criterion, will offer any such excess as may in fact be available at the time for which it is sought and for such period as the selling party shall determine for purchase in accordance with reasonable industry practice by the other party to meet such other party's own reserve requirement.
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| The parties will provide such amounts of spinning reserve as may be adequate to avoid the imposition of unreasonable demands on the other party(ies) in meeting the normal contingencies of operating its (their) system(s). However, in no circumstances shall such spinning reserve requirement exceed the installed reserve requirement.
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| (d) Interconnections will not be limited to low voltages when higher voltages are available from applicant's installed facilities in the area where interconnection is desired and when the proposed arrangement is found to be technically and economically feasible.
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| (e) Interconnection and reserve coordination agreements will not embody provisions which impose limitations upon the use or resale of power and energy sold or exchanges pursuant to the agreement. Further, such arrangements will not prohibit the participants from entering into other interconnection and coordination arrangements, but may include appropriate provisions to assure that (i) applicant receives adequate notice of such additional interconnection or coordination, (ii) the parties will jointly consider and agree upon such measures, if any, as are reasonably necessary to protect the reliability of the interconnected systems and to prevent undue burdens from being imposed on any system, and (iii) applicant will be fully compensated for its costs. Reasonable industry practice as developed in the area from time to time will satisfy this provision.
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| : 3. Applicant currently has on file, and may hereafter file, with the Federal Energy Regulatory Commission contracts with neighboring entity(ies) providing for the sale and exchange of short-term power and energy, limited term power and energy, economy energy, nondisplacement energy, and emergency capacity and energy. Applicant will enter into contracts providing for the same or for like transactions with any neighboring entity on terms which enable applicant to recover the full costs allocable to such transaction.
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| Renewed License No. DPR-38 Amendment No. 371
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| -6
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| : 4. Applicant currently sells capacity and energy in bulk on a full requirements basis to several entities engaging in the distribution of electric power at retail. In addition, applicant supplies electricity directly to ultimate users in a number of municipalities. Should any such entity(ies) or municipality(ies) desire to become a neighboring entity as defined in ~'1 (b) hereof (either alone or through combination with other), applicant will assist in facilitating the necessary transition through the sale of partial requirements firm power and energy. The provision of such firm partial requirements service shall be under such rates, terms and conditions as shall be found by the Federal Energy Regulatory Commission to provide for the recovery of applicant's costs. Applicant will sell capacity and energy in bulk on a full requirements basis to any municipality currently served by applicant when such municipality lawfully engages in the distribution of electric power at retail.
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| : 5. (a) Applicant will facilitate the exchange of electric power in bulk in wholesale transactions over its transmission facilities (1) between or among two or more neighboring entities, including distribution systems with which it is interconnected or may be interconnected in the future, and (2) between any such entity(ies) and any other electric system engaging in bulk power supply between whose facilities applicant's transmission lines and other transmission lines would form a continuous electric path, provided that permission to utilize such other transmission lines has been obtained. Such transaction shall be undertaken provided that the particular transaction reasonably can be accommodated by applicant's transmission system from a functional and technical standpoint and does not constitute the wheeling of power to a retail customer. Such transmission shall be on terms that fully compensate applicant for its cost. Any entity(ies) requesting such transmission arrangements shall give reasonable notice of its (their) schedule and requirements.
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| (b) Applicant will include in its planning and construction program, sufficient transmission capacity as required for the transactions referred to in subparagraph (a) of this paragraph, provided that (1) the neighboring entity(ies) gives applicant sufficient advance notice as may be necessary reasonably to accommodate its (their) requirements from a functional and technical standpoint and (2) that such entity(ies) fully compensates applicant for its cost. In carrying out this subparagraph (b), however, applicant shall not be required to construct or add transmission facilities which (a) will be of no demonstrable present or future benefit to applicant, or (b) which could be constructed by the requesting entity(ies) without duplicating any portion of applicant's existing transmission lines, or (c) which would jeopardize applicant's ability to finance or construct on reasonable terms facilities needed to meet its own anticipated system requirements. Where regulatory or environmental approvals are required for the construction or addition of transmission Renewed License No. DPR-38 Amendment No. 371
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| -7 facilities, needed for the transactions referred to in subparagraph (a) of this paragraph, it shall be the responsibility of the entity(ies) seeking the transaction to participate in obtaining such approvals, including sharing in the cost thereof.
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| : 6. To increase the possibility of achieving greater reliability and economy of electric generation and transmission facilities, applicant will discuss load projections and system development plans with any neighboring entity(ies).
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| : 7. When applicant's plans for future nuclear generating units (for which application will hereafter be made to the Nuclear Regulatory Commission) have reached the stage of serious planning, but before firm decisions have been made as to the size and desired completion date of the proposed nuclear units, applicant will notify all neighboring entities, including distribution systems with peak loads smaller than applicant's, that applicant plans to construct such nuclear units. Neither the timing nor the information provided need be such as to jeopardize obtaining the required site at the lowest possible cost.
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| : 8. The foregoing commitments shall be implemented in a manner consistent with the provisions of the Federal Power Act and all other lawful local, State and Federal regulation and authority. Nothing in these commitments is intended to determine in advance the resolution of issues which are properly raised at the Federal Energy Regulatory Commission concerning such commitments, including allocation of costs or the rates to be charged. Applicant will negotiate (including the execution of a contingent statement of intent) with respect to the foregoing commitments with any neighboring entity including distribution systems where applicable engaging in or proposing to engage in bulk power supply transactions, but applicant shall not be required to enter into any final arrangement prior to resolution of any substantial questions as to the lawful authority of an entity to engage in the transactions. In addition, applicant shall not be obligated to enter into a given bulk power supply transaction if: (1) to do so would violate, or incapacitate it from performing any existing lawful contract it has with a third party; (2) there is contemporaneously available to it, a competing or alternative arrangement which affords it greater benefits which would be mutually exclusive of such arrangement; (3) to do so would adversely affect its system operations or the reliability of power supply to its customers; or (4) if to do so would jeopardize applicant's ability to finance or construct on reasonable terms facilities needed to meet its own anticipated system requirements.
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| Renewed License No. DPR-38 Amendment No. 371
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| -8 D. Fire Protection Duke Energy Carolinas, LLC, 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 revised licensee's amendment request dated April 14, 2010, supplemented by letters dated:
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| January 30, 2009, February 9, 2009, February 23, 2009, May 31, 2009, August 3, 2009, September 29, 2009, November 30, 2009, September 13, 2010, September 27,2010, October 14,2010, November 19, 2010, and December 22, 2010, approved in the NRC safety evaluation (SE) dated December 29, 2010. 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.
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| Risk-Informed Changes that May Be Made Without Prior NRC Approval:
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| 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.
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| Due to the need for the licensee to have an industry full-scope peer review of its Fire PRA and to resolve the findings of that peer review, the licensee is not allowed to self-approve quantitative risk-informed fire protection program changes, except those implementation items needing a plant change evaluation as part of the Transition License Condition below. To enable self-approval of quantitative risk informed fire protection program changes, the licensee will need to make a 10 CFR 50.90 submittal to the NRC requesting to change this license condition. The submittal should describe how the licensee has addressed each of the peer review findings and justify the adequacy of its Fire PRA for use in this application.
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| Other Changes that May Be Made Without Prior NRC Approval:
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| : 1) Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3 fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3 element is functionally equivalent to the corresponding Renewed License No. DPR-38 Amendment No. 371
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| - 9 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.
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| 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:
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| * "Fire Alarm and Detection Systems" (Section 3.8);
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| * "Automatic and Manual Water-Based Fire Suppression Systems" (Section 3.9);
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| * "Gaseous Fire Suppression Systems" (Section 3.10); and
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| * "Passive Fire Protection Features" (Section 3.11)
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| This License Condition does not apply to any demonstration of equivalency under Section 1.7 of NFPA 805.
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| : 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 SE dated December 29,2010, to determine that certain fire protection program changes meet the minimal risk 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.
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| Transition License Conditions
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| : 1) The licensee shall complete the items described in Section 2.9, Table 2.9-1, "Implementation Items," in the NRC SE dated December 29, 2010, prior to January 1, 2013. Implementation items that result in a risk increase, as part of a plant change evaluation, can be self-approved by the licensee, as long as the overall transition risk remains a decrease (i.e., collective risk increases of transition and implementation are offset by the PSW modification risk decrease).
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| : 2) To complete the transition to full compliance with 10 CFR 50.48(c), the licensee shall implement the modifications listed in Section 2.8, Table 2.8.1-1, "Committed Plant Modifications," in the NRC SE dated December 29,2010.
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| Renewed License No. DPR-38 Amendment No. 371
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| - 10
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| : 3) The licensee shall maintain appropriate compensatory measures in place until completion of all modifications and implementation items delineated above.
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| E. Physical Protection Duke Energy Carolinas, LLC, 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 27817 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: "Duke Energy Physical Security Plan" submitted by letter dated September 8, 2004, and supplemented on September 30, 2004, October 15, 2004, October 21,2004, and October 27,2004.
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| F. In the update to the UFSAR required pursuant to 10 CFR 50.71 (e)(4) scheduled for July, 2001, the licensee shall update the UFSAR to include the UFSAR supplement submitted pursuant to 10 CFR 54.21(d) as revised on March 27, 2000. Until the UFSAR update is complete, the licensee may make changes to the programs described in its UFSAR 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.
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| G. The licensee's UFSAR supplement submitted pursuant to 10 CFR 54.21(d), as revised on March 27, 2000, describes certain future inspection activities to be completed before the period of extended operation. The licensee shall complete these activities no later than February 6, 2013.
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| H. Mitigation Strategy License Condition Develop and maintain strategies for addressing large fires and explosions and that include the following key areas:
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| (a) Fire fighting response strategy with the following elements:
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| : 1. Pre-defined coordinated fire response strategy and guidance
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| : 2. Assessment of mutual aid fire fighting assets
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| : 3. Designated staging areas for equipment and materials
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| : 4. Command and control
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| : 5. Training of response personnel Renewed License No. DPR-38 Amendment No. 371
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| - 11 (b) Operations to mitigate fuel damage considering the following:
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| : 1. Protection and use of personnel assets
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| : 2. Communications
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| : 3. Minimizing fire spread
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| : 4. Procedures for implementing integrated fire response strategy
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| : 5. Identification of readily-available pre-staged equipment
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| : 6. Training on integrated fire response strategy
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| : 7. SFP mitigation measures (c) Actions to minimize release to include consideration of:
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| : 1. Water spray scrubbing
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| : 2. Dose to onsite responders
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| : 4. This renewed license is effective as of the date of issuance and shall expire at midnight on February 6, 2033.
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| FOR THE NUCLEAR REGULATORY COMMISSION Original signed by Roy P. Zimmerman Roy Zimmerman, Acting Director Office of Nuclear Reactor Regulation
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| ==Attachment:==
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| : 1) Appendix A - Technical Specifications Renewed License No. DPR-38 Date of Issuance: May 23, 2000 Renewed License No. DPR-38 Amendment No. 371 I
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| -2 On the basis of the foregoing findings regarding this facility, Facility Operating License No.
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| DPR-47, issued on October 6, 1973, is superseded by Renewed Facility Operating License No.
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| DPR-47, which is hereby issued to Duke Energy Carolinas, LLC, to read as follows:
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| : 1. This license applies to the Oconee Nuclear Station, Unit 2, a pressurized water reactor and associated equipment (the facility) owned and operated by Duke Energy Carolinas, LLC. The facility is located in eastern Oconee County, about eight miles northeast of Seneca, South Carolina, and is described in the "Updated Final Safety Analysis Report" (UFSAR) as supplemented and amended and the Environmental Report as supplemented and amended.
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| : 2. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses Duke Energy Carolinas, LLC (the licensee):
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| A. Pursuant to Section 104b of the Act and 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," to possess, use, and operate the facility at the designated location on the Oconee Nuclear Station Site in accordance with the procedures and limitations set forth in this license; B. Pursuant to the Act and 10 CFR Part 70, to receive, possess, and use at any time special nuclear material as reactor fuel in accordance with the limitations for storage and amounts required for reactor operation, as described in the UFSAR as supplemented and amended; C. Pursuant to the Act and 10 CFR Parts 30, 40 and 70 to receive, possess, and use at any time 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; D. Pursuant to the Act and 10 CFR Parts 30, 40 and 70 to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components; E. 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 Oconee Nuclear Station, Units 1, 2 and 3.
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| : 3. This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I, 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; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
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| Renewed License No. DPR-47 Amendment No. 373
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| -3 A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal.
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| B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 373, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
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| : c. This license is subject to the following antitrust conditions:
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| Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity.
| |
| Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in ~1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.
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| : 1. As used herein:
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| (a) "Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another.
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| (b) "Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and transmission of electricity which meets each of Renewed License No. DPR-47 Amendment No. 373
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| -4 the following criteria: (1) its existing or proposed facilities are economically and technically feasible of interconnection with those of the applicant and (2) with the exception of municipalities, cooperatives, governmental agencies or authorities, and associations, it is, or upon commencement of operations will be, a public utility and subject to regulation with respect to rates and service under the laws of North Carolina or South Carolina or under the Federal Power Act; provided, however, that as to associations, each member of such association is either a public utility as discussed in this clause (2) or a municipality, a cooperative or a governmental agency or authority.
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| (c) Where the phrase "neighboring entity" is intended to include entities engaging or proposing to engage only in the distribution of electricity, this is indicated by adding the phrase "including distribution systems".
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| (d) "Cost" means any appropriate operating and maintenance expenses, together with all other costs, including a reasonable return on applicant's investment, which are reasonably allocable to a transaction. However, no value shall be included for loss of revenues due to the loss of any wholesale or retail customer as a result of any transaction hereafter described.
| |
| : 2. (a) Applicant will interconnect and coordinate reserves by means of the sale and exchange of emergency and scheduled maintenance bulk power with any neighboring entity(ies), when there are net benefits to each party, on terms that will provide for all of applicant's properly assignable costs as may be determined by the Federal Energy Regulatory Commission and consistent with such cost assignment will allow the other party the fullest possible benefits of such coordination.
| |
| (b) Emergency service and/or scheduled maintenance service to be provided by each party will be furnished to the fullest extent available from the supplying party and desired by the party in need. Applicant and each party will provide to the other emergency service and/or scheduled maintenance service if and when available from its own generation and, in accordance with recognized industry practice, from generation of others to the extent it can do so without impairing service to its customers, including other electric systems to whom it has firm commitments.
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| Renewed License No. DPR-47 Amendment No. 373 I
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| | |
| - 5 (c) Each party to a reserve coordination arrangement will establish its own reserve criteria, but in no event shall the minimum installed reserve on each system be less than 15%, calculated as a percentage of estimated peak load responsibility. Either party, if it has, or has firmly planned, installed reserves in excess of the amount called for by its own reserve criterion, will offer any such excess as may in fact be available at the time for which it is sought and for such period as the selling party shall determine for purchase in accordance with reasonable industry practice by the other party to meet such other party's own reserve requirement. The parties will provide such amounts of spinning reserve as may be adequate to avoid the imposition of unreasonable demands on the other party(ies) in meeting the normal contingencies of operating its (their) system(s). However, in no circumstances shall such spinning reserve requirement exceed the installed reserve requirement.
| |
| (d) Interconnections will not be limited to low voltages when higher voltages are available from applicant's installed facilities in the area where interconnection is desired and when the proposed arrangement is found to be technically and economically feasible.
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| (e) Interconnection and reserve coordination agreements will not embody provisions which impose limitations upon the use or resale of power and energy sold or exchanges pursuant to the agreement. Further, such arrangements will not prohibit the participants from entering into other interconnection and coordination arrangements, but may include appropriate provisions to assure that (i) applicant receives adequate notice of such additional interconnection or coordination, (ii) the parties will jointly consider and agree upon such measures, if any, as are reasonably necessary to protect the reliability of the interconnected systems and to prevent undue burdens from being imposed on any system, and (iii) applicant will be fully compensated for its costs. Reasonable industry practice as developed in the area from time to time will satisfy this provision.
| |
| : 3. Applicant currently has on file, and may hereafter file, with the Federal Energy Regulatory Commission contracts with neighboring entity(ies) providing for the sale and exchange of short-term power and energy, limited term power and energy, economy energy, nondisplacement energy, and emergency capacity and energy. Applicant will enter into contracts providing for the same or for like transactions with any neighboring entity on terms which enable applicant to recover the full costs allocable to such transaction.
| |
| Renewed License No. DPR-47 Amendment No. 373
| |
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| -6
| |
| : 4. Applicant currently sells capacity and energy in bulk on a full requirements basis to several entities engaging in the distribution of electric power at retail. In addition, applicant supplies electricity directly to ultimate users in a number of municipalities. Should any such entity(ies) or municipality(ies) desire to become a neighboring entity as defined in 111 (b) hereof (either alone or through combination with other), applicant will assist in facilitating the necessary transition through the sale of partial requirements firm power and energy. The provision of such firm partial requirements service shall be under such rates, terms and conditions as shall be found by the Federal Energy Regulatory Commission to provide for the recovery of applicant's costs. Applicant will sell capacity and energy in bulk on a full requirements basis to any municipality currently served by applicant when such municipality lawfully engages in the distribution of electric power at retail.
| |
| : 5. (a) Applicant will facilitate the exchange of electric power in bulk in wholesale transactions over its transmission facilities (1) between or among two or more neighboring entities, including distribution systems with which it is interconnected or may be interconnected in the future, and (2) between any such entity(ies) and any other electric system engaging in bulk power supply between whose facilities applicant's transmission lines and other transmission lines would form a continuous electric path, provided that permission to utilize such other transmission lines has been obtained. Such transaction shall be undertaken provided that the particular transaction reasonably can be accommodated by applicant's transmission system from a functional and technical standpoint and does not constitute the wheeling of power to a retail customer. Such transmission shall be on terms that fully compensate applicant for its cost. Any entity(ies) requesting such transmission arrangements shall give reasonable notice of its (their) schedule and requirements.
| |
| (b) Applicant will include in its planning and construction program, sufficient transmission capacity as required for the transactions referred to in subparagraph (a) of this paragraph, provided that (1) the neighboring entity(ies) gives applicant sufficient advance notice as may be necessary reasonably to accommodate its (their) requirements from a functional and technical standpoint and (2) that such entity(ies) fully compensates applicant for its cost. In carrying out this subparagraph (b), however, applicant shall not be required to construct or add transmission facilities which (a) will be of no demonstrable present or future benefit to applicant, or (b) which could be constructed by the requesting entity(ies) without duplicating any portion of applicant's existing transmission lines, or Renewed License No. DPR-47 Amendment No. 373
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| | |
| -7 (c) which would jeopardize applicant's ability to finance or construct on reasonable terms facilities needed to meet its own anticipated system requirements. Where regulatory or environmental approvals are required for the construction or addition of transmission facilities, needed for the transactions referred to in subparagraph (a) of this paragraph, it shall be the responsibility of the entity(ies) seeking the transaction to participate in obtaining such approvals, including sharing in the cost thereof.
| |
| : 6. To increase the possibility of achieving greater reliability and economy of electric generation and transmission facilities, applicant will discuss load projections and system development plans with any neighboring entity(ies).
| |
| : 7. When applicant's plans for future nuclear generating units (for which application will hereafter be made to the Nuclear Regulatory Commission) have reached the stage of serious planning, but before firm decisions have been made as to the size and desired completion date of the proposed nuclear units, applicant will notify all neighboring entities, including distribution systems with peak loads smaller than applicant's, that applicant plans to construct such nuclear units. Neither the timing nor the information provided need be such as to jeopardize obtaining the required site at the lowest possible cost.
| |
| : 8. The foregoing commitments shall be implemented in a manner consistent with the provisions of the Federal Power Act and all other lawful local, State and Federal regulation and authority. Nothing in these commitments is intended to determine in advance the resolution of issues which are properly raised at the Federal Energy Regulatory Commission concerning such commitments, including allocation of costs or the rates to be charged. Applicant will negotiate (including the execution of a contingent statement of intent) with respect to the foregoing commitments with any neighboring entity including distribution systems where applicable engaging in or proposing to engage in bulk power supply transactions, but applicant shall not be required to enter into any final arrangement prior to resolution of any substantial questions as to the lawful authority of an entity to engage in the transactions. In addition, applicant shall not be obligated to enter into a given bulk power supply transaction if: (1) to do so would violate, or incapacitate it from performing any existing lawful contract it has with a third party; (2) there is contemporaneously available to it, a competing or alternative arrangement which affords it greater benefits which would be mutually exclusive of such arrangement; (3) to do so would adversely affect its system operations or the reliability of power supply to its customers; or (4) if to do so would jeopardize applicant's ability to finance or construct on reasonable terms facilities needed to meet its own anticipated system requirements.
| |
| Renewed License No. DPR-47 Amendment No. 373
| |
| | |
| -8 D. Fire Protection Duke Energy Carolinas, LLC, 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 revised licensee's amendment request dated April 14, 2010, supplemented by letters dated:
| |
| January 30, 2009, February 9, 2009, February 23, 2009, May 31,2009, August 3, 2009, September 29, 2009, November 30, 2009, September 13, 2010, September 27,2010, October 14,2010, November 19, 2010, and December 22, 2010, approved in the NRC safety evaluation (SE) dated December 29, 2010. 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.
| |
| Risk-Informed Changes that May Be Made Without Prior NRC Approval:
| |
| 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.
| |
| Due to the need for the licensee to have an industry full-scope peer review of its Fire PRA and to resolve the findings of that peer review, the licensee is not allowed to self approve quantitative risk-informed fire protection program changes, except those implementation items needing a plant change evaluation as part of the Transition License Condition below. To enable self-approval of quantitative risk-informed fire protection program changes, the licensee will need to make a 10 CFR 50.90 submittal to the NRC requesting to change this license condition. The submittal should describe how the licensee has addressed each of the peer review findings and justify the adequacy of its Fire PRA for use in this application.
| |
| Other Changes that May Be Made Without Prior NRC Approval:
| |
| : 1) Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3 fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3 element is functionally equivalent to the corresponding Renewed License No. DPR-47 Amendment No. 373
| |
| | |
| - 9 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 SE dated December 29, 2010, to determine that certain fire protection program changes meet the minimal risk 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.
| |
| Transition License Conditions
| |
| : 1) The licensee shall complete the items described in Section 2.9, Table 2.9-1, "Implementation Items," in the NRC SE dated December 29, 2010, prior to January 1, 2013. Implementation items that result in a risk increase, as part of a plant change evaluation, can be self-approved by the licensee, as long as the overall transition risk remains a decrease (i.e., collective risk increases of transition and implementation are offset by the PSW modification risk decrease).
| |
| : 2) To complete the transition to full compliance with 10 CFR 50.48(c), the licensee shall implement the modifications listed in Section 2.8, Table 2.8.1-1, "Committed Plant Modifications," in the NRC SE dated December 29,2010.
| |
| Renewed License No. DPR-47 Amendment No. 373
| |
| | |
| - 10
| |
| : 3) The licensee shall maintain appropriate compensatory measures in place until completion of all modifications and implementation items delineated above.
| |
| E. Physical Protection Duke Energy Carolinas, LLC, 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 27817 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: "Duke Energy Physical Security Plan" submitted by letter dated September 8, 2004, and supplemented on September 30, 2004, October 15, 2004, October 21, 2004, and October 27, 2004.
| |
| F. In the update to the UFSAR required pursuant to 10 CFR 50.71 (e)(4) scheduled for July, 2001, the licensee shall update the UFSAR to include the UFSAR supplement submitted pursuant to 10 CFR 54.21 (d) as revised on March 27, 2000. Until the UFSAR update is complete, the licensee may make changes to the programs described in its UFSAR 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. The licensee's UFSAR supplement submitted pursuant to 10 CFR 54.21(d), as revised on March 27, 2000, describes certain future inspection activities to be completed before the period of extended operation. The licensee shall complete these activities no later than February 6, 2013.
| |
| H. Mitigation Strategy License Condition 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 Renewed License No. DPR-47 Amendment No. 373
| |
| | |
| - 11 (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. SFP mitigation measures (c) Actions to minimize release to include consideration of:
| |
| : 1. Water spray scrubbing
| |
| : 2. Dose to onsite responders
| |
| : 4. This renewed license is effective as of the date of issuance and shall expire at midnight on October 6, 2033.
| |
| FOR THE NUCLEAR REGULATORY COMMISSION Original Signed By Roy P. Zimmerman Roy P. Zimmerman, Acting Director Office of Nuclear Reactor Regulation
| |
| | |
| ==Attachment:==
| |
| : 1) Appendix A - Technical Specifications Renewed License No. DPR-47 Date of issuance: May 23, 2000 Renewed License No. DPR-47 Amendment No. 373
| |
| | |
| -2 On the basis of the foregoing findings regarding this facility, Facility Operating License No.
| |
| OPR-55, issued on July 19, 1974, is superseded by Renewed Facility Operating License No.
| |
| OPR-55, which is hereby issued to Duke Energy Carolinas, LLC, to read as follows:
| |
| : 1. This license applies to the Oconee Nuclear Station, Unit 3, a pressurized water reactor and associated equipment (the facility) owned and operated by Duke Energy Carolinas, LLC. The facility is located in eastern Oconee County, about eight miles northeast of Seneca, South Carolina, and is described in the "Updated Final Safety Analysis Report" (UFSAR) as supplemented and amended and the Environmental Report as supplemented and amended.
| |
| : 2. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses Duke Energy Carolinas, LLC (the licensee):
| |
| A. Pursuant to Section 104b of the Act and 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," to possess, use, and operate the facility at the designated location on the Oconee Nuclear Station Site in accordance with the procedures and limitations set forth in this license; B. Pursuant to the Act and 10 CFR Part 70, to receive, possess, and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the UFSAR as supplemented and amended; C. Pursuant to the Act and 10 CFR Parts 30, 40 and 70 to receive, possess, and use at any time 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; O. Pursuant to the Act and 10 CFR Parts 30, 40 and 70 to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components; E. 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 Oconee Nuclear Station, Units 1, 2 and 3.
| |
| : 3. This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I, 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; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
| |
| Renewed License No. OPR-55 Amendment No. 372 I
| |
| | |
| -3 A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal.
| |
| B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 372, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
| |
| : c. This license is subject to the following antitrust conditions:
| |
| Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity.
| |
| Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in 111 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.
| |
| : 1. As used herein:
| |
| (a) "Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another.
| |
| (b) "Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and transmission of electricity which meets each of Renewed License No. OPR-55 Amendment No. 372
| |
| | |
| -4 the following criteria: (1) its existing or proposed facilities are economically and technically feasible of interconnection with those of the applicant and (2) with the exception of municipalities, cooperatives, governmental agencies or authorities, and associations, it is, or upon commencement of operations will be, a public utility and subject to regulation with respect to rates and service under the laws of North Carolina or South Carolina or under the Federal Power Act; provided, however, that as to associations, each member of such association is either a public utility as discussed in this clause (2) or a municipality, a cooperative or a governmental agency or authority.
| |
| (c) Where the phrase "neighboring entity" is intended to include entities engaging or proposing to engage only in the distribution of electricity, this is indicated by adding the phrase "including distribution systems".
| |
| (d) "Cost" means any appropriate operating and maintenance expenses, together with all other costs, including a reasonable return on applicant's investment, which are reasonably allocable to a transaction. However, no value shall be included for loss of revenues due to the loss of any wholesale or retail customer as a result of any transaction hereafter described.
| |
| : 2. (a) Applicant will interconnect and coordinate reserves by means of the sale and exchange of emergency and scheduled maintenance bulk power with any neighboring entity(ies), when there are net benefits to each party, on terms that will provide for all of applicant's properly assignable costs as may be determined by the Federal Energy Regulatory Commission and consistent with such cost assignment will allow the other party the fullest possible benefits of such coordination.
| |
| (b) Emergency service and/or scheduled maintenance service to be provided by each party will be furnished to the fullest extent available from the supplying party and desired by the party in need. Applicant and each party will provide to the other emergency service and/or scheduled maintenance service if and when available from its own generation and, in accordance with recognized industry practice, from generation of others to the extent it can do so without impairing service to its customers, including other electric systems to whom it has firm commitments.
| |
| Renewed License No. DPR-55 Amendment No. 372
| |
| | |
| -5 (c) Each party to a reserve coordination arrangement will establish its own reserve criteria, but in no event shall the minimum installed reserve on each system be less than 15%, calculated as a percentage of estimated peak load responsibility. Either party, if it has, or has firmly planned, installed reserves in excess of the amount called for by its own reserve criterion, will offer any such excess as may in fact be available at the time for which it is sought and for such period as the selling party shall determine for purchase in accordance with reasonable industry practice by the other party to meet such other party's own reserve requirement.
| |
| The parties will provide such amounts of spinning reserve as may be adequate to avoid the imposition of unreasonable demands on the other party(ies) in meeting the normal contingencies of operating its (their) system(s). However, in no circumstances shall such spinning reserve requirement exceed the installed reserve requirement.
| |
| (d) Interconnections will not be limited to low voltages when higher voltages are available from applicant's installed facilities in the area where interconnection is desired and when the proposed arrangement is found to be technically and economically feasible.
| |
| (e) Interconnection and reserve coordination agreements will not embody provisions which impose limitations upon the use or resale of power and energy sold or exchanges pursuant to the agreement. Further, such arrangements will not prohibit the participants from entering into other interconnection and coordination arrangements, but may include appropriate provisions to assure that (i) applicant receives adequate notice of such additional interconnection or coordination, (ii) the parties will jointly consider and agree upon such measures, if any, as are reasonably necessary to protect the reliability of the interconnected systems and to prevent undue burdens from being imposed on any system, and (iii) applicant will be fully compensated for its costs. Reasonable industry practice as developed in the area from time to time will satisfy this provision.
| |
| : 3. Applicant currently has on file, and may hereafter file, with the Federal Energy Regulatory Commission contracts with neighboring entity(ies) providing for the sale and exchange of short-term power and energy, limited term power and energy, economy energy, nondisplacement energy, and emergency capacity and energy. Applicant will enter into contracts providing for the same or for like transactions with any neighboring entity on terms which enable applicant to recover the full costs allocable to such transaction.
| |
| Renewed License No. OPR-55 Amendment No. 372
| |
| | |
| - 6
| |
| : 4. Applicant currently sells capacity and energy in bulk on a full requirements basis to several entities engaging in the distribution of electric power at retail. In addition, applicant supplies electricity directly to ultimate users in a number of municipalities. Should any such entity(ies) or municipality(ies) desire to become a neighboring entity as defined in
| |
| ~1 (b) hereof (either alone or through combination with other), applicant will assist in facilitating the necessary transition through the sale of partial requirements firm power and energy. The provision of such firm partial requirements service shall be under such rates, terms and conditions as shall be found by the Federal Energy Regulatory Commission to provide for the recovery of applicant's costs. Applicant will sell capacity and energy in bulk on a full requirements basis to any municipality currently served by applicant when such municipality lawfully engages in the distribution of electric power at retail.
| |
| : 5. (a) Applicant will facilitate the exchange of electric power in bulk in wholesale transactions over its transmission facilities (1) between or among two or more neighboring entities, including distribution systems with which it is interconnected or may be interconnected in the future, and (2) between any such entity(ies) and any other electric system engaging in bulk power supply between whose facilities applicant's transmission lines and other transmission lines would form a continuous electric path, provided that permission to utilize such other transmission lines has been obtained. Such transaction shall be undertaken provided that the particular transaction reasonably can be accommodated by applicant's transmission system from a functional and technical standpoint and does not constitute the wheeling of power to a retail customer.
| |
| Such transmission shall be on terms that fully compensate applicant for its cost. Any entity(ies) requesting such transmission arrangements shall give reasonable notice of its (their) schedule and requirements.
| |
| (b) Applicant will include in its planning and construction program, sufficient transmission capacity as required for the transactions referred to in subparagraph (a) of this paragraph, provided that (1) the neighboring entity(ies) gives applicant sufficient advance notice as may be necessary reasonably to accommodate its (their) requirements from a functional and technical standpoint and (2) that such entity(ies) fully compensates applicant for its cost.
| |
| In carrying out this subparagraph (b), however, applicant shall not be required to construct or add transmission facilities which (a) will be of no demonstrable present or future benefit to applicant, or (b) which could be constructed by the requesting entity(ies) without duplicating any portion of applicant's existing transmission lines, or (c) which would jeopardize applicant's ability to finance or construct Renewed License No. OPR-55 Amendment No. 372
| |
| | |
| -7 on reasonable terms facilities needed to meet its own anticipated system requirements. Where regulatory or environmental approvals are required for the construction or addition of transmission facilities, needed for the transactions referred to in subparagraph (a) of this paragraph, it shall be the responsibility of the entity(ies) seeking the transaction to participate in obtaining such approvals, including sharing in the cost thereof.
| |
| : 6. To increase the possibility of achieving greater reliability and economy of electric generation and transmission facilities, applicant will discuss load projections and system development plans with any neighboring entity(ies).
| |
| : 7. When applicant's plans for future nuclear generating units (for which application will hereafter be made to the Nuclear Regulatory Commission) have reached the stage of serious planning, but before firm decisions have been made as to the size and desired completion date of the proposed nuclear units, applicant will notify all neighboring entities, including distribution systems with peak loads smaller than applicant's, that applicant plans to construct such nuclear units. Neither the timing nor the information provided need be such as to jeopardize obtaining the required site at the lowest possible cost.
| |
| : 8. The foregoing commitments shall be implemented in a manner consistent with the provisions of the Federal Power Act and all other lawful local, State and Federal regulation and authority. Nothing in these commitments is intended to determine in advance the resolution of issues which are properly raised at the Federal Energy Regulatory Commission concerning such commitments, including allocation of costs or the rates to be charged. Applicant will negotiate (including the execution of a contingent statement of intent) with respect to the foregoing commitments with any neighboring entity including distribution systems where applicable engaging in or proposing to engage in bulk power supply transactions, but applicant shall not be required to enter into any final arrangement prior to resolution of any substantial questions as to the lawful authority of an entity to engage in the transactions. In addition, applicant shall not be obligated to enter into a given bulk power supply transaction if: (1) to do so would violate, or incapacitate it from performing any existing lawful contract it has with a third party; (2) there is contemporaneously available to it, a competing or alternative arrangement which affords it greater benefits which would be mutually exclusive of such arrangement; (3) to do so would adversely affect its system operations or the reliability of power supply to its customers; or (4) if to do so would jeopardize applicant's ability to finance or construct on reasonable terms facilities needed to meet its own anticipated system requirements.
| |
| Renewed License No. DPR-55 Amendment No. 372
| |
| | |
| -8 D. Fire Protection Duke Energy Carolinas, LLC, 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 revised licensee's amendment request dated April 14, 2010, supplemented by letters dated:
| |
| January 30,2009, February 9, 2009, February 23,2009, May 31,2009, August 3, 2009, September 29, 2009, November 30, 2009, September 13, 2010, September 27,2010, October 14, 2010, November 19, 2010, and December 22, 2010, approved in the NRC safety evaluation (SE) dated December 29, 2010. 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.
| |
| Risk-Informed Changes that May Be Made Without Prior NRC Approval:
| |
| 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.
| |
| Due to the need for the licensee to have an industry full-scope peer review of its Fire PRA and to resolve the findings of that peer review, the licensee is not allowed to self-approve quantitative risk-informed fire protection program changes, except those implementation items needing a plant change evaluation as part of the Transition License Condition below. To enable self approval of quantitative risk-informed fire protection program changes, the licensee will need to make a 10 CFR 50.90 submittal to the NRC requesting to change this license condition. The submittal should describe how the licensee has addressed each of the peer review findings and justify the adequacy of its Fire PRA for use in this application.
| |
| Other Changes that May Be Made Without Prior NRC Approval:
| |
| : 1) Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3 fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3 element is functionally equivalent to the Renewed License No. DPR-55 Amendment No. 372
| |
| | |
| -9 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 SE dated December 29, 2010, to determine that certain fire protection program changes meet the minimal risk 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.
| |
| Transition License Conditions
| |
| : 1) The licensee shall complete the items described in Section 2.9, Table 2.9-1, "Implementation Items," in the NRC SE dated December 29, 2010, prior to January 1, 2013. Implementation items that result in a risk increase, as part of a plant change evaluation, can be self-approved by the licensee, as long as the overall transition risk remains a decrease (i.e., collective risk increases of transition and implementation are offset by the PSW modification risk decrease).
| |
| : 2) To complete the transition to full compliance with 10 CFR 50.48(c), the licensee shall implement the modifications listed in Section 2.8, Table 2.8.1-1, "Committed Plant Modifications," in the NRC SE dated December 29, 2010.
| |
| Renewed License No. DPR-55 Amendment No. 372
| |
| | |
| - 10
| |
| : 3) The licensee shall maintain appropriate compensatory measures in place until completion of all modifications and implementation items delineated above.
| |
| E. Physical Protection Duke Energy Carolinas, LLC, 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 27817 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: "Duke Energy Physical Security Plan" submitted by letter dated September 8, 2004, and supplemented on September 30, 2004, October 15, 2004, October 21, 2004, and October 27, 2004.
| |
| F. In the update to the UFSAR required pursuant to 10 CFR 50.71 (e)(4) scheduled for July, 2001, the licensee shall update the UFSAR to include the UFSAR supplement submitted pursuant to 10 CFR 54.21(d) as revised on March 27, 2000. Until the UFSAR update is complete, the licensee may make changes to the programs described in its UFSAR 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. The licensee's UFSAR supplement submitted pursuant to 10 CFR 54.21 (d),
| |
| as revised on March 27, 2000, describes certain future inspection activities to be completed before the period of extended operation. The licensee shall complete these activities no later than February 6, 2013.
| |
| H. Mitigation Strategy License Condition 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 Renewed License No. DPR-55 Amendment No. 372 I
| |
| | |
| - 11 (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. SFP mitigation measures (c) Actions to minimize release to include consideration of:
| |
| : 1. Water spray scrubbing
| |
| : 2. Dose to onsite responders
| |
| : 4. This renewed license is effective as of the date of issuance and shall expire at midnight on July 19, 2034.
| |
| FOR THE NUCLEAR REGULATORY COMMISSION Original Signed By Roy P. Zimmerman Roy P. Zimmerman, Acting Director Office of Nuclear Reactor Regulation
| |
| | |
| ==Attachment:==
| |
| : 1) Appendix A - Technical Specifications Renewed License No. DPR-55 Date of issuance: May 23, 2000 Renewed License No. DPR-55 Amendment No. 372
| |
| | |
| 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:
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| : a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978;
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| : 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;
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| : c. Quality assurance for effluent and environmental monitoring; and
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| : d. All programs specified in Specification 5.5.
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| OCONEE UNITS 1, 2, & 3 5.0-6 Amendment Nos. 371, 373, 372
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| O~~ICIAL USE ONLY SECURITY RELATED IN~ORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION TRANSITION TO A RISK-INFORMED. PERFORMANCE-BASED FIRE PROTECTION PROGRAM IN ACCORDANCE WITH 10 CFR 50.48(c)
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| RELATED TO AMENDMENT NO. 371 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-38 AMENDMENT NO. 373 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-47 AND AMENDMENT NO. 372 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-55 DUKE ENERGY CAROLINAS, LLC OCONEE NUCLEAR STATION, UNITS 1. 2, AND 3 DOCKET NOS. 50-269, 50-270. AND 50-287 Security-related information pursuant to Title 10 of Code of Regulations (10 CFR), Section 2.390 has been redacted from this document.
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| Redacted information is identified by blank space enclosed within double brackets as shown here [[ ]].
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| ENCLOSURE 4 O~~ICIAL USE ONLY SECURITY RELATED IN~ORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION TABLE OF CONTENTS TABLE OF CONTENTS i ACRONyMS iv
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| ==1.0 INTRODUCTION==
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| 1 1.1. Background 1 1.2 Requested Licensing Action 1
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| ==2.0 REGULATORY EVALUATION==
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| 2 2.1. Applicable Regulations 4 2.2. Applicable Staff Guidance 5 2.3. Interim Staff Positions (NFPA 805 Frequently Asked Questions Process) 6 2.4. Orders, License Conditions and Technical Specifications 7 2.4.1. Orders 7 2.4.2. License Conditions 7 2.4.3. Technical Specifications 8 2.5. Updated Final Safety Analysis Report (UFSAR) 9 2.6. Rescission of Exemptions 9 2.7. Self Approval Process for Post-Transition FPP Changes 10 2.7.1. Self Approval Using the Plant Change Evaluation Process 11 2.7.2. Self Approval of Changes to NFPA 805, Chapter 3, Requirements 13 2.8. Implementation 14 2.8.1. Modifications 14 2.8.2. Schedule 18 2.9. Summary of Implementation Items 18
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| ==3.0 TECHNICAL EVALUATION==
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| 25 3.1. NFPA 805 Fundamental FPP Elements and Minimum Design Requirements 26 3.1.1. Compliance with NFPA 805, Chapter 3, Requirements 27 3.1.2. Identification of Power Block 30 3.1.3. Performance-Based Methods for NFPA 805, Chapter 3, Elements 30 3.1.3.1 Use of Non-treated Wood 31 3.1.3.2 Use of Compressed Flammable Gas Storage in the Power Block 33 3.1.3.3 Use of Non-listed / Unapproved Wiring above the Suspended Ceiling 35 3.1.3.5 Allow Reactor Coolant Pump (RCP) Oil Mist Without Collection 39 3.1.3.6 Fire Pump Circulation Relief Valves 41 3.1.3.7 Insufficient Pressure for Reactor Building Hose Stations 42 3.1.3.8 Fire Pump Automatic/Remote Stop Feature .45 3.1.3.9 KHS Fire Main and Standpipe Use 46 OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE Of:l.JLY SECURITY RELATED INFORMATION ii 3.1.3.10 Use of Dual Purpose Fire Protection Water Supplies 48 3.1.3.11 KHS Fire Protection Fire Pump 50 3.1.4. Conclusion for Section 3.1 52 3.2. Nuclear Safety Capability Assessment (NSCA) Methods 52 3.2.1. Compliance with Nuclear Safety Capability Assessment Methods 54 3.2.2. Applicability of Feed and Bleed 63 3.2.3. Assessment of Multiple Spurious Operations (MSOs) 64 3.2.4. Transition of Operator Manual Actions to Recovery Actions 65 3.2.5. Conclusion for Section 3.2 68 3.3. Fire Modeling 68 3.4. Fire Risk Assessments 69 3.4.1. Maintaining Defense-in-Depth and Safety Margins 70 3.4.2. Fire Risk Evaluation 72 3.4.3. Quality of the Probabilistic Risk Assessment ~ 73 3.4.4. Additional Risk Presented by Recovery Actions 79 3.4.5. Risk-Informed or Performance-Based Alternatives to NFPA 805 80 3.4.6. Cumulative Risk and Combined Changes 81 3.4.7. Conclusion for Section 3.4 84 3.5. Nuclear Safety Capability Assessment Results 85 3.5.1. Nuclear Safety Capability Assessment Results by Fire Area 86 3.5.2. Fire Protection During Non-Power Operational Modes 93 3.5.3. Conclusion for Section 3.5 99 3.6. Radioactive Release Performance Criteria 101 3.7. Monitoring Program 103 3.8. Program Documentation, Configuration Control, and Quality Assurance 107 3.8.1. Documentation 107 3.8.2. Configuration Control 107 3.8.3. Quality 108 3.8.4. Fire Protection Quality Assurance Program 110 3.8.5. Conclusion for Section 3.8 111 4.0 FIRE PROTECTION LICENSE CONDITION 111 5.0
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| ==SUMMARY==
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| 112
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| ==6.0 STATE CONSULTATION==
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| 112
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| ==7.0 ENVIRONMENTAL CONSIDERATION==
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| 112
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| ==8.0 CONCLUSION==
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| 112
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| ==9.0 REFERENCES==
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| 113 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix 120 OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION iii Attachment B, Nuclear Safety Capability Assessment Method Review 156 Attachment C, Fire Risk Evaluation Tables 180 Attachment D, Nuclear Safety Capability Assessment Results by Fire Area 225 Attachment E, Radioactive Release Tables 308 OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION iv ACRONYMS AB auxiliary building AC alternating current ADAMS Agencywide Documents Access and Management System AFW auxiliary feedwater system AHJ authority having jurisdiction AHU air-handling unit ANS American Nuclear Society AOPs abnormal operating procedures AOV air-operated valve ARP auxiliary relay panel AS accident sequence ASD alternative shutdown ASME American Society of Mechanical Engineers ASTM American Society for Testing and Materials ASW auxiliary service water ATWS anticipated transient without scram Aux auxiliary AWT activity waste tank BH block house BHUT bleed holdup tank Bid bleed B&WOG Babock & Wilcox Owners Group BTP Branch Technical Position BWR boiling-water reactor BWST borated water storage tank CAP corrective action program CBAST concentrated boric acid storage tank CCs capability categories CCDP conditional core damage probability CCF common cause failure CCW condenser circulating water CCWS component cooling water system CDF core damage frequency CFAST Consolidate Model of Fire Growth and Smoke Transport CFR Code of Federal Regulations Chem Chemistry Clrs coolers Cmp component CRD control rod drive CSA Canadian Standards Association CSD cold shutdown CSIP charging/safety injection pump CRS Control Room Supervisor CT(s) current transformers DBD design basis document DBS design basis specification DC direct current DECON decontamination Demin demineralizer OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION v
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| DG diesel generator DHR decay heat removal DID defense-in-depth EC engineering change EEE engineering equivalency evaluation EEEEs existing engineering equivalency evaluation EFW emergency feedwater EHC electro-hydraulic control EOP emergency operating procedures EPRI Electric Power Research Institute ERFBS electrical raceway fire barrier system ERFIS emergency response facility information system ES engineered safeguards ESFAS engineered safety features actuation system ESV essential siphon vacuum Evp evaporator EWST elevated water storage tank F&O(s) facts and observations F&O(s) findings and observations FACP fire alarm control panel FAQ(s) frequently asked questions Fltr filter FPE fire protection engineer FPIE full power internal event FPDID fire protection defense-in-depth FPP fire protection program FPRA fire probabilistic risk assessment FR Federal Register FRE fire risk evaluation FSA fire safety analyses FSAR Final Safety Analysis Report GDC General Design Criterion GL U.S. NRC Generic Letter GPM gallons per minute HEAF high energy arcing faults HEP human error probabilities HFEs human failure events HGL hot gas layer HPI high-pressure injection HPSW high-pressure service water HR human reliability HRA human reliability analysis HRE high(er) risk evolution(s)
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| HRR heat release rates HSS high safety significant HUT holdup tank HVAC heating ventilation air conditioning HWP hotwell pump I&E instrument and electronic IEEE Institute of Electrical and Electronic Engineers OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION vi IMC NRC Inspection Manual Chapter IN U.S. NRC Information Notice ISFSI Independent Spent Fuel Storage Installation ISLOCA interfacing system loss-of-coolant accident KEO Keowee KHS Keowee Hydro Station KSF(s) key safety function(s)
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| LAR license amendment request LD letdown LDST letdown storage tank LERF large early release frequency LlH low and high LPI low-pressure injection LPIP low-pressure injection pump LPSW low-pressure service water MCB main control board MCC motor control center MCR main control room MDEFDW motor-driven emergency feedwater MFW main feedwater MOV(s) motor-operated valves MR modification required MREM milia roentgen MS moisture separator MSH(s) main steam headers MSO(s) multiple spurious operation(s)
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| MSRH moisture separator reheater MSV milia sievert MT main turbine MTOT main turbine oil tank MWHT main waste holdup tank NEI Nuclear Energy Institute NEIL Nuclear Electric Insurance Limited NFPA National Fire Protection Association NPO non-power operation NRC U.S. Nuclear Regulatory Commission NSCA nuclear safety capability assessment NSD Nuclear System Directive NPSH net positive suction head NUREG documents prepared by the NRC staff OMA(s) operator manual action(s)
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| ONS Oconee Nuclear Station, Units 1, 2, and 3 OSC Oconee site calculation OSFD ONS simplified flow diagram OS&Y outside screw and yoke PALS post-accident liquid sample PB performance-based PCS primary control station PDS plant damage states PIC process instrumentation cabinet OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION vii P&ID piping and instrument drawings PIP problem identification program PMG performance monitoring group Pmp pump POS(s) plant operating state(s)
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| PORV(s) power-operated relief valve(s)
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| PRA probabilistic risk assessment PSA probabilistic safety assessment PSI pounds per square inch PSW protected service water PVC poly-vinyl chloride PWR pressurized-water reactor PWROG PWR Owners Group QA quality assurance RA(s) recovery action(s)
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| RAB reactor auxiliary building RAI request for additional information RAW risk achievement worth RB reactor building RBS reactor building spray RBES reactor building emergency sump RBSP reactor building spray pump RC reactor coolant RCA radiation-controlled area RCMU reactor coolant makeup RCP reactor coolant pump RCS reactor coolant system Res resin RG regulatory guide RHR residual heat removal RI risk-informed RIAs radiation indicating alarms RI/PB risk-informed, performance-based RI/PB FPP risk-informed, performance-based fire protection program RIS Regulatory Information Summary RMA radioactive material area RP radiation protection RWST refueling water storage tank SC success criteria SO Site Directive SDQA software and data quality assurance SE safety evaluation SFP spent fuel pool SFPE Society of Fire Protection Engineers SG steam generator SGTR steam generator tube rupture SI safety injection SLCs selected licensee commitments SM safety margin SOG Standard Operating Guide OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION viii Spt spent SRs supporting requirements SRP standard review plan SRST spent resin storage tank SSC structures, systems, and components SSD safe shutdown SSE SSD earthquake SSEL SSD equipment list SSF standby shutdown facility SSHR secondary side heat removal SSPS solid state protection system SWGR switchgear SY systems SYD switchyard SW service water TB turbine building TDEFDW turbine-driven emergency feedwater TEDE total effective dose equivalent TH thermal-hydraulic TP thermoplastic TR Technical Report TS(s) technical specification(s)
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| UFSAR Updated Final Safety Analysis Report UL Underwriters Laboratory V volt V&V verification and validation VAC volts alternating current VCT volume control tank VEWFDS very early warning fire detection system VFDR variation from deterministic requirements Iyr per year Xfer transfer ZOI zone of influence OFFICIAL USE ONLY SECURITY REU\TED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 1
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| ==1.0 INTRODUCTION==
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| 1.1 Background On June 16, 2004, the U.S. Nuclear Regulatory Commission (NRC or the Commission) revised its regulation Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.48 to include a new paragraph 50.48(c). The new paragraph incorporates by reference National Fire Protection Association Standard 805 (NFPA 805), "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants 2001 Edition," (Reference 1) hereafter referred to as NFPA 805. This change to the NRC's fire protection regulations provides licensees with the opportunity to adopt a performance-based (PB) fire protection program (FPP) as an alternative to the existing, deterministic fire protection regulations.
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| Specifically, NFPA 805 allows the use of PB methods, such as fire modeling, and risk-informed (RI) methods, such as fire probabilistic risk assessment (PRA), to demonstrate compliance with the nuclear safety performance criteria.
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| In the related license amendment request (LAR) and this safety evaluation (SE), extensive reference is made to NFPA 805. In particular, when this SE refers to an FPP element as being in compliance with, or meeting the requirements of, NFPA 805, the NRC staff intends this to indicate that the element is in compliance with 10 CFR 50.48(c) as well as the applicable portions of NFPA 805.
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| 1.2 Requested Licensing Action By application dated May 30, 2008 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML081650475) (Reference 2), as supplemented by letters dated June 30, 2008 (ADAMS Accession No. ML081890193), April 21 ,2008 (ADAMS Accession No.
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| ML091170546), February 9,2009 (ADAMS Accession No. ML090480143), February 23,2009 (ADAMS Accession No. ML090700134), May 31,2009 (ADAMS Accession No. ML091590045),
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| August 3,2009 (ADAMS Accession No. ML092190212), September 29,2009 (ADAMS Accession No. ML092740624), and November 30,2009 (ADAMS Accession No. ML093410007)
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| (References 3, 4, 5, 6, 7, 8, 9, and 10, respectively), Duke Energy Carolinas, LLC (Duke, the licensee), requested a license amendment to allow the licensee to maintain a FPP in accordance with 10 CFR 50.48(c) for the Oconee Nuclear Station, Units 1, 2, and 3 (ONS).
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| By letter dated April 14, 2010 (ADAMS Accession No. ML101121042 (Reference 11), the licensee resubmitted the LAR and superseded the contents of the LAR submitted by letters dated May 30,2008, and October 31,2008. This resubmitted LAR, however, does not .
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| supersede the supplements dated January 30, 2009 (ADAMS Accession No. ML091040205),
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| February 9, 2009, February 23, 2009, May 31, 2009, August 3, 2009, September 29, 2009, and November 30, 2009. By letters dated September 13, 2010 (ADAMS Accession No.
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| ML102640110) (Reference 12), September 27, 2010 (ADAMS Accession No. ML102720409)
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| (Reference 13), October 14,2010 (ADAMS Accession No. ML102910093) (Reference 54),
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| November 19, 2010 (ADAMS Accession No. ML103300227) (Reference 52), and December 22, 2010 (ADAMS Accession No. ML103620105) (Reference 59), the licensee supplemented the LAR on April 14, 2010.
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| Pursuant to Section 2.390 of Title 10 of the Code of Federal Regulations (10 CFR), by letter dated December 6, 2010, the NRC sent the licensee the draft Safety Evaluation and provided the licensee with an opportunity to comment on any proprietary or security-related aspects of OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 2
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| the draft SE. By letter dated December 22,2010, the licensee provided comments. The NRC reviewed and accepted all comments made by the licensee.
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| In addition, the December 6, 2010, letter also requested the licensee to provide comments on factual errors or clarity concerns contained in the draft SE. By letter dated December 22, 2010, the licensee provided comments. The NRC has considered each comment and changed the Safety Evaluation as appropriate.
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| The licensee is requesting amendments to the ONS renewed operating facility licenses and technical specifications (TSs) to establish and maintain a PB FPP in accordance with the requirements of 10 CFR 50.48(c). Specifically, the licensee requests to transition from the existing deterministic fire protection licensing basis established in accordance with 10 CFR 50.48(b) and 10 CFR Appendix R to a 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 ONS is referred to as risk-informed, performance-based (RI/PB) FPP throughout this SE.
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| The licensee has proposed a new fire protection license condition reflecting the new RI/PB FPP licensing basis, as well as revisions to the TSs that address this change to the current FPP licensing basis. Section 2.4.2 and Section 4.0 of this SE discuss in detail the license condition, and Section 2.4.3 discusses the TS changes.
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| By letter dated April 14, 2010, (Reference 11), the licensee stated in their Section 4.2.3 that safe shutdown (SSD) requirements for fire protection, turbine building (TB) flooding, and physical security requirements were resolved by NRC approval of the station standby shutdown facility (SSF) design in an SE dated April 28, 1983 (ADAMS Accession No. ML103370444) (Reference 24). The fire protection portions of the approval have been incorporated into the Nuclear Safety Capability Assessment (NSCA) and so Reference 24 is no longer applicable to the ONS FPP.
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| ==2.0 REGULATORY EVALUATION==
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| Section 50.48, "Fire protection," of 10 CFR provides the NRC requirements for nuclear power plant fire protection. Paragraph 50.48(c) of 10 CFR outlines the NRC requirements applicable to licensees that choose to adopt a PB FPP as an alternative to meeting the requirements of 10 CFR 50.48(b) for plants licensed to operate before January 1, 1979, or the approved fire protection license conditions for plants licensed to operate after January 1, 1979. ONS Units 1, 2 and 3 received -their operating licenses prior to January 1, 1979.
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| The NRC regulations include specific procedural requirements for implementing an RI/PB FPP based on the provisions of NFPA 805. In particular, 10 CFR 50.48(c)(3)(i) requires licensees which choose to adopt an RI/PB FPP in compliance with NFPA 805 to submit an LAR to the NRC that identifies any orders and license conditions that must be revised or superseded, and contains any necessary revisions to the plant's TSs and the bases thereof. The license conditions issued with these amendments will supersede the current fire protection license condition with a condition that allows implementation of an FPP in accordance with NFPA 805.
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| 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 FPP or nuclear power plant as permitted by NFPA 805."
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELJ\TED INFORMATION 3
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| The intent of this paragraph is given in the statement of considerations for the final rule, which was published in the Federal Register on June 16, 2004 (69 FR 33536). The statement of considerations states:
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| This paragraph requires licensees to complete all of the Chapter 2 methodology (including evaluations and analyses) and to modify their fire protection plan before making changes to the fire protection program or to the plant configuration. This process ensures that the transition to an NFPA 805 configuration is conducted in a complete, controlled, integrated, and organized manner. This requirement also precludes licensees from implementing NFPA 805 on a partial or selective basis (e.g., in some fire areas and not others, or truncating the methodology within a given fire area).
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| The evaluations and analyses process in Chapter 2 of NFPA 805 provides for the establishment of the fundamental fire protection program, identification of fire area boundaries and fire hazards, determination by analysis that the plant design satisfies the performance criteria, identification 'of the structures, systems and components (SSCs) required to achieve the performance criteria, conduct of plant change evaluations, establishment of a monitoring program, development of documentation, and configuration control. Chapter 2 of NFPA 805 also provides for the use of a deterministic or performance-based approach to determine that the performance criteria are satisfied and provides for the use of tools such as engineering analyses, fire models, nuclear safety capability assessments, and fire risk evaluations to support development of these approaches. The methodology for the use of these tools is established in Chapter 4 of NFPA 805 (69 FR 33548).
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| In the LAR, the licensee has provided a description of the revised FPP 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 staffs evaluation of the licensee's amendment request and concludes that:
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| (1) The licensee has 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). The NRC staff finds this adequate.
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| (2) The licensee has completed its implementation of the methodology in Chapter 2, "Methodology," of NFPA 805, including completion of all the required evaluations and analyses outlined by the statement of considerations, and the NRC staff has approved the licensee's modified FPP, which reflects the decision to comply with NFPA 805, consistent with 10 CFR 50.48(c)(3)(ii).
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| Since items (1) and (2) satisfy the requirements of 10 CFR 50.48(c)(3), the NRC staff concludes that the licensee's implementation of the modified FPP that aligns with NFPA 805, including physical plant modifications as described in the LAR and supplements, in accordance with the implementation schedule set forth in this SE and the accompanying license condition, is sufficient to demonstrate compliance with 10 CFR 50.48(c).
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| The regulations also allow for flexibility that was not originally included in the NFPA 805 standard. Licensees that choose to adopt 10 CFR 50.48(c), but wish to use the PB methods permitted elsewhere in the standard to meet the fire protection requirements of NFPA 805, OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 4
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| Chapter 3, "Fundamental Fire Protection Program and Design Elements," may do so by submitting an LAR in accordance with 10 CFR 50.48(c)(2)(vii). 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).
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| In addition to the conditions outlined by the rule that require licensees to submit an LAR for NRC review and approval in order to adopt an RI/PB FPP, licensees may also submit additional elements of their FPP for which they wish to receive specific NRC review and approval, as set forth in Regulatory Guide (RG) 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," Revision 1, Regulatory Position C.2.2.1, published in the Federal Register on December 18,2009 (74 FR 67253; Reference 14). 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. However, any submittal addressing these additional FPP elements should include sufficient detail to allow the NRC staff to assess whether the licensee's treatment of these elements meets the 10 CFR 50.48(c) requirements.
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| The purpose of the FPP established by NFPA 805 is to provide assurance, through a defense in-depth (DID) philosophy, that the fire protection objectives are satisfied.
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| NFPA 805, Section 1.2, "Defense-in-Depth," states the following:
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| 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:
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| (1) Preventing fires from starting (2) Rapidly detecting and controlling and extinguishing promptly those fires that do occur, thereby limiting fire damage (3) Providing an adequate level of fire protection for structures, systems, and components important to safety, so that a fire that is not promptly extinguished will not prevent essential plant safety functions from being performed In addition, in accordance with General Design Criterion (GDC) 3, "Fire protection," of Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, fire protection systems must be designed such that their failure or inadvertent operation does not adversely impact the ability of the SSCs important to safety to perform their intended safety functions.
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| 2.1. Applicable Regulations The licensee's FPP will generally be considered acceptable if it meets the applicable regulatory criteria established by the following regulations:
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| * 10 CFR Part 50, Appendix A, GDC 3, "Fire protection," establishes the general criteria for fire and explosion protection of SSCs important to safety.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 5
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| * 10 CFR Part 50, Appendix A, GDC 5, "Sharing of Systems, Structures, and Components," relates to shared fire protection systems and potential fire impacts on shared SSCs important to safety.
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| * 10 CFR 50.48(a), requires that each operating nuclear power plant have a fire protection plan that meets the requirements of GDC 3.
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| * 10 CFR 50.48(c), incorporates NFPA 805 (2001 Edition) by reference, with certain exceptions, modifications, and supplementation. This regulation establishes the requirements for using a PB FPP in conformance with NFPA 805 as an alternative to the requirements associated with 10 CFR 50.48(b) and Appendix R, "Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979," to 10 CFR Part 50, or the specific plant license condition(s) related to fire protection.
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| Because NFPA 805 was incorporated by reference into 10 CFR, all requirements of the endorsed standard must be met, unless an exemption is granted by the NRC as allowed in 10 CFR 50.12, "Specific Exemptions."
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| * 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."
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| 2.2. Applicable Staff Guidance The NRC staff's review also relied on the following additional codes, RGs, and standards:
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| * RG 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," Revision 1, issued December 2009 (ADAMS Accession No.
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| ML092730314), (Reference 14), which provides guidance to licensees for implementing an RI/PB FPP in compliance with 10 CFR 50.48(c).
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| * RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 1, issued November 2002 (ADAMS Accession No. ML023240437), (Reference 15), which provides guidance to licensees on acceptability limits for RI changes to the licensing basis.
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| * RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, issued March 2009 (ADAMS Accession No. ML090410014), (Reference 16), which provides guidance to licensees on methods for determining the technical adequacy of probabilistic risk assessment (PRA) results when used for RI changes to the licensing basis.
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| * RG 1.189, "Fire Protection for Operating Nuclear Power Plants," Revision 2, issued October 2009 (ADAMS Accession No. ML092580550), (Reference 17), which provides guidance to licensees on the proper content and quality of engineering equivalency evaluations used to support the FPP.
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| * NUREG-0800, Section 9.5.1.2, "Risk-Informed, Performance-Based Fire Protection Program," Revision 0, issued December 2009 (ADAMS Accession No. ML092590527),
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| (Reference 18), which provides the NRC staff with guidance for evaluating LARs that seek to implement a PB FPP in accordance with 10 CFR 50.48(c).
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 6
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| * NUREG-0800, Section 19.1, "Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, issued June 2007 (ADAMS Accession No. ML071700657), (Reference 19), which provides the NRC staff with guidance for evaluating the technical adequacy of a licensee's PRA results when used to request RI changes to the licensing basis.
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| * 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 (ADAMS Accession No. ML071700658), (Reference 20), which provides the NRC staff with guidance for evaluating the risk information used by a licensee to support permanent, RI changes to the licensing basis.
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| It should be noted that during the course of the review of the ONS NFPA 805 LAR, several of the above guidance documents were revised to incorporate updated information and lessons learned during the course of the transition process. As such, the original ONS NFPA 805 LAR was submitted against earlier revisions of some of these documents (e.g., RG 1.205). The revised LAR submitted on April 14, 2010 (Reference 11), incorporated into the application many of the positions in the new document revisions. Accordingly, the NRC staff considers that the NFPA 805 revised LAR meets the intent of the current document revisions, and was reviewed as such.
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| 2.3. Interim Staff Positions (NFPA 805 Frequently Asked Questions Process)
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| During the ongoing NFPA 805 pilot transition process, as well as throughout the subsequent non-pilot reviews, the NRC staff, industry, and other interested stakeholders expect to gain experience and develop lessons learned during the submission and subsequent review of each LAR to transition a licensee to an RI/PB FPP. The lessons learned are often converted into interim staff positions, which apply to the ongoing review until they can be formally incorporated into the NFPA 805 guidance documents such as Nuclear Energy Institute (NEI) document NEI 04-02, "Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c)" (ADAMS Accession No. ML081130188), (Reference 21), as endorsed, and RG 1.205.
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| The lessons learned and interim staff positions address the NRC's performance goals of maintaining safety, improving effectiveness and efficiency, reducing regulatory burden, and increasing public confidence. In most cases, the meetings and other interactions involved in promulgating interim staff positions are open to the public and feedback is welcomed. With respect to the NFPA 805 LARs, the NRC established the frequently asked questions (FAQs) process as described in Regulatory Information Summary (RIS) 2007-19, "Process for Communicating Clarifications of Staff Positions Provided in Regulatory Guide 1.205 Concerning Issues Identified during the Pilot Application of National Fire Protection Association Standard 805," (ADAMS Accession No. ML071590227), (Reference 22), to clarify issues encountered during the pilot transition process.
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| The FAQ process provides a means for the NRC staff to establish and communicate interim positions on technical and regulatory issues that emerge as experience is gained during review of the NFPA 805 LARs. Approved interim staff positions documented through the FAQ process are used where applicable in reviewing those portions of the LAR to which they apply.
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| The following table provides the current set of FAQs the NRC staff used in the preparation of this SE, as well as the SE section to which the FAQ was applied.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE O~JLY SECURITY RELATED INFORMATION 7
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| Table 2.3-1: NFPA 805 Frequently Asked Questions Closure SE Memo Section FAQ# Rev. FAQ Title ADAMS Accession Nos.
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| 06-0008 9 Fire Protection Enqineerinq Evaluations ML073380976 4.0 06-0022 3 Acceptable Electrical Cable Construction Tests ML091240278 3.1 07-0032 2 10 CFR 50.48(a) and GDC Clarification ML081400292 2.0 07-0039 2 Provide Update for NEI 04-02, Table B-2 ML091320068 3.2 07-0040 4 Non-Power Operations Clarification ML082200528 3.5 08-0048 0 NUREG/CR-6850 Revised Fire Ignition Frequencies ML092190457 3.5 (ADAMS Accession No. ML052580075) 2.4. Orders, License Conditions and Technical Specifications Paragraph 50.48(c)(3)(i) of 10 CFR Part 50 states that the LAR "must identify any orders and license conditions that must be revised or superseded, and contain any necessary revisions to the plant's technical specifications and the bases thereof."
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| 2.4.1. Orders The NRC staff reviewed Section 5.2.3, "Orders and Exemptions," and Attachment 0, "Orders and Exemptions," of ONS's NFPA 805 License Amendment Request Transition Report, as revised on April 14, 2010 (Reference 11), hereafter referred to simply as the LAR, with regard to NRC-issued Orders pertinent to ONS that are being revised or superseded by the NFPA 805 transition process. The licensee determined that no Orders need to be superseded or revised to implement an FPP at ONS that complies with 10 CFR 50.48(c).
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| This review, conducted by the licensee, included an assessment of docketed correspondence files and electronic searches, including internal ONS records and ADAMS. The review was performed to ensure that compliance with the physical protection requirements, security orders, and adherence to commitments applicable to ONS are maintained. The NRC staff accepts the licensee's determination that no Orders need to be superseded or revised to implement NFPA 805 at ONS.
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| In addition, a specific review was performed of the license amendment that incorporated the mitigation strategies required by Section B.5.b of Commission Order EA-02-026 (ADAMS Accession No. ML072260290), (Reference 23) to ensure that any changes being made in order to comply with 10 CFR 50.48(c) do not invalidate existing commitments applicable to ONS. The licensee's review of this Order and the related license amendment demonstrated that changes to the FPP during transition to NFPA 805 will not affect the mitigation measures required by Section B.5.b.
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| 2.4.2. License Conditions The NRC staff reviewed LAR Section 5.2.1, "License Condition Changes," and Attachment M, "License Condition Changes," regarding changes the licensee is seeking to make to the ONS fire protection license condition in order to adopt NFPA 805, as required by 10 CFR 50.48(c)(3).
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICI,J\L USE ONLY SECURITY RELATED INFORMATION 8
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| The NRC staff reviewed the revised license condition the licensee requested, which supersedes the current ONS fire protection license condition 3.0, for consistency with the format and content guidance outlined by Regulatory Position C.3.1 of RG 1.205, Revision 1. This section of RG 1.205 outlines an approach acceptable to the NRC staff for promulgating a fire protection license condition in accordance with the requirements of NFPA 805. Overall, the licensee's replacement license condition conforms to the guidance in RG 1.205, Revision 1.
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| Furthermore, the revised license condition, as specified by the sample license condition, identifies the plant-specific modifications outlined in the LAR, and associated implementation schedules, which must be accomplished at ONS to complete the transition to NFPA 805. In addition, the revised license condition includes a requirement that appropriate compensatory measures will remain in place until implementation of the specified plant modifications are completed. The modifications, implementation schedules, and compensatory measures ensure that completion of the transition to NFPA 805 at ONS will be orderly and conducted in accordance with the applicable regulations and license conditions.
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| Once these and other implementation issues are completed, NFPA 805 will be fully in effect at ONS, and provided that the licensee implements the RI/PB FPP as described in the LAR, as supplemented, the licensee will be in full compliance with 10 CFR 50.48(c). These modifications and implementation schedules are identical to those identified in the LAR, as discussed in SE Sections 2.8.1 and 2.8.2, and explicitly reviewed in Section 3.0, of this SE.
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| The licensee's proposed license condition is consistent with the content and format of the sample license condition in RG 1.205, Revision 1. Section 4.0 of this SE discusses the proposed ONS FPP license condition.
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| 2.4.3. Technical Specifications The NRC staff reviewed LAR Section 5.2.2, "Technical Specifications" and Attachment N, "Technical Specification Changes," with regard to proposed changes to the ONS TSs that are being revised or superseded during the NFPA 805 transition process. According to the LAR, the licensee conducted a review of the ONS TSs, including proposed TS changes that have been submitted to the NRC for approval, to determine which TS sections will be impacted by the transition to an RI/PB FPP based on 10 CFR 50.48(c) The licensee identified three changes.
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| The first change is to delete TS Section 5.4.1. TS 5.4.1 currently states that written procedures shall be established, implemented, and maintained covering activities that include FPP implementation. As discussed in the LAR, TS Section 5.4.1 is being deleted because, after completion of the transition to NFPA 805, the requirement for establishing, implementing, and maintaining fire protection procedures is contained in 10 CFR 50.48(c), as specifically outlined in Section 3.2.3, "Procedures," of NFPA 805. The licensee has stated that the RI/PB FPP at ONS complies with the requirements of NFPA 805 Section 3.2.3 (see Section 3.1.1 of this SE).
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| The second change is to revise the bases of ONS TS 3.10.1, "Standby Shutdown Facility (SSF)"
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| to delete reference to Appendix R of 10 CFR Part 50. The bases for TS 3.10.1 currently refer to "10 CFR 50 Appendix R fire" four different times. As discussed in the LAR, the bases for TS 3.10.1 are being changed since 10 CFR Part 50, Appendix R, is no longer an appropriate basis for the ONS FPP.
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| The third change is to revise the bases of ONS TS 3.10.2, "Standby Shutdown Facility (SSF)
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| Battery Cell Parameters," to delete reference to Appendix R of 10 CFR Part 50. The bases for OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 9
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| TS 3.10.2 currently refer to "a 10 CFR 50 Appendix R fire" two different times. As discussed in the LAR, the bases for TS 3.10.2 are being changed since 10 CFR Part 50, Appendix R, is no longer an appropriate basis for the ONS FPP.
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| 2.5. Updated Final Safety Analysis Report (UFSAR)
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| The NRC staff reviewed LAR Section 5.4 "Revision to the ONS UFSAR" and Attachment Q, "UFSAR Changes" with regard to the proposed changes to the UFSAR as a result of transitioning to NFPA 805. Attachment Q states that the ONS UFSAR will be revised in accordance with 10 CFR 50.71(e) after this SE is issued.
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| The ONS transition to NFPA 805 represents a complete change in the licensing basis for their FPP. The NRC staff performed a review in order to determine that the licensee's proposed UFSAR changes are consistent with the RI/PB FPP described in the LAR (see below). The licensee's proposed changes to the UFSAR impact Section 9.5.1, "Fire Protection." Attachment Q provides an outline of the major sections and anticipated content of UFSAR Section 9.5.1 when it is revised. The major sections include:
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| * Section 9.5.1.1, "Design Basis Summary," will contain a general discussion of compliance with NFPA 805, Chapter 3, "Fundamental Fire Protection Program and Design Elements," a general discussion on NFPA 805, Chapter 4) "Performance Goal, Objectives and Criteria,"
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| (which includes discussions of nuclear safety for power and non-power conditions and a discussion of defense-in-depth), and a summary of radioactive release.
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| * Section 9.5.1.2, "Systems Description," will include a definition of power block structures, NSCA and equipment selection criteria, and required fire protection systems and features selection criteria.
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| * Section 9.5.1.3, "Safety Evaluation," will describe the methodology used to identify fire hazards, identify NSCA compliance strategies at power and non-power conditions on a fire area basis, demonstrate compliance with radioactive release criteria, summarize Fire PRA results, and summarize conclusions regarding compliance with NFPA 805.
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| * Section 9.5.1.4, "Inspection and Testing," will include information on inspection, testing, and surveillance methodologies and the monitoring program methodology.
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| * Section 9.5.1.5, "Personnel Qualification and Training," will include information on qualification and training of FPP personnel and the fire brigade.
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| 2.6. Exemptions The NRC staff reviewed LAR Section 5.2.3, "Orders and Exemptions," Attachment 0, "Orders and Exemptions," and Attachment K, "Existing Licensing Action Transition" with regard to previously-approved exemptions to Appendix R to 10 CFR Part 50, which the transition to a FPP licensing basis in conformance with NFPA 805 will supersede. The licensee requested and received NRC approval for exemptions from 10 CFR Part 50, Appendix R. The licensee identified the following eight exemptions to 10 CFR Part 50, Appendix R, that are being superseded by the ONS FPP that complies with 10 CFR 50.48(c):
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 10
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| : 1. Auxiliary Building (AB) Lack of 3-hour fire rated Barrier (ADAMS Accession No.
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| ML012000058), (Reference 27). This is an exemption from 10 CFR Part 50, Appendix R, Section III.G.2.a for the lack of 3-hour rated barrier separation between SSD circuits between the West Penetration Room Fire Areas and the Balance of Plant Fire Area.
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| : 2. AB Lack of 3-hour fire rated penetration seals (Reference 27). This is an exemption from 10 CFR Part 50, Appendix R, Section III.G.2.a for the lack of 3-hour fire rated barrier pipe penetrations separation between SSD circuits between the West Penetration Room Fire Areas and the Balance of Plant Fire Area.
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| : 3. AB Non-rated Expansion Joints (Reference 27). This is an exemption from 10 CFR Part 50, Appendix R, Section III.G.2.a for the lack of 3-hour fire rated cork in the expansion joint at the ceiling between the West Penetration Room Fire Areas and the Balance of Plant Fire Area.
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| : 4. Lack of Control Room Suppression (ADAMS Accession No. ML011990218), (Reference 38). This is an exemption from 10 CFR Part 50, Appendix R, Section III.G.3 for the lack of fixed suppression in the Control Rooms.
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| : 5. Outside and SSF Emergency Lighting (ADAMS Accession No. ML011990375),
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| (Reference 39). This is an exemption from 10 CFR Part 50, Appendix R, Section III.J for the lack of 8-hour emergency lighting.
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| : 6. Reactor Building (RB) 20 feet Separation without Intervening Combustibles (Reference 27). This is an exemption from 10 CFR Part 50, Appendix R, Section III.G.2.d for the lack of 20 feet horizontal distance separation between SSD circuits in the RB with no intervening combustibles.
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| : 7. RB Unrated Containment Mechanical Penetrations (Reference 27). This is an exemption from 10 CFR Part 50, Appendix R, Section III.G.2.a for the lack of 3-hour fire rated barrier pipe penetrations separation between the West Penetration Fire Areas and the RB Fire Areas.
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| : 8. SSF Lack of Instrumentation per II I. L.2 (ADAMS Accession No. ML091310038),
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| (Reference 40). This is an exemption from 10 CFR Part 50, Appendix R, Section 1I1.L.2 for the lack of a source range flux monitor and steam generator (SG) pressure indication at the SSF.
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| The NRC staff individually addresses the applicability and continuing validity of these exemptions as incorporated into the NFPA 805 FPP as part of the staff's review of the appropriate section or fire area involved (see SE Sections 3.2 and 3.3).
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| 2.7. Self Approval Process for Post-Transition Fire Protection Program Changes Upon completion of the implementation of the PB FPP and issuance of the license conditions, changes to the approved FPP must be evaluated to ensure that they are acceptable.
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| NFPA 805, Section 2.2.9, "Plant Change Evaluation," states the following:
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| In the event of a change to a previously approved fire protection program element, a risk-informed plant change evaluation shall be performed and the results used as described in 2.4.4 to ensure that the public risk associated with OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 11 fire-induced nuclear fuel damage accidents is low and that adequate defense-in depth and safety margins are maintained.
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| NFPA 805, Section 2.4.4, "Plant Change Evaluation," states:
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| 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.
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| As stated in RG 1.205, Regulatory Position C.3.1, the NRC may allow licensees to implement certain changes without prior NRC review and approval. A plant change evaluation must be performed for changes to the previously approved FPP, as stated above. An exception is for changes to certain NFPA 805, Chapter 3, requirements; this is discussed in Section 2.7.2. The specific implementation guidance documents associated with NFPA 805 (NEI 04-02, Section 5.3, and RG 1.205, Regulatory Position C.3.2) address the screening process and other requirements necessary to allow self approval of plant changes with the potential to impact the RI/PB FPP. Changes that do not meet the acceptance criteria of the license condition may either be cancelled or the licensee may request a change to the FPP under 10 CFR 50.90.
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| 2.7.1. Self Approval Using the Plant Change Evaluation Process The NRC staff reviewed LAR Section 4.7.2, "Compliance with Configuration Control Requirements in Section 2.7.2 of NFPA 805," and LAR Attachment M, "License Condition Changes," for compliance with the NFPA 805 Plant Change Evaluation requirements.
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| The licensee will utilize a multi-step process for identifying and evaluating proposed changes to the plant that impact the FPP. The first step of the process is an initial review of the proposed plant change to determine if it has the potential to impact (change) the NFPA 805 FPP. This is accomplished through a series of questions/checklists contained in current ONS procedures.
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| Initial reviews that identify potential FPP changes are further reviewed by a team of qualified individuals having relevant experience (i.e., Fire Protection, SSD/NSCA, Fire PRA) to determine the specific FPP changes, if any. If FPP changes are determined to exist as a result of the proposed plant change, a plant change evaluation must be performed. If the plant change is determined to comply with NFPA 805, Chapter 3 and/or Section 4.2.3, then a deterministic approach can be used.
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| By letter dated September 13, 2010, (Reference 12), the licensee stated that the licensee's change evaluation process consists of the following four subtasks:
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| * Defining the Change
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| * Performing the Preliminary Risk Screening
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| * Performing the Risk Evaluation
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| * Evaluating the Acceptance Criteria The licensee's plant change evaluation process starts with defining the change or altered condition to be evaluated and a review of the baseline configuration as defined by the existing licensing basis (i.e., the approved NFPA 805 FPP element).
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| Once the change has been defined, along with its relationship to the deterministically compliant condition or previously approved FPP element, a preliminary risk screening is performed. The OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 12 licensee's preliminary risk screening process is modeled after the process provided in NEI 02 03, Revision. 0, "Guidance for Performing a Regulatory Review of Proposed Changes to the Approved Fire Protection Program," (ADAMS Accession No. ML031780500), (Reference 50),
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| which it expects to address most administrative changes (e.g., changes to the combustible control program, organizational changes, etc.). The staff notes that t\lEI 02-03, although discussed in NEI 04-02, Revision 2, was not endorsed in RG 1.205, Revision 1 (i.e., it was not reviewed and endorsed by the NRC since it was only a "referenced document"). The NRC staff reviewed the licensee's preliminary risk screening process and determined that it meets the "assessment of the acceptability of risk" requirement of Section 2.4.4 of NFPA 805.
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| If the change to be evaluated does not screen out during the preliminary risk screening, the licensee's process allows a more detailed risk evaluation to be performed. These detailed evaluations may include fire modeling and risk assessment techniques. By letter dated August 3,2009 (ADAMS Accession No. ML092190212 (Reference 8), the licensee stated that post-transition (to NFPA 805) plant changes requiring a detailed risk evaluation will be evaluated using the Fire PRA. The licensee also stated that its process for ensuring configuration control of the Fire PRA model complies with the American Society of Mechanical Engineers (ASME) Standard for PRA quality and ensures that the licensee maintains an as built, as-operated PRA model of the plant. Section 3.4.3 of this SE discusses the technical adequacy of the licensee's Fire PRA, including the licensee's process to ensure that the Fire PRA remains current. In Section 3.4.3 of this SE, the NRC staff concludes that the licensee's PRA used to perform the risk assessments in accordance with NFPA 805 Section 2.4.4 (plant change evaluations) and Section 4.2.4.2 (fire risk evaluation) is of sufficient quality to support this application to transition to 10 CFR 50.48(c) because the remaining resolutions of findings on the internal events PRA and Fire PRA are not expected to change the substantial estimated risk decrease associated with this transition into a risk increase.
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| The proposed license condition as discussed in Section 4.0 of this SE does not allow the licensee to self-approve risk-informed changes to the FPP. The proposed license condition requires the licensee to submit a license amendment application (per 10 CFR 50.90) requesting such self-approval capability.
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| Based on the licensee's described process, the detailed risk evaluation will involve risk calculations for both CDF and LERF that will be used to model the proposed change and calculate the change in risk (i.e., LiCDF and LiLERF) with respect to the baseline configuration. Consistent with RG 1.205, Revision 1, Regulatory Position C.2.2.4.3, the post-transition baseline risk (used to evaluate cumulative risk impacts) is the risk of the plant at the point of full implementation of NFPA 805 (i.e., after completing all plant modifications and implementation items that the licensee has committed to make).
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| The final step in the plant change evaluation process involves determining whether the proposed change is acceptable with respect to risk, DID, and safety margin (SM), such that prior NRC review and approval is not required to implement the change. This step utilizes the guidance provided in NEI 04-02 and RG 1.205, Revision 1. As stated above, before achieving full compliance with 10 CFR 50.48(c) by implementing the plant modifications and implementation items listed in SE Sections 2.8 and 2.9 and subject to the NFPA 805 license condition and other license conditions (i.e., during full implementation of the transition to NFPA 805), RI changes to the licensee's FPP 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 using the initial review and/or preliminary risk screening process discussed above (i.e., use of the detailed risk evaluation is not approved at this time). In addition, the licensee is required to ensure that fire protection DID and SMs are maintained during the transition process. The OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 13 I'JFPA 805 license condition includes the appropriate limitations, acceptance criteria, and other attributes to form an acceptable method for meeting Regulatory Position C.3.1 of RG 1.205, Revision 1, with respect to the requirements for FPP changes during transition, and therefore demonstrate compliance with 10 CFR 50.48(c).
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| 2.7.2. Self Approval of Changes to NFPA 805, Chapter 3, Requirements The NFPA 805 license condition also includes a provision for self-approval of changes to the NFPA 805, Chapter 3 fundamental FPP elements and design requirements for which an engineering evaluation demonstrates that the alternative to the NFPA 805, Chapter 3, element is functionally equivalent or adequate for the hazard. These two types of engineering evaluations, discussed in detail below, are not plant change evaluations because they conclude that the change to the NFPA 805, Chapter 3, requirement still maintains the function of the NFPA 805 requirement.
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| The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3 element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement (i.e., has not impacted its contribution toward meeting the nuclear safety and radioactive release performance criteria), using a relevant technical requirement or standard.
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| These fire protection engineering evaluations can use qualitative analyses. The basis of approval for a functionally equivalent evaluation is that it maintains the function of the NFPA 805 requirement. As such, the determination that the condition is functionally equivalent means that the evaluated condition complies with the code requirement.
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| Use of this approach does not fall under NFPA 805, Section 1.7, "Equivalency," because the condition can be shown to meet the NFPA 805, Chapter 3, requirement. Section 1.7 of NFPA 805 is a standard format used throughout NFPA standards. It is intended to allow owner/operators to utilize the latest state-of-the-art fire protection features, systems, and equipment, provided the alternatives are of equal or superior quality, strength, fire resistance, durability, and safety. However, the intent is to require approval from the authority having jurisdiction (AHJ) for Section 1.7 type equivalencies 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.
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| Alternatively, 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 (with respect to the ability to meet the nuclear safety and radioactive release performance criteria), using a relevant technical requirement or standard.
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| The four specific sections of NFPA 805, Chapter 3, for which prior NRC review and approval are not required to implement alternatives that an engineering evaluation has demonstrated are adequate for the hazard are as follows:
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| * Fire Alarm and Detection Systems (Section 3.8)
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 14
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| * Automatic and Manual Water-Based Fire Suppression Systems (Section 3.9)
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| * Gaseous Fire Suppression Systems (Section 3.10)
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| * Passive Fire Protection Features (Section 3.11)
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| The engineering evaluations described above (i.e., functionally equivalent and adequate for the hazard) are engineering analyses governed by the NFPA 805 guidelines. In particular, this means that the evaluations must meet the requirements of NFPA 805, Section 2.4, "Engineering Analyses," and NFPA 805, Section 2.7, "Program Documentation, Configuration Control, and Quality." Specifically, the effectiveness of the fire protection features under review must be evaluated and found acceptable in relation to their ability to detect, control, suppress, and extinguish a fire and provide passive protection to achieve the performance criteria and not exceed the damage threshold for the plant being analyzed. The associated evaluations must also meet the documentation content (as outlined by NFPA 805, Section 2.7.1, "Content") and quality requirements (as outlined by NFPA 805, Section 2.7.3, "Quality") of the standard in order to be considered adequate. The NRC staff's review of the licensee's compliance with NFPA 805, Sections 2.7.1 and 2.7.3 is provided in SE Section 3.8.
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| 2.8. Implementation Regulatory Position C.3.1 of RG 1.205, Revision 1, provides guidance that the NFPA 805 license condition presented in the LAR should include the following: (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 commitment to maintain appropriate compensatory measures in place until implementation of the modifications is completed.
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| 2.8.1. Modifications The NRC staff reviewed LAR Attachment S, "Plant Modifications and Confirmatory Items," which describes the ONS plant modifications necessary to implement the NFPA 805 licensing basis as proposed. These modifications are identified in the LAR as necessary to bring ONS into compliance with either the deterministic or PB requirements of NFPA 805. LAR Table S-1 in Attachment S provides a description of each of the proposed plant modifications and presents the problem statement explaining why the modification is needed. This table also explains for each modification, as appropriate, that compensatory measures are currently in place for existing deficiencies associated with 10 CFR Part 50, Appendix R compliance, and that compensatory measures will be established when the NFPA 805 FPP becomes effective and will remain in effect until the modification is completed.
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| The NRC staff's review confirmed that the modifications identified in LAR Table S-1 are the same as those identified in LAR Table B-3, "Fire Area Transition," on a fire area basis, as the modifications being credited in the proposed NFPA 805 plant configuration and licensing basis.
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| The NRC staff also confirmed that the LAR Table S-1 modifications and associated implementation schedule are the same as those provided in the licensee's proposed NFPA 805 license condition (LAR Attachment N), and for which the licensee has committed to keep the appropriate compensatory measures in place until the modifications have been completed.
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| The plant modifications committed to in LAR Table S-1 must be completed in order for ONS to be in full compliance with 10 CFR 50.48(c) (NFPA 805). As discussed above, these modifications will be implemented in accordance with the schedule provided in the NFPA 805 license condition.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 15 In addition, the licensee has committed to keep the appropriate compensatory measures in place for each modification until the modification has been fully implemented. Table 2.8.1-1 presents a simplified version of LAR Table 8-1 and incorporates supplementary information provided by the licensee (Reference 59). The NRC will perform follow-up inspections to ensure that all items in Table 2.8.1 below have been completed prior to implementation of the license amendments.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 16 Table 2.8.1-1: Committed Plant Modifications Proposed Modification Item No. Problem Statement Modification Description Compensatory Completion Measures*
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| [[ II OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 17 Proposed Modification Item No. Problem Statement Modification Description Compensatory Completion Measures*
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 18 2.8.2. Schedule In the LAR Section 5.5 and as supplemented, the licensee provided the overall schedule for completing the NFPA 805 transition at ONS. The licensee stated that it would complete the implementation of the new program, including procedure changes, process updates, and training for affected plant personnel, within 24 months after NRC approval, as conveyed by the date of issuance of this SE.
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| LAR Attachment S provides an implementation completion schedule for each of the identified plant modifications. This implementation schedule is provided in Table 2.8.1-1, and in the proposed license condition. In addition, the proposed license condition includes a statement that appropriate compensatory measures will remain in place until implementation of these modifications is fully implemented (see Section 4.'0 of this SE).
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| 2.9. Summary of Implementation Items LAR Table S-2 in Attachment S provides a list of "Confirmatory Items" for ONS. These confirmatory items, referred to by the NRC as implementation items, are items that the licensee has not fully completed or implemented as of the issuance date of the SE, but which will be completed during implementation of the license amendment to transition to NFPA 805 (e.g.,
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| procedure changes that are still in process, NFPA 805 programs that have not been fully implemented, personnel training that is still underway, etc.). These items do not impact the bases for the safety conclusion made by the NRC staff in the associated SE.
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| The NRC staff, during a future fire protection inspection, may choose to examine the closure of the items, with the expectation that any variations discovered during this review, or concerns with regard to adequate completion of the items, would be tracked and dispositioned appropriately under the licensee's corrective action program.
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| As a result of its review of the ONS LAR, the NRC staff identified additional items that are contained in Table 2.9-1. For tracking purposes, the NRC staff has assigned a unique identifying number to each item. The table also specifies the associated section of the SE in which the item is identified, as well as the appropriate licensee document, which denotes that the action associated with the item is still ongoing, and provides some additional level of detail regarding what the change will entail.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 19 Table 2.9-1: Implementation Items Item SE Section Implementation Item Description LAR Section
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| : 1. Attachment A: The Design Basis Specification for Fire LAR Table S-2, Section 3.2.2.4, Protection will be updated to include the and Attachment A, Management Policy statement that the NRC is the AHJ for Subsection 3.2.2.4 onAHJ fire protection changes requiring approval.
| |
| : 2. Attachment A: Fleet Directive NSD-313, "Control of LAR Table S-2, and Section 3.3.1.2(2), Flammable and Combustible Materials," Attachment A, Control of Combustible will be updated to include the statement Subsection 3.3.1.2.2 Materials that plastic-sheeting materials shall conform to the requirements of NFPA 701 or equivalent.
| |
| : 3. Attachment A: Appropriate station procedure(s) for leak LAR Table S-2, and Section 3.3.1.3.3, or air flow testing will be updated to Attachment A, Control of Ignition Sources preclude the use of open flames or Subsection 3.3.1.3.3 for Leak TestinQ combustion Qenerated smoke.
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| : 4. Attachment A: Fleet Directive NSD-318 "Coatings LAR Table S-2, and Section 3.3.3, Program," will be updated to include the Attachment A Interior Finishes specifications for Class A walls/ceilings Subsection 3.3.3 and Class I floor finishes.
| |
| : 5. Attachment A: Appropriate station electrical LAR Table S-2, and Section 3.3.5.2, specifications will be updated to specify Attachment A Electrical Raceway only metal tray and metal conduits shall Subsection 3.3.5.2 Construction Limits be used for electrical raceways. Thin wall metallic tubing shall not be used for power, instrumentation, or control cables.
| |
| : 6. Attachment A: Transformer deluge system flow test LAR Table S-2, and Section 3.3.9, Transformers procedures will be updated to include Attachment A drainage inspections as part of the Subsection 3.3.9 annual flow tests.
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| : 7. Attachment A: Training and Station Fire Brigade Training LAR Table S-2, and Drills, Subsection documentation will be updated to Attachment A 3.4.3.(c)(3) include guidance to ensure fire drills are (Subsection 3.4.3.(c)(3>>
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| conducted in various plant areas, especially in those areas identified to be essential to plant operation and to contain siQnificant fire hazards.
| |
| : 8. Section 3.7 and Attachment Implement the monitoring program LAR Table S-2, and A: Section 3.2.3, described in SE Section 3.7. Attachment A Subsection 3.2.3.(3) Subsection 3.2.3.(3)
| |
| : 9. Attachment A: Pre-fire Plans will be updated to include LAR Table S-2, and Section 3.4.2.1 any changes to equipment important to Attachment A nuclear safety and other updates Subsection 3.4.2.1 pertinent to the NFPA 805 Transition.
| |
| : 10. Attachment A: Standard Operating Guidelines (SOGs) LAR Table S-2, and Section 3.4.2.3 will be updated to include a SOG with Attachment A the location of the Pre-Fire Plans. Subsection 3.4.2.3 OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 20
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| : 11. Attachment A: ONS code compliance calculation will LAR Table S-2, and Section 3.8.2, be updated to ensure required 'fire Attachment A detection devices are installed in Subsection 3.8.2 accordance with NFPA 72, 2007 Edition.
| |
| : 12. Attachment A: Validate hydraulics calculations for all LAR Table S-2, and Section 3.9.1 required automatic or manual water Attachment A based suppression systems. Subsection 3.9.1
| |
| : 13. Attachment D, The SSD procedure and analysis will be LAR Table S-2, Fire Area [[ ]] updated to incorporate the monitoring Attachment C, and/or adjustment of the following Fire Area [[ ]]
| |
| parameters required during operation of the SSF diesel generator (DG):
| |
| generator current, voltage, power and frequency. The controls and indications required to monitor and adjust these parameters are currently not included in the SSD analysis.
| |
| : 14. Section 3.2.4: Recovery Actions - Station procedures LAR Table S-2, Transition of will be updated to reflect new NSCA Attachment C, Operator Manual strategies (including supporting Fire Areas [[ ]],
| |
| Actions to Recovery communication coverage) and perform [[ ]], and [[ ]] and Actions training as necessary. The following Attachment G actions will be performed:
| |
| : 1) An evaluation to ensure that the hand-held radios operate in the locations of the recovery actions when needed, either with or without repeaters.
| |
| : 2) Development of SSD procedures for
| |
| [[
| |
| ]]
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| : 3) Provide training to the operators on the new SSD procedures for [[
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| ]]
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| : 4) Conduct drills to ensure viability on the new [[ ]] safe shutdown procedures.
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| : 15. Section 3.5.2: Revise Fleet Directive NSD-403 and LAR Attachment D Fire Protection Site Directive (SD) 1.3.5 with the VFDR ID #
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| during NPO Modes definition of high(er) risk evolution Oconee site calculation (HRE) to address non-power operation (OSC)-9268-01 (NPO) criteria, e.g., Plant Operating State (POS) 18. Also, reconcile NSD 403 and SD 1.3.5 Thermal Margin Criteria with the criteria in FAQ 07-0040 as needed.
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| : 16. Section 3.5.2: Develop a process to evaluate the LAR Attachment D Fire Protection potential effects of a fire upon VFDR ID#
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| during NPO Modes habitability and the impact of increased OSC-9268-02 DID fire protection actions that can be added to the establishment of high OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 21 confidence [[
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| ]] per Fleet Directive NSD-403.
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| : 17. Section 3.5.2: Implement procedural guidance to LAR Attachment 0 Fire Protection monitor [[ VFDR 10 #
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| during NPO Modes OSC-9268-03
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| ]]
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| : 18. Section 3.5.2: Develop procedural controls to monitor LAR Attachment 0 Fire Protection [[ VFDR 10 #
| |
| during NPO Modes OSC-9268-04
| |
| ]] flow path during higher risk evolutions (HREs) for the outage risk manaQement procedures.
| |
| : 19. Section 3.5.2: Develop procedural controls for use of LAR Attachment 0 Fire Protection [[ VFDR 10 #
| |
| during NPO Modes OSC-9268-05
| |
| ]] during HREs for the outage risk management procedures applicable to NPO key safety function (KSF).
| |
| : 20. Section 3.5.2: Develop procedural controls on the LAR Attachment 0 Fire Protection [[ VFDR 10 #
| |
| during NPO Modes ]] for the outage OSC-9268-06 risk management procedures.
| |
| : 21. Section 3.5.2: Ensure capability to access (Le., an LAR Attachment 0 Fire Protection operator can be dispatched to manually VFDR 10 #
| |
| during NPO Modes throttle) motor-operated valves (MOVs) OSC-9268-07
| |
| [[
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| 11
| |
| : 22. Section 3.5.2: Ensure capability to access (Le., an LAR Attachment 0 Fire Protection operator can be dispatched to manually VFDR 10 #
| |
| during NPO Modes open and close, respectively) manual OSC-9268-08 valves [[
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| 11
| |
| : 23. Section 3.5.2: Complete the analysis of NPO fire LAR Attachment 0 Fire Protection impacts for fire zones following VFDR 10 #
| |
| during NPO Modes installation of the NFPA 805 committed OSC-9313-02 modifications. After implementation, update Oconee Site Calculation (OSC) 9313 and its NPO recommendations for affected fire zones.
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| : 24. Section 3.5.2: Develop procedure guidance for pre LAR Attachment 0 Fire Protection emptive re-alignment of and the removal VFDR 10 #
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| during NPO Modes of power from the MOVs [[ OSC-9313-03
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| ]]
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| : 25. Section 3.5.2: Revise NSD-403, SO 1.3.5 and ONS LAR Attachment 0 Fire Protection technical procedures to implement the VFDR 10 #
| |
| during NPO Modes recommendations in OSC-9313, OSC-9313-07 Attachment 1, subject to resolution of open Items (Le., Items 15 throuQh 24).
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATIO~J 22
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| : 26. Section 3.1.3.10: Revise FPP Design Basis Specification LAR Attachment L Fire Hose Standpipes Use for the [[ ]] fire hydrants. Approval Request #10 of outside Fire Hydrant Appurtenances
| |
| : 27. Section 3.8.2: Configuration control procedures which LAR Section 4.7.2 Configuration Control govern the various ONS documents and databases will be revised to reflect the new RI/PB FPP licensing bases.
| |
| : 28. Section 3.8.3: Training Position Specific Guides will be LAR Section 4.7.3 Quality developed to identify and document required training and mentoring to ensure individuals are appropriately qualified per NFPA 805, Section 2.7.3.4.
| |
| : 29. Section 3.8.3: Post-transition quality requirements from LAR Section 4.7.3 Quality NFPA 805 that are not currently part of the ONS processes will be revised to include any additional requirements.
| |
| : 30. Attachment D: Operator Guidance - ONS procedures LAR Attachment C
| |
| [[ will be updated to include the following: [[
| |
| ]] 1) Guidance for maintaining the plant ]]
| |
| safe and stable following loss of all
| |
| [[ ]]
| |
| : 2) Guidance for operation of [[
| |
| ]]
| |
| : 31. Attachment D: Resolve the physical location issue of LAR Attachment C
| |
| [[ ]] the [[ [[ ]]
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| ]] requirements by revising the fire risk evaluation to denote the physical separation aspects of the
| |
| [[ ]]
| |
| : 32. Attachment D: Incorporate [[ ]] into LAR Attachment C
| |
| [[ ]] FPP site documents after the modification is implemented.
| |
| : 33. Section 3.2.1: Incorporating all related non-coordinated LAR Sections 4.2, Section 3.3.1.7, information in the NSCA and NPO Pinch 4.3, & 4.5, Section 3.3.3.3, Point Analysis, and updating the Fire RAI Response Section 3.5.2.4, PRA model, to include the results of the (Reference 54)
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| Section 3.5.2.5, breaker coordination study
| |
| : 34. Section 3.1.3.4 Future acceptable cable construction LAR Section 4.1, Use of Unqualified Video / qualifications will be included in the RAI Response Communication / Data Power Generation Electrical Discipline (Reference 12)
| |
| Cables Design Criteria Manual. A specific line item will be added that video /
| |
| communication / data cables shall be plenum rated and/or tested in accordance with Institute of Electrical and Electronic Engineers (IEEE) 383 1974, IEEE 1202-1991, CSA 22.2 No.
| |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 23 0.3, NFPA 262, UL 44, UL 83, UL 1581, UL 1666, or UL 1685 as accepted in FAQ 06-0022. Electrical wiring, including video, phone, and communications, installed above a suspended ceiling shall be rated for plenum use, routed in metallic conduit, routed in cable tray with solid metal top and bottom covers, or armored cable.
| |
| : 35. Attachment A: Appropriate directives will be updated to LAR Section 4.1, Section 3.3.1.3.4 clearly indicate that only portable request for additional Plant Administrative electric heaters are permitted to be used information (RA)I Procedures in plant areas with equipment important Response (Reference to nuclear safety or where there is the 52) potential for radiological release due to fire. Portable fuel-fired heaters are not permissible in these areas.
| |
| : 36. Section 3.1.3.7 The fire brigade will develop a SOG for LAR Section 4.1 ,
| |
| [[ fighting a fire [[ ]] Training is RAI Response
| |
| ]] already performed on tactics for fighting (Reference 52) fires of this nature but training will be reinforced with a new SOG. The Fire Brigade Administrator will review the Pre-Fire Plans to determine if enhancement is necessary.
| |
| : 37. Section 3.7 Develop instructions for the software LAR Section 4.6, Monitoring Program program to collect availability and RAI Response reliability data on SSCs in the (Reference 52)
| |
| Monitoring Program.
| |
| : 38. Section 3.2.1 Revise the B-2 Table to include LAR Section 4.2, NSCA Methods additional clarification of alignment with RAI Response the NEI guidance. Reference (12)
| |
| : 39. Section 3.2.1 Development and documentation of a LAR Section 4.2, NSCA Methods long term SSD program including RAI Response analysis, equipment reviews, recovery (References 52 & 54) actions, modifications, and procedural guidance.
| |
| : 40. Section 3.2.1 Complete activities needed to provide LAR Section 4.2, NSCA Methods assurance that fire-induced open RAI Response secondary circuits of current (Reference 12) transformers will not impact the ability to achieve and maintain the fuel in a safe and stable condition.
| |
| : 41. Section 3.4.3 With regard to the Internal Events PRA, LAR Section 4.5, PRA Quality complete the following: RAI Response
| |
| * Determine whether the HRA model (Reference 59) needs to be updated or upgraded.
| |
| * Update/upgrade the HRA model.
| |
| * If HRA model was upgraded conduct a focus-scope peer review of the revised internal events PRA model OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 24 with respect to HRA.
| |
| I
| |
| * Disposition all findings from the peer review and revise the internal events PRA, as appropriate.
| |
| 42 Section 3.4.3 With regard to the Fire PRA, complete LAR Section 4.5, Fire PRA Quality the following: RAI Response
| |
| * Update/upgrade the Fire PRA as (Reference 59) appropriate to resolve NRC staff review findings in SE Attachment C, Table 3.4-2.
| |
| * Complete an industry full-scope peer review of the revised Fire PRA that is performed to the ASME/ANS RA Sa-2009 PRA standard, as endorsed by RG 1.200, Rev. 2. The full-scope peer review will include specific focus on the following elements:
| |
| * Influence on the target set from fire propagation beyond the ignition source due to intervening combustibles and cables on:
| |
| * Expanding the zone of influence (ZOI), both vertically and horizontally, and
| |
| * hot gas layer (HGL) formation, including the effects on fire detection and brigade response.
| |
| * Modeling of high-energy arcing faults on [[ ]]
| |
| bus ducts.
| |
| * Deviation from NUREG/CR-6850 guidance and as modified by closed FAQs will be treated as described in NEI 07-12 (Fire Probabilistic Risk Assessment Peer Review Guidelines) and the fire aspects of ASME/ANS PRA Standard, as endorsed by RG 1.200.
| |
| * Disposition findings from the full-scope Fire PRA peer review and revise the Fire PRA as appropriate.
| |
| 43 Section 3.4.6 Cumulative Confirm that the risk decrease from the LAR Section 4.2, Risk and Combined as-built [[ ]] continues to bound the RAI Response Changes cumulative VFDR transition risk once (Reference 52) the [[ ]]are installed 44 Section 3.2.1: The breaker coordination study will be LAR Section 4.2, Section 3.3.1.7, updated to include all new NFPA 805 RAI Response Section 3.3.3.3, SSD equipment list (SSEL)-related (Reference 52)
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 25 Section 3.5.2.4, power supplies (i.e., PSW) for power Section 3.5.2.5, and non-power operations, and additional plant modification will be defined if necessary to ensure that the assumptions of the Fire PRA and NSCA remain valid.
| |
| : 45. Section 3.8.1: The ONS "Fire Protection Program LAR Section 4.7.1 Documentation Design Basis Document" and supporting documentation will be revised to incorporate NFPA 805 documents.
| |
| 46 Section 3.2.1 Licensee agreed to eliminate the "10 LAR Section 4.2, NSCA Methods minute free of fire damage" assumption. RAI Response The ONS FPP and supporting (Reference 52).
| |
| documentation (including the B-2 Table, B-3 Table, all applicable fire risk evaluations, Fire PRA, NSCA, and operator manual action(s) (OMA) feasibility calculations) will be revised to eliminate the assumptions. Compliance will be demonstrated consistent with NFPA 805, Section 4.2.4.2.
| |
| 47 Attachment B: Revised calculation OSC-9291, NFPA RAI Response Section 3.1.1.7 805 Transition B-2 Table, Section (Reference 12) 3.1.1.7 to reword the alignment basis to clearly state that [[ ]] is not credited for deterministic analysis and therefore not analyzed for its availability in the deterministic analysis. The licensee also states that alignment statement will be revised to ensure the proper relationship with the alignment basis.
| |
| | |
| ==3.0 TECHNICAL EVALUATION==
| |
| | |
| The following sections evaluate the technical aspects of the requested license amendment to transition the FPP at ONS to one based on NFPA 805 in accordance with 10 CFR 50.48(c).
| |
| While performing the technical evaluation of the licensee's submittal, the NRC staff utilized the guidance provided in NUREG-0800, Chapter 9.5.1.2, "Risk-Informed, Performance-Based Fire Protection" (Reference 18), 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 SE Section 2.0. Specifically:
| |
| * Section 3.1 provides the results of the NRC staff's review of the licensee's transition of the FPP from the existing deterministic guidance to that of NFPA 805, Chapter 3, "Fundamental Fire Protection Program and Design Elements."
| |
| * Section 3.2 provides the results of the NRC staff's review of the methods used by the licensee to demonstrate the ability to meet the nuclear safety performance criteria.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 26
| |
| * Section 3.3 provides the results of the NRC staff's review of the fire modeling methods to demonstrate the ability to meet the nuclear safety performance criteria using a fire modeling PB approach.
| |
| * Section 3.4 provides the results of the NRC staff's review of the fire risk assessments used by the licensee to demonstrate the ability to meet the nuclear safety performance criteria using a FRE PB approach.
| |
| * Section 3.5 provides the results of the NRC staffs review of the licensee's NSCA results by fire area.
| |
| * Section 3.6 provides the results of the NRC staff's review of the methods used by the licensee to demonstrate the ability to meet the radioactive release performance criteria.
| |
| * Section 3.7 provides the results of the NRC staff's review of the NFPA 805 monitoring program developed as a part of the transition to the a RI/PB FPP based on NFPA 805.
| |
| * Section 3.8 provides the results of the NRC staff's review of the licensee's approach to program documentation, quality assurance, and configuration management.
| |
| Attachments A - E to this SE provides additional detailed information that was evaluated and/or dispositioned by the NRC staff to support the licensee's request for transition to an RI/PB FPP in accordance with NFPA 805 (i.e., 10 CFR 50.48(c)). These attachments are discussed as appropriate in the associated section of the SE.
| |
| 3.1. NFPA 805 Fundamental FPP Elements and Minimum Design Requirements NFPA 805, Chapter 3, contains the fundamental elements of the FPP and specifies the minimum design requirements for fire protection systems and features that are necessary to meet the standard. 10 CFR 50.48(c) takes exception to three specific requirements of NFPA 805, Chapter 3, and provides alternative requirements as follows:
| |
| * 10 CFR 50.48(c)(2)(v) - Existing cables. In lieu of installing cables meeting flame propagation tests as required by Section 3.3.5.3 of NFPA 805, a 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 of NFPA 805 is not endorsed.
| |
| * 10 CFR 50.48(c)(2)(vi) - Water supply and distribution. The italicized exception to Section 3.6.4 of NFPA 805 is not endorsed. Licensees who wish to use the exception to Section 3.6.4 of NFPA 805 must submit a request for a license amendment in accordance with 10 CFR 50.48(c)(2)(vii).
| |
| * 10 CFR 50.48(c)(2)(vii) - Performance-based methods. While Section 3.1 of NFPA 805 prohibits the use of performance-based methods to demonstrate compliance with the NFPA 80S, Chapter 3, requirements, 10 CFR 50.48(c)(2)(vii) specifically permits that the FPP elements and minimum design requirements of NFPA 805, Chapter 3, may be subject to the performance-based methods permitted elsewhere in the standard.
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| OFFICIAL USE ONLY SECURITY RELJ\TED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 27 Furthermore, Section 3.1 of NFPA 805 specifically allows the use of alternatives to the NFPA 805, Chapter 3, fundamental FPP requirements that have been previously approved by the NRC, which is the authority having jurisdiction (AHJ), as denoted in the NFPA 805 standard.
| |
| The licensee identified an implementation action to modify the ONS FPP to include the statement that the NRC is the AHJ (SE Section 2.9, Table 2.9-1, Item 1).
| |
| 3.1.1. Compliance with NFPA 805, Chapter 3, Requirements The licensee used the systematic approach described in NEI 04-02, Revision 2 (Reference 21),
| |
| as endorsed by the NRC in RG 1.205, Revision 1 (Reference 14), to assess the proposed ONS FPP against the NFPA 805, Chapter 3, requirements.
| |
| As part of this assessment, the licensee reviewed each section and subsection of NFPA 805, Chapter 3, against the existing FPP and provided specific compliance statements for each NFPA 805, Chapter 3, attribute that contained applicable requirements. As discussed below, some subsections of NFPA 805, Chapter 3, do not contain requirements, or are otherwise not applicable to ONS.
| |
| The methods used by the licensee for achieving compliance with the NFPA 805, Chapter 3, fundamental FPP elements and minimum design requirements are as follows:
| |
| : 1. The existing FPP element directly complies with the requirement; noted in LAR Attachment A, "NEI 04-02 Table B-1, Transition of Fundamental FPP and Design Elements (NFPA 805, Chapter 3)," also called the B-1 Table, as "Comply."
| |
| : 2. The existing FPP element complies through the use of an explanation or clarification; noted in the B-1 Table as "Complies with Clarification."
| |
| : 3. The existing FPP element complies with the requirement based on prior NRC approval of an alternative to the fundamental FPP attribute and the bases for the NRC approval remain valid; noted in the B-1 Table as "Complies by Previous NRC ApprovaL"
| |
| : 4. The existing FPP element complies through the use of existing engineering equivalency evaluation (EEEEs) whose bases remain valid and are of sufficient quality; noted in the B-1 Table as "Complies with use of EEEE."
| |
| : 5. The existing FPP element does not comply with the requirement, but the licensee is requesting specific approval for a performance-based method in accordance with 10 CFR 50.48(c)(2)(vii); noted in the B-1 Table as "Submit for NRC ApprovaL" The licensee stated in LAR Section 4.2.2, "Engineering Equivalency Evaluation Transition," that they had evaluated the EEEEs used to demonstrate compliance with the NFPA 805, Chapter 3, requirements in order to ensure continued appropriateness, quality, and applicability to the current ONS plant configuration. The licensee determined that no EEEE used to support compliance with NFPA 805 required NRC approval.
| |
| Additionally, the licensee stated in LAR Section 4.2.3, "Licensing Action Transition," that the existing licensing actions included a provision to demonstrate compliance have been evaluated to ensure that their bases remain valid. The results of these licensing action evaluations were provided in the LAR Attachment K.
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| OFFICIAL USE ONLY SECURITY RELATED I~JFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 28 Attachment A,Table 3.1-1, "NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix,"
| |
| in Attachment A to this SE, provides the specific FPP elements and minimum design requirements from NFPA 805, Chapter 3, as appropriately modified by 10 CFR 50.48(c). In addition, the table describes each fundamental FPP element from NFPA 805, Chapter 3, and identifies which of the methods the licensee used as the means for achieving compliance with the requirement.
| |
| Attachment A, Table 3.1-1 also provides the results of the NRC staff's evaluation of the licensee's compliance statement for each FPP element. LAR Attachment A (the NEI 04-02 B-1 Table) provides further details regarding the licensee's compliance strategy for specific NFPA 805, Chapter 3, requirements, including references to where compliance is documented.
| |
| For approximately 68 percent of the NFPA 805, Chapter 3, requirements, as modified by 10 CFR 50.48(c)(2), the licensee determined that the RI/PB FPP complies directly with the fundamental FPP element. In these instances, based on the validity of the licensee's statements, the NRC staff finds the licensee's compliance method/strategy acceptable.
| |
| For approximately 2 percent of the NFPA 805, Chapter 3, requirements, the licensee provided additional clarification when describing its means of compliance with the fundamental FPP element. In these instances, the NRC staff reviewed the additional clarifications and concludes that the licensee meets the underlying requirement for the FPP element as clarified.
| |
| For approximately 14 percent of the NFPA 805, Chapter 3, requirements, the licensee demonstrated compliance with the fundamental FPP element through the use of EEEEs. Based on the licensee's statement of validity provided in Tables B-1 and B-3, the NRC staff finds the licensee's statements of compliance in these instances acceptable.
| |
| Approximately 1 percent of the requirements were supplanted by an alternative that was previously approved by the NRC. In two instances, NRC approval was documented in the original August 11, 1978, FPP SE report (Reference 26), and the other two instances were approved in an NRC Exemption (ADAMS Accession No. ML012000058) dated August 21,1989 (Reference 27). NFPA 805 allows the justification for exemptions to be carried forward in the transition to NFPA 805 as PB evaluations.
| |
| In each instance, 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 that previous NRC approval has been demonstrated using suitable documentation that meets the approved guidance contained in RG 1.205, Revision 1. Based on the licensee's justification of the previously approved alternatives to the NFPA 805, Chapter 3, requirements, as well as the NRC staff's review of this information, the NRC staff finds the licensee's statements of compliance in these instances acceptable.
| |
| In the compliance statements for approximately 12 percent (13 of the requirements) of the NFPA 805, Chapter 3, requirements, the licensee used more than one of the above strategies to demonstrate compliance with all aspects of the fundamental FPP elements. In each of these cases, the NRC staff found the compliance statements acceptable, for the reasons outlined above.
| |
| In 11 instances, the licensee requested approval for the use of PB methods to demonstrate compliance with a fundamental FPP element. In accordance with 10 CFR 50.48(c)(2)(vii), the licensee requested specific approval be included in the license amendment approving the OFFICIAL USE ONLY SECURITY RELATED INFORMATIO~J
| |
| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 29 transition to NFPA 805. The requested PB methods pertain to the following requirements further discussed in Section 3.1.3 in this SE.
| |
| Some NFPA 805, Chapter 3, sections either do not apply to the transition to a RI/PB FPP at ONS, or have no technical requirements. Accordingly, the NRC staff did not review these sections for acceptability. The non-reviewed sections fall into one of two categories:
| |
| * Sections that do not contain any technical requirements (e.g., NFPA 80S, Chapter 3, Section 3.1, and Section 3.4.5).
| |
| * Sections that are not applicable to ONS because of the following:
| |
| The licensee states that ONS does not have systems of this type installed (e.g., the NFPA 80S, Chapter 3, Section 3.9.4 requirements for diesel-driven water fire pumps, Section 3.10 requirements for gaseous suppression systems, and Section 3.11.5 requirements for electrical raceway fire barrier systems).
| |
| The requirements are structured with an applicability statement (e.g., NFPA 805, Chapter 3, Section 3.4.1 (a)(2) and Section 3.4.1 (a)(3), wherein the determination of which NFPA code(s) apply to the fire brigade depends on the type of brigade specified in the FPP).
| |
| As documented in Attachment A, SE Table 3.1-1 and discussed above, the NRC staff evaluated the results of the licensee's assessment of the proposed ONS RI/PB FPP against the NFPA 80S, Chapter 3, fundamental FPP elements and minimum design requirements, as modified by the exceptions, modifications, and supplementations in 10CFR 50.48(c)(2). Based on this review of the licensee's LAR, and supplements, the NRC staff finds the RI/PB FPP acceptable with respect to the fundamental FPP elements and minimum design requirements of NFPA 805, Chapter 3, as modified by 10 CFR 50.48(c)(2), because the licensee accomplished the following:
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| * used an overall process consistent with NRC staff approved guidance to determine the state of compliance with each of the applicable NFPA 805, Chapter 3, requirements.
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| * provided appropriate documentation of ONS's state of compliance with the NFPA 805, Chapter 3, requirements, which adequately demonstrated compliance in that the licensee was able to substantiate that it complied:
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| - with the requirement directly.
| |
| - with the intent of the requirement (or element) given adequate justification.
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| via previous NRC staff approval of an alternative to the requirement.
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| through the use of an EEEE.
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| - through the use of a combination of the above methods.
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| through the use of a PB method that the NRC staff has specifically approved in accordance with 10 CFR 50.48(c)(2)(vii).
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 30 3.1.2. Identification of Power Block The NRC staff reviewed the ONS structures identified in LAR Table 1-1, "ONS Power Block Definition," 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 ONS RI/PB FPP that have additional requirements in accordance with 10 CFR 50.48(c) and NFPA 805. As stated in the LAR, power block equipment includes SSCs required for the safe and reliable operation of the station. It includes all safety-related and balance-of-plant systems and components required for operation, including radioactive waste processing and storage, the 230 kV switchyard, Keowee Dam and associated structures, and the PSW Facility and associated electrical duct banks.
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| This equipment does not include buildings or structures that support station staff, such as offices or storage structures, or the ventilation and support systems focused only on habitability of those structures. The NRC staff finds that the licensee has appropriately evaluated the structures and equipment at ONS, and adequately documented a list of those structures that fall under the definition of "power block" in NFPA 805.
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| 3.1.3. Performance-Based Methods for NFPA 805, Chapter 3, Elements Performance-Based Methods, Section 50.48(c)(2)(vii):
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| The prohibition in Section 3.1 of NFPA 805 that does not permit the use of performance based methods for the Chapter 3 fundamental fire protection program elements and minimum design criteria is not endorsed. The NRC takes this exception in order to provide licensees greater flexibility in meeting the fire protection program elements and minimum design requirements of Chapter 3 by the use of performance-based methods (including the use of risk-informed methods) described in the NFPA 805 standard. This approach is acceptable to NRC because the rule requires NRC review and approval prior to the licensee's use of those methods, and the rule sets forth criteria for evaluating the acceptability of the licensee's proposed use of performance-based methods in meeting the fire protection program elements and minimum design requirements.
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| Final Rule, Voluntary Fire Protection Requirements for Light Water Reactors; Adoption of NFPA 805 as a Risk-Informed, Performance-Based Alternative, (69 FR 33536, 33543, June 16, 2004)
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| (describing Performance-Based Methods of Section50.48(c)(2)(vii)).
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| 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 of NFPA 805, Chapter 3. According to 10 CFR 50.48(c)(2)(vii), an acceptable PB approach accomplishes the following:
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| (A) Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; (B) Maintains safety margins; and (C) Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire SSD capability).
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OffiCIAL USE ONLY SECURITY RELATED INfORMATIO~J 31 In LAR Section 4.1.2.3, "NFPA 805, Chapter 3, Requirements Not Specifically Met nor Previously Approved by NRC," and Attachment L, "NFPA 805, Chapter 3, Requirements for Approval," the licensee requested NRC staff review and approval of 11 PB methods to demonstrate an equivalent level of fire protection for the following NFPA 805, Chapter 3, requirements:
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| * Section 3.3.1.2, which concerns the use of untreated wood for use in concrete forming at ONS,
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| * Section 3.3.5.1, which concerns the use of wiring above suspended ceilings at ONS,
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| * Section 3.3.5.3, which concerns the use of low-voltage cable at ONS that does not comply with an acceptable flame propagation test,
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| * Section 3.3.7.1, which concerns the storage of bulk quantities of flammable gas cylinders at ONS,
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| * Section 3.3.12(1), which concerns collection of oil mist from the reactor coolant pump (RCP) oil system at ONS,
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| * Section 3.5.3, which concerns the omission of relief valves on the high-pressure service water (HPSW) and Keowee fire pumps at ONS,
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| * Sections 3.5.3,3.5.16,3.6.1, and 3.6.2, which concerns the use of fire hose stations and fire pumps at ONS that do not meet NFPA 14 and NFPA 20, respectively,
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| * Section 3.5.6, which concerns the use of fire pumps that have an automatic stop function at ONS,
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| * Sections 3.5.7, 3.5.10, and 3.5.15, which concerns the use of a fire protection system at ONS that does not have sectionalizing valves and does not meet NFPA 24,
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| * Section 3.5.16, which concerns the use of the fire protection fire water system at ONS that has dual purposes, and
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| * Sections 3.5.3 and 3.5.4, which concerns the use of a fire protection system at KHS that cannot provide 100 percent of the required flow rate and pressure at all times.
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| The NRC staff's review of the licensee's proposed methods is provided below.
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| 3.1.3.1 Use of Non-treated Wood In LAR Attachment L, "NFPA 805, Chapter 3, Requirements for Approval," the licensee requested the NRC staff's review and approval of a PB method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805, Section 3.3.1.2, "Control of Combustible Material," regarding the use of untreated wood for concrete forming. Specifically the licensee has requested approval of a PB method to justify the use of non-pressure impregnated or fire retardant (untreated) wood within the ONS power block for concrete forming.
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| OffiCIAL USE ONLY SECURITY RELATED INfORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 32 The licensee stated that the basis for the approval request of this PB method is:
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| * In some cases, the chemicals used in the treatment of fire-retardant wood affect concrete curing.
| |
| * A small quantity of untreated wood used for concrete forming is acceptable because the magnitude of the additive combustible material would be insignificant as compared to the total fire load in the area.
| |
| * The locations of concrete forming are generally not in close proximity to ignition sources.
| |
| * Concrete forming is for temporary use and not for permanent plant installation.
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| The licensee stated that the use of untreated wood for concrete forming would have no adverse impact on nuclear safety performance because (1) concrete forming is used infrequently within the ONS power block, (2) it is generally in such small quantities that it would have a negligible impact to the in-situ fire load and would be within the permissible transient fire load, and (3) if the quantity of untreated wood exceeds the permissible limits established in ONS administrative controls, a fire protection engineer (FPE) review would be conducted and result in the identification and implementation of special precautions or limitations, as necessary, on the use of the untreated wood.
| |
| By letter dated September 13, 2010, (Reference 12), the licensee stated, the proposed PB method maintains SMs and conservatisms because (1) the quantity of untreated wood used in concrete forming is minimal and the quantity is reviewed with special precautions or limitations identified as necessary in order to minimize fire risk, and (2) the precautions and limitations ensure that the quantity of these materials is maintained within the limitations and assumptions of the Fire PRA.
| |
| By letter dated September 13, 2010, (Reference 12), the licensee stated that the introduction of untreated wood for concrete forming does not impact fire protection DID because (1) its use is administered under the ONS combustible control program, and (2) automatic or manual fire suppression functions, fire protection for systems and structures, and post-fire SSD capability are not compromised since quantities of untreated wood used for concrete forming cannot be introduced such that they may challenge any elements of the FPP without appropriate compensatory measures being identified during the work review process.
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| By letter dated September 13, 2010, (Reference 12), the licensee stated that the use of untreated wood for concrete forming will have no impact on the radiological release performance criteria because (1) the introduction of untreated wood does not change the conclusion of the radiological release evaluation performed for each fire zone that potentially contaminated water is contained and smoke is monitored since fire brigade control of water runoff and smoke is not hindered because of the existence of the small quantity of untreated wood and (2) untreated wood does not add additional radiological materials to the area or challenge systems boundaries that contain such.
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| The NRC staff finds that the proposed PB method:
| |
| * does not impact the NFPA 805 nuclear safety performance measures (goals, objectives, and performance criteria) because (1) the quantity of untreated wood is expected to be relatively small when compared to the total combustible loading in the area, (2) the difference in combustible loading between treated and untreated wood for small quantities OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 33 exposed during a limited amount of time will not present a challenge to the fire protection features in place, and (3) administrative controls used while untreated wood is in the plant provides additional assurance of minimal impact should the untreated wood be exposed to a nearby fire,
| |
| * maintains the SMs of the licensee's analyses based on the licensee's statement that the precautions and limitations identified to minimize fire risk ensure that the quantity of untreated wood used for concrete forming is maintained within the limitations and assumptions of the Fire PRA,
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| * maintains fire protection DID since automatic or manual fire suppression functions, fire protection for systems and structures, and post-fire SSD capability are not compromised, and
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| * will have no effect on the NFPA 805 radiological release performance measures (goals, objectives, and performance criteria) since there will be no impact on fire suppression activities.
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| In accordance with 10 CFR 50.48(c)(2)(vii), the NRC staff finds the proposed PB method acceptable for application at ONS in lieu of the corresponding NFPA 805, Section 3.3.1.2 requirement because it satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release, maintains sufficient SM, and maintains adequate fire protection DID.
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| 3.1.3.2 Use of Compressed Flammable Gas Storage in the Power Block In LAR Attachment L, "NFPA 805, Chapter 3, Requirements for Approval," the licensee requested NRC staff review and approval of a PB method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805, Section 3.3.7, "Bulk Flammable Gas Storage."
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| The licensee has requested approval of a PB method to justify the storage of flammable gas cylinders in four locations in the AB where gas bottles are already installed. The locations identified are the chemistry labs and the post-accident monitoring instrumentation rooms.
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| Plant administrative controls contain specific instructions for segregating and storing compressed gas cylinders. Chemistry labs use hydrogen on a daily basis and have reserve tanks staged for continued use. The Chemistry labs and reference gases are located in the ONS AB, which also contains systems, equipment, and components important to nuclear safety.
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| The hydrogen cylinders are stored and controlled in accordance with ONS administrative and operating procedures.
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| The licensee stated that the basis for the approval request of this PB method is:
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| * Staging of flammable gas cylinders is required in four locations, which house systems, equipment, or components important to nuclear safety. Typically, one bottle is connected to the system and the minimum number of required bottles are staged in the area for continued use.
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| * The flammable gas cylinders in this evaluation exist in the plant ora fire protection engineering review will be performed prior to any new installation.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 34
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| * Gas cylinders staged but not in use are segregated and stored in accordance with ONS administrative procedures and design review processes.
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| * The flammable gas cylinders are stored in locations that do not impact equipment important to nuclear safety:
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| o The Chemistry Labs are located on the 796' elevation of the Units 2 and 3 AB (Fire Zones 90 and 86).
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| o The post-accident monitoring instrumentation is located on the 838' elevation of the AB in the Units 1/2 air-handling unit (AHU) and Unit 3 AHU rooms (Fire Zones 119 and 116).
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| By letter dated September 13, 2010, (Reference 12), the licensee stated that the storage of flammable gas cylinders in the four identified locations will have no impact on the nuclear safety performance criteria because (1) the four locations have been analyzed in the Fire PRA in the current configuration which includes the presence of the flammable gas cylinders, and (2) hydrogen fires in the four locations do not impact any targets.
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| Similarly, the licensee stated by letter dated September 13, 2010, (Reference 12) that the proposed PB method maintains SMs and conservatisms because (1) the four locations have been analyzed in the current Fire PRA in their current configuration which includes the presence of the flammable gas cylinders, (2) hydrogen fires in these locations do not impact any targets, and (3) the method does not change the assumptions and limitations of the analytical methods used in the development of the Fire PRA.
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| By letter dated September 13, 2010, (Reference 12), the licensee stated that the storage of flammable gas cylinders in the four locations does not impact fire protection DID because (1)
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| ONS administrative controls require the introduction of a new compressed gas cylinder be evaluated by the fire hazard review process, and (2) the introduction of flammable gas cylinders does not result in compromising automatic or manual fire suppression functions, fire protection for systems and structures, or post-fire SSD capability.
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| By letter dated September 13, 2010, (Reference 12), the licensee stated that the proposed PB method has no impact on the radiological release performance criteria because (1) the introduction of flammable gas cylinders in the four locations does not change the conclusion of the radiological release evaluation performed for each fire zone that potentially contaminated water is contained and smoke is monitored since fire brigade control of water runoff*and smoke is not hindered because of the existence of the gas cylinders, and (2) flammable gas cylinders do not add additional radiological materials to the area or challenge systems boundaries that contain such.
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| The NRC staff finds that the proposed PB method:
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| * does not impact the NFPA 805 nuclear safety performance measures (goals, objectives, and performance criteria) because (1) the specific locations of the cylinders have been analyzed in the Fire PRA for the proposed configuration, (2) the bottle use and storage are controlled by procedure, (3) gas bottles are segregated by procedure and design process, and (4) hydrogen fires due to failure of the gas cylinders do not impact any targets, OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE O~JLY SECURITY RELATED INFORMATION 35
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| * maintains the SMs of the licensee's analyses based on the licensee's statement that the method does not change the assumptions and limitations of the analytical methods used in the development of the Fire PRA,
| |
| * maintains fire protection DID since automatic or manual fire suppression functions, fire protection for systems and structures, and post-fire SSD capability are not compromised, and
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| * will have no effect on the NFPA 805 radiological release performance measures (goals, objectives, and performance criteria) since there will be no impact on fire suppression activities.
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| In accordance with 10 CFR 50.48(c)(2)(vii), the NRC staff finds the proposed PB method acceptable for application at ONS in lieu of the corresponding NFPA 805, Section 3.3.7 requirement because it satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release, maintains sufficient SM, and maintains adequate fire protection DID.
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| 3.1.3.3 Use of Non-listed I Unapproved Wiring above the Suspended Ceiling In LAR Attachment L, "NFPA 805, Chapter 3, Requirements for Approval," the licensee requested the NRC staff's review and approval of a PB method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805, Section 3.3.5.1 regarding wiring above suspended ceilings. The licensee has requested approval of a PB method to justify the use of existing wiring above suspended ceilings that may not comply with the requirements of this section. As described by the licensee, this concerns areas at ONS currently with suspended ceilings inside the power block consisting of offices, labs, elevator lobbies, corridors, and change rooms. The areas include:
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| * Control Rooms / Lobbies
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| * TB office areas
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| * AB stair and/or elevator lobbies
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| * AB office areas (Health Physics/Chemistry)
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| * AB Change Areas
| |
| * 838' elevation AB corridor With the exception of the Control Rooms, the licensee stated that these areas are not risk significant.
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| The licensee stated that the basis for the approval request of this PB method is:
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| * Power and control cables comply with requirements of "plenum rated" equivalent or armored.
| |
| * The wiring above ceilings in offices, lobbies, and laboratories, do not pose a hazard:
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 36 o Low voltage is not susceptible to shorts causing a fire. The licensee defines that video/communication/data cables are low voltage.
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| o By eliminating cables with the potential shorts, this eliminates ignition sources and therefore the jacketing of cable is not relevant.
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| o There is no equipment important to nuclear safety in the vicinity of these cables.
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| o Beginning in 2006, any new cables installed and the replacement of existing cables as part of upgrades are "plenum rated."
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| * The installation of detection above the Control Room ceilings will promptly identify a fire thereby enhancing fire brigade response time. The installation of detection is a committed plant modification (Section 2.8.1 of this SE; Modification 5).
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| * New power, control or instrumentation cable installed is constructed similar to or superior to the original cable and meets the requirements of IEEE-383, "IEEE Standard for Type Test of Class IE Electric Cables, Field Splices, and Connections for Nuclear Power Generating .
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| Stations," IEEE Standard 383.
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| The licensee further stated that for the cabling above the suspended ceilings in the control rooms has a very low possibility of a fire due to limited combustible loading, discontinuity of combustibles, and the inherent features of the electrical circuit design. In addition, the ventilation in the control rooms is a closed-loop system, which recirculates the air where either the existing detection or the control room operators who are continuously present in the area would quickly identify the smoke. The licensee indicated that an engineering change (EC) has been developed to install detection above the suspended ceiling area in the control room. The installation of new detection above the control room ceilings can promptly identify a fire thereby enhancing fire brigade response time and minimizing the impact to fire risk.
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| The licensee stated that the wiring above the suspended ceiling that may not comply with the requirement of NFPA 805, Section 3.3.5.1, does not impact the nuclear safety performance criteria because (1) with the exception of the control room, wiring above suspended ceilings is not in the vicinity of nuclear safety equipment, (2) power and control cables are armor jacketed, in metallic conduit, or plenum rated, and (3) low voltage cable is not susceptible to shorts that would result in a fire.
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| By letter dated September 13, 2010, (Reference 12), the licensee stated that the proposed PB method maintains SMs and conservatisms because the method does not change the assumptions and limitations of the analytical methods used in the development of the Fire PRA.
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| By letter dated September 13, 2010, (Reference 12), the licensee stated that the wiring above suspended ceiling that may not comply with the requirement of NFPA 805, Section 3.3.5.1 does not impact fire protection DID because the video/communication/data cables do not result in compromising automatic or manual fire suppression functions, fire protection for systems and structures, or post-fire SSD capability.
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| By letter dated September 13, 2010, (Reference 12), the licensee stated that the proposed PB method has no impact on the radiological release performance criteria because (1) the location of cables above suspended ceilings does not change the conclusion of the radiological release evaluation performed for each fire zone that potentially contaminated water is contained and OFFICIAL USE O~JLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 37 smoke is monitored since fire brigade control of water runoff and smoke is not hindered because of the existence of the low-voltage cables, and (2) the cables do not add additional radiological materials to the area or challenge systems boundaries that contain such.
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| The NRC staff finds that the proposed PB method:
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| * does not impact the nuclear safety performance measures (goals, objectives and performance criteria) because (1) the space enclosing these cables are non-combustible, (2) the location of wiring above suspended ceilings has a minimum amount of nearby ignition sources considering the adjacent armored power and control cables, (3) the video/communication/data cables are low energy and therefore pose a low fire ignition hazard due to hot shorts, and (4) fire detection will be installed above the suspended ceilings in the control rooms,
| |
| * maintains the SMs of the licensee's analyses based on the licensee's statements that (1) power and control cables comply with requirements of "plenum rated" equivalent or armored, (2) limited unqualified low-voltage wiring, and (3) that the method does not change the assumptions and limitations of the analytical methods used in the development of the Fire PRA,
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| * maintains fire protection DID since automatic and manual fire suppression functions, fire protection for systems and structures, and post-fire SSD capability are not compromised, and
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| * will have no effect on the radiological release performance measures (goals, objectives, and performance criteria) since there will be no impact on fire suppression activities.
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| The NRC staff finds the proposed PB method acceptable for application at ONS in lieu of the corresponding NFPA 805, Section 3.3.5.1, requirement because it satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release, maintains sufficient SM, and maintains adequate fire protection DID.
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| 3.1.3.4 Use of Unqualified Video/Communication/Data Cables In LAR Attachment L, "NFPA 805, Chapter 3, Requirements for Approval," the licensee requested NRC staff review and approval of a PB method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805, Section 3.3.5.3 regarding an acceptable flame propagation test for electric cable construction. The licensee has requested approval of a PB method to justify the use of existing wiring that may not comply with the requirements of this code section. As described by the licensee, video/communication/data cables installed at ONS are not necessarily tested in accordance with the flame propagation test requirements of IEEE 383 or any other qualification standard outlined in FAQ 06-0022 as endorsed by the NRC.
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| The licensee stated that the basis for the approval request of this PB method is:
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| * Power and control cable installed is constructed similar to or superior to the original cable and meets the requirements of IEEE-383.
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| * All new power, control or instrumentation cable installed will be constructed similar to or superior to the original cable and will meet the requirements of IEEE-383 or plenum rated OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 38 (NFPA 262, "Standard Method of Test for Flame Travel and Smoke of Wires and Cables for Use in Air Handling Spaces").
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| * Video/communication/data cables are low voltage and not susceptible to cause shorts and fires.
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| By letter dated September 13, 2010, (Reference 12), the licensee stated that acceptable cable construction qualifications will be included in the Power Generation Electrical Discipline Design Criteria Manual. The licensee identified an implementation action to modify this manual to add a specific line item that video/communication/data cables shall be plenum-rated and/or tested in accordance with IEEE 383-1974, IEEE 1202-1991, CSA (Canadian Standards Association) 22.2 No. 0.3, NFPA 262, UL (Underwriters Laboratory) 44, UL 83, UL 1581, UL 1666, or UL 1685 as accepted in FAQ 06-0022 (SE Section 2.9, Table 2.9-1, Item 34). Electrical wiring, including video, phone, and communications, installed above suspended ceilings shall be rated for plenum use, routed in metallic conduit, routed in cable tray with solid metal top and bottom covers, or armored cable.
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| By letter dated September 13, 2010, (Reference 12) the licensee stated that the proposed PB method maintains SMs and conservatisms because the method does not change the assumptions and limitations of the analytical methods used in the development of the Fire PRA.
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| By letter dated September 13, 2010, (Reference 12), the licensee stated that the video/communication/data cables that may not comply with the requirement of NFPA 805, Section 3.3.5.3 do not impact fire protection DID because (1) cable flame spread criteria are controlled by the licensee's design processes, and (2) video/communication/data cable construction does not result in compromising automatic or manual fire suppression functions, fire protection for systems and structures, or post-fire SSD capability.
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| By letter dated September 13, 2010, (Reference 12), the licensee stated that the proposed PB method has no impact on the radiological release performance criteria because (1) the construction of cables does not change the radiological release evaluation performed for each fire zone that potentially contaminated water is contained and smoke is monitored since fire brigade control of water runoff and smoke is not hindered because of the existence of the cables, and (2) cables do not add additional radiological materials to the area or challenge systems boundaries that contain such.
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| The NRC staff finds that the proposed PB method:
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| * does not impact the nuclear safety performance measures (goals, objectives and performance criteria) because (1) existing video/communication/data cables do not constitute a sjgnificant fire hazard, (2) future installation of cable will require appropriately approved flame spread criteria, (3) adjacent power and control cables are stated by the licensee to have acceptable flame spread qualities and therefore do not contribute significantly to the hazard, and (4) the licensee will require that all future installations of cable will comply with NFPA 805, Section 3.3.5.3,
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| * maintains the SMs of the licensee's analyses based on the licensee's statements that the method does not change the assumptions and limitations of the analytical methods used in the development of the Fire PRA, OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 39
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| * maintains fire protection DID since automatic and manual fire suppression functions, fire protection for systems and structures, and post-fire SSD capability are not compromised, and
| |
| * will have no effect on the radiological release performance measures (goals, objectives, and performance criteria) since there will be no impact on fire suppression activities.
| |
| In accordance with 10 CFR 50.48(c)(2)(vii), the NRC staff finds the proposed PB method acceptable for application at ONS in lieu of the corresponding NFPA 805, Section 3.3.5.1, requirement because it satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release, maintains sufficient SM, and maintains adequate fire protection DID.
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| 3.1.3.5 Allow Reactor Coolant Pump (RCP) Oil Mist Without Collection In LAR Attachment L, "NFPA 805, Chapter 3, Requirements for Approval," the licensee requested the NRC staff's review and approval of a PB method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805, Section 3.3.12(1) regarding the ONS RCP oil collection system. The licensee has requested approval of a PB method to justify not collecting oil mist resulting from pump/motor operation. This system was designed and was reviewed in accordance with 10 CFR Part 50, Appendix R, Section 111.0 to collect leakage from pressurized and nonpressurized leakage sites in the RCP oil system.
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| This however did not include collection of oil mist as a result of RCPs pump/motor operation.
| |
| As stated in Attachment L of the LAR, oil misting is not leakage due to equipment failure, but an inherent occurrence in the operation of large rotating equipment. It is normal for large motors to lose some oil throUgh seals and the oil to potentially become 'atomized' in the ventilation system. 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 oil mist resulting from normal operation will not adversely impact the ability of a plant to achieve and maintain SSD even if ignition occurred.
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| The licensee stated that the basis for the approval request of this PB method is:
| |
| * The oil collection system is designed to collect leakage from pressurized and nonpressurized 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 released as misting does not account for an appreciable heat release rate or accumulation near potential ignition sources or non-insulated reactor coolant piping.
| |
| * The RCPs use a synthetic oil of higher flash point, approximately 450 OF.
| |
| * There are redundant RCPs and they are not required to achieve or maintain fire SSD.
| |
| The licensee stated that the lack of an oil mist collection system for the RCPs does not impact the nuclear safety performance criteria because (1) oil mist does not significantly contribute to a fire heat release rate, (2) the synthetic oil used has a higher flashpoint of approximately 450°F, and (3) the equipment is not required for SSD.
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| OFFICIAL USE O~JLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 40 By letter dated September 13, 2010, (Reference 12), the licensee stated that the proposed PB method maintains SMs and conservatisms because (1) the oil mist resultant from normal operation will not adversely impact the ability of a plant to achieve and maintain fire SSD even if ignition occurred, (2) the RCPs are not required to achieve and maintain fire SSD, and (3) the PB method does not change the assumptions and limitations of the analytical methods used in the development of the Fire PRA.
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| By letter dated September 13, 2010, (Reference 12), the licensee stated that the oil mist from normal pump operation does not impact fire protection DID because (1) the RCP oil collection systems are controlled by the licensee's design processes, (2) oil misting does not result in compromising automatic or manual fire suppression functions, and (3) does not impact the post fire SSD capability.
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| By letter dated September 13, 2010, (Reference 12), the licensee stated that the proposed PB method has no impact on the radiological release performance criteria because (1) the entire RB during power operations is an environmentally sealed radiological area, (2) the potential for oil mist from the RCPs does not change the radiological release evaluation performed for each fire zone that potentially contaminated water is contained and smoke is monitored since fire brigade control of water runoff and smoke is not hindered because of the existence of the misting, and (3) the oil mist does not add additional radiological materials to the area or challenge systems boundaries that contain such.
| |
| The NRC staff finds that the proposed PB method:
| |
| * does not impact the nuclear safety performance measures (goals, objectives and performance criteria) because (1) RCP oil collection fire does not impact the ability to achieve SSD, (2) RCPs are not components identified as necessary for SSD, (3) oil mist is not in the immediate proximity to an ignition source or non-insulated reactor coolant piping, (4) oil collection is provided for in all areas where leakage from pressurized and non pressurized leak sites exist in the oil system, (5) the oil collection system has been seismically qualified to prevent oil spillage reaching areas which may be above the flash point of the lubricating oil, and (6) upper and lower oil pots have been modified with a shield to catch oil and carry it to a tank to reduce fire potential,
| |
| * maintains the SMs of the licensee's analyses based on the licensee's statements that (1) the RCPs are not required to achieve and maintain SSD following a fire in the vicinity and (2) the method does not change the assumptions and limitations of the analytical methods used in the development of the Fire PRA,
| |
| * maintains fire protection DID since automatic and manual fire suppression functions, fire protection for systems and structures, and post-fire SSD capability are not compromised, and
| |
| * will have no effect on the radiological release performance measures (goals, objectives, and performance criteria) since there will be no impact on fire suppression activities.
| |
| In accordance with 10 CFR 50.48(c)(2)(vii), the NRC staff finds the proposed PB method acceptable for application at ONS in lieu of the corresponding NFPA 805, Section 3.3.12(1),
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| requirement because it satisfies the performance goals, performance objectives, and OFFICIAL USE O~JLY SECURITY RELJ\TED INFORMATIO~J
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 41 performance criteria specified in NFPA 805 related to nuclear safety and radiological release, maintains sufficient SM, and maintains adequate fire protection DID.
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| 3.1.3.6 Fire Pump Circulation Relief Valves In LAR Attachment L, "NFPA 805, Chapter 3, Requirements for Approval," the licensee requested the NRC staff's review and approval of a PB method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805, Section 3.5.3 regarding fire pump design and installation compliance with NFPA 20, Section 5.11. The licensee has requested approval of a PB method to justify the omission of circulation relief valves on the HPSW and Keowee fire pumps.
| |
| NFPA 805 has incorporated the requirement for circulation relief valves as specified in NFPA 20, Section 5.11.1.2, "The valve shall provide flow of sufficient water to prevent the pump from overheating when operating with no discharge."
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| The HPSW pumps are not standard fire pumps but large industrial pumps and subsequently built to different original standards. The HPSW pumps are utilized as the ONS fire pumps.
| |
| These pumps have a dual function: to supply water for fire suppression and to provide sealing/cooling water to various components. The two electric-driven HPSW pumps are provided with cooling lines designed to cool the pump motor and also provide some flow to prevent the pump from overheating.
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| The HPSW pumps do not generally operate without flow. When the HPSW pump operates in a fire event, the ONS fire brigade response procedure instructs that a deluge system or hydrant be opened if the flow on the system is assessed less than 1,450 gallons per minute (gpm) in order to maintain greater than the manufacturer's minimum flow.
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| The Keowee fire pump is not provided with an automatic relief valve. However, the Keowee pump automatically shuts down when flow stops such that it will not run at shutoff pressure.
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| The licensee stated that the basis for the approval request of this PB method is:
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| * The HPSW pumps have procedures in place to ensure there is acceptable flow to prevent overheating therefore circulation relief valves are not necessary.
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| * The Keowee fire pump has an automatic shutdown feature to prevent overheating therefore circulation relief valve is not necessary.
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| The licensee stated that omission of circulation relief valves on the HPSW and Keowee fire pumps does not impact the nuclear safety performance criteria because the pumps are operable and measures are in place to ensure the pumps do not overheat.
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| By letter dated September 13, 2010, (Reference 12), the licensee stated that the proposed PB method maintains SMs and conservatisms because (1) alternative measures are provided to ensure the HPSW and Keowee fire pumps will not overheat and (2) the PB method does not change the assumptions and limitations of the analytical methods used in the development of the Fire PRA.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY REL,lWED INFORMATION 42 By letter dated September 13, 2010, (Reference 12), the licensee stated that the omission of circulation relief valves on the HPSW and Keowee fire pumps does not impact fire protection DID because (1) procedures ensure the HPSW pumps do not overheat by manually opening an excess flow path, (2) the Keowee pump is maintained free of overheating by the auto-stop feature, and (3) the lack of circulation relief valves does not result in compromising automatic or manual fire suppression functions, fire protection for systems and structures, or post-fire SSD capability since the pumps are functional and measures are in place to ensure the pumps do not overheat.
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| By letter dated September 13, 2010, (Reference 12), the licensee stated that the omission of circulation relief valves on the HPSW and Keowee fire pumps has no impact on the radiological release performance criteria because (1) the features of the fire pumps do not change the radiological release evaluation performed for each fire zone that potentially contaminated water is contained and smoke is monitored since fire brigade control of water runoff and smoke is not hindered because of the lack of circulation relief valves, and (2) the fire pumps provide radiological clean water to the HPSW system and do not cross-tie to contaminated water piping.
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| The NRC staff finds that the proposed PB method:
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| * does not impact the nuclear safety performance measures (goals, objectives and performance criteria because the (1) Keowee pump is designed using shutoff interlocks to prevent pump damage, (2) the HPSW pumps are procedurally controlled during operations to ensure the pumps do not overheat, and (3) the HPSW pumps are addressed in the brigade procedures to ensure at least a minimum flow is maintained,
| |
| * maintains the SMs of the licensee's analyses based on the licensee's statements that (1) alternative measures are provided to ensure the HPSW and Keowee fire pumps will not overheat, and (2) the method does not change the assumptions and limitations of the analytical methods used in the development of the Fire PRA,
| |
| * maintains fire protection DID since (1) procedures ensure the HPSW pumps do not overheat by manually opening an excess flow path, (2) the Keowee pump is maintained free of overheating by the auto-stop feature, and (3) automatic and manual fire suppression functions, fire protection for systems and structures, and post-fire SSD capability are not compromised, and
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| * will have no effect on the radiological release performance measures (goals, objectives, and performance criteria) since there will be no impact on fire suppression activities.
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| In accordance with 10 CFR 50.48(c)(2)(vii), the NRC staff finds the proposed PB method acceptable for application at ONS in lieu of the corresponding NFPA 805, Section 3.5.3 requirement because it satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release, maintains sufficient SM, and maintains adequate fire protection DID.
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| 3.1.3.7 Insufficient Pressure for Reactor Building Hose Stations In LAR Attachment L, "NFPA 805, Chapter 3, Requirements for Approval," the licensee requested the NRC staff's review and approval of a PB method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805, Sections 3.5.3, 3.5.16, 3.6.1, and 3.6.2 regarding fire protection water supply to the RBs. The licensee has requested approval of a PB OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 43 method to justify the existing ONS fire pumps, standpipes, and water mains for the RBs that do not meet certain aspects of NFPA 20, "Standard for the Installation of Stationary Pumps for Fire Protection." Approval is requested for the use of the low-pressure service water (LPSW) system to supply the RB hose stations/standpipes at less than the required pressure for the RB hose stations/standpipes.
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| The pressure at various standpipe elevations does not meet the minimum pressure prescribed by NFPA 14, "Standard for the Installation of Standpipe and Hose Stations." The licensee stated that the RB hose stations cannot meet the system demands of pressure and flow as required for fire fighting with hose stations in the RBs.
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| The hydraulic calculations indicate that a flow of 100 gpm, provides the residual pressure of approximately 21 pounds per square inch (psi) at the highest elevation inside the RB and 56 psi in the lower elevations. Both of these values are less than the required pressure of 65 psi (Code of Record: NFPA 14 - 1978).
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| The licensee stated that the basis for the approval request of this PB method is:
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| * Quick detection and suppression of a fire by the fire brigade is generally an inherent assumption in the Fire PRA, but in the case of a fire in the RB, no credit for manual suppression is given in the containment Fire PRA model.
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| * The licensee committed to the NRC to use the LPSW to supply the hose stations/standpipes in the RBs. The LPSW pumps were never designed to be able to provide the required pressures for the hose stations in RB.
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| * The fire hazards in the RBs are minimized and higher pressure for hose station operations are provided at the lower elevations where there is a higher concentration of combustibles.
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| * The ONS fire brigade has low-pressure nozzles available and is trained on their use.
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| * There are six additional carbon dioxide fire extinguishers staged at each RB personnel hatch.
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| * The primary purpose of the hose stations in containment is to act as back-up manual suppression during non-power operation (NPO) modes.
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| By letter dated November 19, 2010 (Reference 52), the licensee stated that during power operations, the expected response to a fire in containment is not to enter the RB and let the fire burn out either via fuel consumption or lack of oxygen. The fire brigade is trained and standard operating procedures direct them to preferably only enter an area to fight a fire with a charged hose line. The charged hose line would be connected to the HPSW system via hose stations in the AB with an alternative connection supplied from a yard fire hydrant. The HPSW system is the normal plant fire protection water system. There is a hose station located in the immediate area adjacent to each personnel hatch in the AB. During NPO modes a fire could be attacked using the existing RB hose stations (if at the fire brigade's discretion that the fire is within the capabilities of the LPSW hose stations), or with a hose line connected to the HPSW system as described above, or with a hose line connected to the HPSW system through the yard hydrants if the exterior equipment hatch is open.
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| The licensee stated that the fire brigade will develop a Standard Operating Guide (SOG) for fighting a fire in the RB. Training is already performed on tactics for fighting fires of this nature OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 44 but training will be reinforced with a new SOG. The Fire Brigade Administrator will review the Fire Plans to determine if enhancement is necessary. This item is being tracked by the ONS Corrective Action Program and is an implementation item (Section 2.9, Table 2.9-1, Item 36).
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| By letters dated September 13, 2010 and November 19, 2010, (References 12 and 52) the licensee stated that low pressure in the RB hose stations does not impact the nuclear safety performance criteria because (1) in the case of a fire in the ONS RB, no credit for manual suppression is given in the containment Fire PRA model, and (2) the RBs are not accessed during power operation unless in an emergency. The fire brigade does not use the hose stations located in containment in the event of a significant fire in the RB.
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| By letter dated November 19, 2010, (Reference 52), the licensee stated that the proposed PB method maintains SMs and conservatisms because (1) the low pressure in the RB hose stations are provided for limited use by trained fire brigade members, (2) alternative equipment such as fire extinguishers and charged hose station from outside the RB can be used to fight a fire, and (3) the PB method does not change the assumptions and limitations of the analytical methods used in the development of the Fire PRA.
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| By letter dated September 13, 2010, (Reference 12), the licensee stated that the RB hose stations having insufficient pressure do not impact fire protection DID because (1) administrative controls to prevent fires are still in place, (2) the availability of alternate suppression means for the RB does not result in compromising automatic or manual fire suppression functions, and (3) post-fire SSD capability remains unaffected because no credit is provided for these hose stations.
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| By letters dated September 13, 2010 and November 19, 2010, (Reference 12 and 52) the licensee stated the low-pressure hose stations in the RBs has no impact on the radiological release performance criteria because (1) the entire RB in which the subject fire hose stations are located is in an environmentally sealed radiological area, and (2) the limited pressure of fire hose stations in the RBs has no impact on the radiological release performance criteria.
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| The NRC staff finds that the proposed PB method:
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| * does not impact the nuclear safety performance measures (goals, objectives and performance criteria because (1) the containment portion of the Fire PRA does not take credit for manual suppression, (2) the analysis assumes the hose stations are not available or immediately accessible during power operations, (3) there is no increased risk or change in delta CDF or LERF as manual suppression is not credited in the containment Fire PRA, and (4) the RB FREs indicate that based on the containment configuration, limited exposed combustibles, volume of the containment, and slow or limited fire propagation will reduce the impact of fire in the containment,
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| * maintains the SMs of the licensee's analyses based on the licensee's statements that the method does not change the assumptions and limitations of the analytical methods used in the development of the Fire PRA,
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| * maintains fire protection DID since automatic and manual fire suppression functions, fire protection for systems and structures, and post-fire SSD capability are not compromised,
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| * will have no effect on the radiological release performance measures (goals, objectives, and performance criteria) since there will be no impact on fire suppression activities.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 45 In accordance with 10 CFR 50.48(c)(2)(vii), the NRC staff finds the proposed PB method acceptable for application at ONS in lieu of the corresponding NFPA 805, Sections 3.5.3, 3.5.16, 3.6.1, and 3.6.2 requirements because it satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release, maintains sufficient SM, and maintains adequate fire protection DID.
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| 3.1.3.8 Fire Pump Automatic/Remote Stop Feature In LAR Attachment L, "NFPA 805, Chapter 3, Requirements for Approval," the licensee requested the NRC staff's review and approval of a PB method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805, Section 3.5.6 regarding automatic stop features on fire pumps, and Section 3.5.3 regarding NFPA 20 requirements for fire pumps. The licensee has requested approval of a PB method to justify the automatic stop function on existing ONS HPSW pumps (fire water) and the Keowee fire pump, and to justify remote operation of the HPSW pumps from the control room. Both of these features are contrary to the NFPA 20, "Standard for the Installation of Stationary Pumps for Fire Protection."
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| As described by the licensee:
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| * The HPSW pumps can be stopped automatically by the level switches in the elevated water storage tank (EWST), manually in the control room, locally at the switchgear breaker, and locally at the pump.
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| * The HPSW system has a jockey pump to maintain normal system pressure during service water (SW) loads. If the pressure falls below the setpoint at which the jockey pump cannot maintain the HPSW system, the altitude valve, located at the base of the EWST, opens to supply system pressure and flow. The HPSW pump(s) stop based upon set fill level of the EWST.
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| * A startlrun/off/base/standby switch is provided in the control room on auxiliary control board, 1AB3, for the HPSW pumps (both A and B). This permits the pumps to be manually operated in order to avoid pressure disruptions in the system.
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| * The Keowee fire pump stops automatically based on sensing low/no water flow.
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| The licensee stated that the basis for the approval request of this PB method is:
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| * The NRC previously accepted the use of the HPSW system for fire protection use.
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| * The Keowee fire pump has an automatic shut off on low/no flow. The system is routinely tested to demonstrate operability.
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| * When the pumps are operating, they are monitored by trained operators who can control the pumps as necessary.
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| The licensee stated that remotely/automatically stopping the fire pumps does not impact the nuclear safety performance criteria because (1) nuclear safety is not affected, and (2) the pumps are available and monitored by trained operators.
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| By letter dated September 13, 2010, (Reference 12), the licensee stated that the proposed PB method maintains SMs and conservatisms because (1) the fire pumps operate automatically OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 46 and are monitored and controlled by trained operators, and (2) the PB method does not change the assumptions and limitations of the analytical methods used in the development of the Fire PRA.
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| By letter dated September 13, 2010, (Reference 12), the licensee stated that the means of remotely/automatically stopping the fire pumps does not impact fire protection DID because (1) pump controls are maintained by procedures to ensure pumps are available by taking manual control, (2) suppression is maintained by the availability of a redundant fire pump, and (3) means are available to ensure fire pumps are functional during a fire event, which does not result in compromising automatic or manual fire suppression functions, fire protection for systems and structures, or post-fire SSD capability.
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| By letter dated September 13, 2010, (Reference 12), the licensee stated that remotely/automatically stopping the fire pumps has no impact on the radiological release performance criteria because (1) the features of the fire pumps do not change the radiological release evaluation performed for each fire zone that potentially contaminated water is contained and smoke is monitored since fire brigade control of water runoff and smoke is not hindered because of the lack of auto-stop feature, and (2) the fire pumps provide radiologically clean water to the HPSW system and do not cross-tie to contaminated water piping.
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| The NRC staff finds that the proposed PB method:
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| * does not impact the nuclear safety performance measures (goals, objectives and performance criteria) because (1) the pumps are available and monitored by trained operators under the controls of procedures, (2) fire brigade operations recognize proper fire water supply conditions, and (3) monitoring programs will be in place to ensure proper operation of the fire water supply pumps,
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| * maintains the SMs of the licensee's analyses based on the licensee's statements that the method does not change the assumptions and limitations of the analytical methods used in the development of the Fire PRA,
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| * maintains fire protection DID since automatic and manual 'fire suppression functions, fire protection for systems and structures, and post-fire SSD capability are not compromised, and
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| * will have no effect on the radiological release performance measures (goals, objectives, and performance criteria) since there will be no impact on fire suppression activities.
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| In accordance with 10 CFR 50.48(c)(2)(vii), the NRC staff finds the proposed PB method acceptable for application at ONS in lieu of the corresponding NFPA 805, Sections 3.5.6 and 3.5.3 requirements because it satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release, maintains sufficient SM, and maintains adequate fire protection DID.
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| 3.1.3.9 KHS Fire Main and Standpipe Use In LAR Attachment L, "NFPA 805, Chapter 3, Requirements for Approval," the licensee requested the NRC staff's review and approval of a PB method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805, Sections 3.5.7, 3.5.10, and 3.5.15 regarding yard loop connections, yard loop design, and hydrant connections. The licensee has OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORM.<\TION 47 requested approval of a PB method to justify the omission of the requirement for a fire main loop and fire hydrants at the KHS .
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| KHS does not have a fire main loop or fire hydrants. However, KHS has two outside fire hose standpipes that use fire hydrant appurtenances as their controlling devices. The outside fire hose standpipes (fire hydrants) are equipped with two 2-1/2" hose connections (no other/larger connections).
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| The KHS underground fire piping consists of an 8-inch pipe that supplies the transformer water spray system and a 4-inch pipe, which tee's and supplies the two-yard hydrants/outside fire hose standpipes. There is no yard fire loop in accordance with the elements on NFPA 805 Sections 3.5.7,3.5.10, and 3.5.15. In addition, in accordance with NFPA 24, Sections 5.2.1 and 13.1, the piping servicing the fire hydrants is not provided with piping greater than 6 inches in diameter.
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| These devices do not meet the requirements for fire hydrants as they are supplied via a 4-inch underground main. It can be best determined that the two locations with fire hydrants' appurtenances are used as external fire hose standpipes because the fire hydrant offers a drain function of the barrel therefore no freeze protection is required.
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| Revision of the Design Basis Specification for Fire Protection to state that the KHS fire hydrants are not designed, nor intended to function as fire hydrants but to act as external automatic wet standpipes for fire brigade/fire department response as required is an implementation item (SE Section 2.9, Table 2.9-1, Item 26).
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| The Jicensee stated that the basis for the approval request of this PB method is:
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| * A yard fire loop is not required given the fire protection water required at the KHS .
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| * The fire hydrants are installed as fire hose standpipes for the fire brigade and act as a wet standpipe.
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| The licensee stated that the layout of the fire service main at the KHS does not impact the nuclear safety performance criteria because (1) the Keowee fire protection system is not required for the overall nuclear fire safety at ONS, (2) a fire is not simultaneously postulated at ONS and KHS and (3) KHS is the emergency power for ONS, such that in the event of a loss of-offsite power, KHS provides the power to shutdown ONS, while if KHS were unavailable, the licensee would proceed on a controlled shutdown using normal power.
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| By letter dated September 13, 2010, (Reference 12), the licensee stated that the proposed PB method maintains SMs and conservatisms because (1) the layout of the fire service main at KHS does not impact fire protection and the fire hydrants are functionally automatic wet standpipes for fire brigade/fire department operations, and (2) the PB method does not change the assumptions and limitations of the analytical methods used in the development of the Fire PRA.
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| By letter dated September 13, 2010, (Reference 12), the licensee stated that the fire service main at KHS does not impact fire protection DID because (1) administrative controls to prevent fires are not affected, (2) suppression is maintained by the inherent design and objectives of the Keowee fire service main, and (3) the presence of this fire service main does not compromise OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 48 automatic or manual fire suppression functions, fire protection for systems and structures, or post-fire SSD capability.
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| By letter dated September 13, 2010, (Reference 12), the licensee stated that the layout of the fire service main at KHS has no impact on the radiological release performance criteria because there are no radiological concerns at the KHS plant location and therefore no capability to influence a potential radiological release.
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| The NRC staff finds that the proposed PB method:
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| * does not impact the nuclear safety performance measures (goals, objectives and performance criteria) because (1) KHS is not required for the overall nuclear fire safety at ONS, (2) the fire service main layout at KHS adequately provides fire suppression water to the limited demands of the hydrant standpipes and transformer deluge systems, (3) fire in these locations do not affect the nuclear safety performance of the power block, and (4)
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| KHS is the emergency power for ONS, such that in the event of a loss-of-offsite power, KHS provides the power to shutdown ONS, while if KHS were unavailable, the licensee would proceed on a controlled shutdown using normal power,
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| * maintains the SMs of the licensee's analyses based on the licensee's statements that the method does not change the assumptions and limitations of the analytical methods used in the development of the Fire PRA,
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| * maintains fire protection DID since (1) administrative controls to prevent fires are not affected, and (2) automatic and manual fire suppression functions, fire protection for systems and structures, and post-fire SSD capability are not compromised, and
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| * will have no effect on the radiological release performance measures (goals, objectives, and performance criteria) because the layout of the fire service main at Keowee has no impact on fire suppression activities.
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| In accordance with 10 CFR 50.48(c)(2)(vii), the NRC staff finds the proposed PB method acceptable for application at ONS in lieu of the corresponding NFPA 805, Sections 3.5.7, 3.5.10, and 3.5.15 requirements because it satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release, maintains sufficient SM, and maintains adequate fire protection DID.
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| 3.1.3.10 Use of Dual Purpose Fire Protection Water Supplies In LAR Attachment L, "NFPA 805, Chapter 3, Requirements for Approval," the licensee requested the NRC staff's review and approval of a PB method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805, Section 3.5.16 regarding dedicated fire protection water supply system. The licensee has requested approval of a PB method to justify the use of the ONS and Keowee service water (SW) systems for purposes other than fire protection water supply. The HPSW system is used for dual purposes including fire protection (suppression systems, hose stations, and fire hydrants) and SW uses including supplying bearing lubrication or cooling water to the condenser circulation water pumps and motors, the primary instrument air compressor, the leak rate test compressors, and backup cooling water to the turbine driven emergency feedwater (EFW) pump oil coolers and the high-pressure injection (HPI) pump motors. In addition, the hydrants and/or hose stations may be used for other functions such as wash down and truck/tank filling at ONS.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 49 As described by the licensee, this usage of yard hydrants and standpipes would require control room notification of fire protection system water for plant evolutions other than fire protection under the following conditions: (1) the Work Control Center Senior Reactor Operator is notified of the evolution, (2) fire brigade procedures provide for steps to secure non-essential use of the fire water system immediately if a fire occurs, and (3) the HPSW pump capacity far exceeds the largest fire suppression water demand.
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| The licensee stated that the use of fire protection water for these non-fire protection system water demands would have no adverse impact on the ability of the fire protection water supply system to provide required flow and pressure based on two redundant 6,000 gpm HPSW pumps. The largest suppression system demand is the Unit 2 TB mezzanine system, which requires 2,723 gpm, plus 1,000 gpm fire hose allowance and 318 gpm additional SW for a total demand of 4,041 gpm. Assuming the maximum of 500 gpm non-fire-related system flow, there is still approximately 1,500 gpm of margin, with just one pump in operation, in excess of the required HPSW system demands. The licensee concluded that neither the flow and pressure available to any automatic water based suppression system, nor the manual fire suppression demands when needed, will be adversely impacted by the proposed change since the non-fire protection water demand would be secured before hose streams were used.
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| The Keowee SW system is used for dual purposes includil1g fire protection (suppression systems and hose stations) and SW uses including dilution flow, supplying cooling to air compressors and heating ventilation air conditioning (HVAC) units, and for tank usage. In addition, the hose stations may be used for other functions such as wash down and truck/tank filling. The SW demands are generally taken before the fire pump. The largest SW demand is the dilution flow line that has a valve that automatically closes upon actuation of the fire pump to allow sufficient flow to the fire pump.
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| The licensee stated that the basis for the approval request of this PB method is:
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| * The HPSW system has excess capacity.
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| * The Keowee SW system has an automatic valve to cease high SW flow demands in the event of the fire pump start.
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| * Appropriate personnel are notified when using the HPSW system.
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| * Fire brigade response procedure includes a step to make a public address announcement to discontinue use of the HPSW system for non-essential purposes.
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| The licensee stated that the use of the HPSW and Keowee SW systems for non-fire protection uses does not impact the nuclear safety performance criteria because (1) the HPSW system has excess capacity to supply the demands of the HPSW system above the greatest sprinkler system demand, and (2) the Keowee SW system has a valve that automatically closes upon actuation of the fire pump.
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| By letter dated September 13, 2010, (Reference 12), the licensee stated that the proposed PB method maintains SMs and conservatisms because (1) the HPSW system has excess capacity to supply the demands of the HPSW system above the greatest sprinkler system demand, (2) the Keowee SW system has a valve that automatically closes upon actuation of the fire pump, and (3) the PB method does not change the assumptions and limitations of the analytical methods used in the development of the Fire PRA.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 50 By letter dated September 13, 2010, (Reference 12), the licensee stated that use of the HPSW system for non-fire protection uses does not impact fire protection DID because (1) the Work Control Center Senior Reactor Operator is notified of the evolution and fire brigade procedures provide for steps to secure non-essential use of the fire water system immediately if a fire occurs, (2) suppression is maintained by excess capacity, operational guidance, and automatic equipment functions to maintain sufficient fire fighting water, and (3) the use of the HPSW pumps and Keowee, SW system do not compromise automatic or manual fire suppression functions, fire protection for systems and structures, or post-fire SSD capability.
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| By letter dated September 13, 2010, (Reference 12), the licensee stated that the use of the HPSW system for non-fire protection uses, including the use of hydrants and hose for purposes other than fire, has no impact on the radiological release performance criteria because (1) there are no radiological hazards at Keowee and (2) the HPSW system is radiologically clean and does not cross-tie to contaminated water piping.
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| The NRC staff finds that the proposed PB method:
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| * does not impact the nuclear safety performance measures (goals, objectives and performance criteria) because (1) the licensee's statements that the HPSW has excess capacity to supply the demands of the greatest sprinkler system demand, (2) the Keowee SW system has an automatic valve to cease high SW flow if the fire pump starts, and (3) plant-wide fire notification measures to stop non-essential water use are in place,
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| * maintains the SMs of the licensee's analyses based on the licensee's statements that (1) that the HPSW system has excess capacity to supply the demands of the HPSW system above the greatest sprinkler system demand, (2) the Keowee SW system has a valve that automatically closes upon actuation of the fire pump, and (3) the method does not change the assumptions and limitations of the analytical methods used in the development of the Fire PRA,
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| * maintains fire protection DID since (1) the HPSW pumps have the excess capacity to supply the demands of the HPSW system in addition to the greatest sprinkler system demand, and (2) both systems automatic and manual fire suppression functions, fire protection for systems and structures, and post-fire SSD capability are not compromised, and
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| * will have no effect on the radiological release performance measures (goals, objectives, and performance criteria) since there will be no impact on suppression activities inside the radiation-controlled area (RCA).
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| In accordance with 10 CFR 50.48(c)(2)(vii), the NRC staff finds the proposed PB method acceptable for application at ONS and Keowee in lieu of the corresponding NFPA 805, Section 3.5.16 requirement because it satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release, maintains sufficient SM, and maintains adequate fire protection DID.
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| 3.1.3.11 KHS Fire Protection Fire Pump By letter dated November 19, 2010, (Reference 52) the licensee requested the NRC staff's review and approval of a PB method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805, Sections 3.5.3, and 3.5.4 regarding the Keowee fire pump. The OFFICIAL USE ONLY SECURITY RELJ\TED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 51 licensee has requested approval of a PB method to justify the omission for the installation of a second fire pump at Keowee.
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| Keowee is provided with one electric motor-driven fire pump. There is no secondary/back-up fire pump. The existing pump is installed in accordance with NFPA 20 and is capable of providing the required flow and pressure to the Keowee hose stations. Keowee is spatially separated from any other Oconee power block areas by approximately 3,000 feet and is an extension of the Oconee "power block" as Keowee is used as emergency power.
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| Keowee is a credited SSD system; however a fire at Keowee does not impact the SSD capability at Oconee using other credited power systems. A fire at Oconee is not postulated concurrent with a fire at Keowee; therefore, a fire at Oconee does not impact the ability of Keowee to provide emergency power.
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| The licensee stated that the basis for the approval request of this PB method is:
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| * A fire at Keowee does not impact the SSD capability at ONS.
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| * A fire at ONS is not concurrent with a fire at Keowee.
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| * Keowee is a separate structure located a significant distance away from any other Oconee power block structures.
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| * The main purpose of the fire pump at Keowee is to supply the Keowee fire hose stations.
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| * Compensatory measures are provided in the event the Keowee fire pump is out of service.
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| The NRC staff finds that the proposed PB method:
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| * does not impact the nuclear safety performance measures (goals, objectives and performance criteria) because (1) a fire is not simultaneously postulated at Oconee and Keowee (2) Keowee is the emergency power source for Oconee. If a fire at Keowee were to render it unavailable, Oconee would proceed to shutdown using normal power, and (3) a fire at Keowee does not impact the SSD capability at ONS.
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| * maintains the SMs of the licensee's analyses based on the licensee's statements that; (1) the single electric fire motor-driven pump at Keowee does not impact "SSD" fire protection for the "power block" or power production areas of the turbine/auxiliary/reactor buildings or the SSF, and (2) the fire pump is used to supply the flow and pressure requirements to the Keowee fire hose stations and has been evaluated in accordance with NFPA 20.
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| * maintains fire protection DID since (1) a fire at Keowee does not impact the SSD capability of ONS, and (2) ONS has compensatory measures in place in the event the Keowee fire pump is out of service.
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| * will have no effect on the radiological release performance measures (goals, objectives, and performance criteria) since there will be no impact on suppression activities inside the RCA.
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| O~~ICIAL USE ONLY SECURITY RELATED IN~ORMATIO~J
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| OFFICIAL USE O~JLY SECURITY RELATED INFORMATION 52 In accordance with 10 CFR 50.48(c)(2)(vii), the NRC staff finds the proposed PB method acceptable for application at Keowee in lieu of the corresponding NFPA 805, Sections 3.5.3 and 3.5.4 requirements because it satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release, maintains sufficient SM, and maintains adequate fire protection DID.
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| 3.1.4. Conclusion for Section 3.1 The NRC staff reviewed ONS's RI/PB FPP for compliance with each of the requirements of NFPA 805, Chapter 3, 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 finds the ONS RI/PB FPP acceptable with respect to the fundamental FPP elements and minimum design requirements of NFPA 805, Chapter 3, as modified by 10 CFR 50.48(c)(2),
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| because the licensee accomplished the following:
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| * Used an overall process consistent with NRC staff-approved guidance to determine the state of compliance with each of the applicable NFPA 805, Chapter 3, requirements.
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| * Provided appropriate documentation of ONS's state of compliance with the NFPA 805, Chapter 3, requirements, which adequately demonstrated compliance in that the licensee was able to substantiate that it complied:
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| With the requirement directly.
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| With the intent of the requirement (or element) given adequate justification.
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| Via previous NRC staff approval of an alternative to the requirement.
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| Through the use of an EEEE.
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| Through the use of a combination of the above methods.
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| * Used PB methods that the NRC staff has specifically approved in accordance with 10 CFR 50.48(c)(2)(vii).
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| 3.2. Nuclear Safety Capability Assessment (NSCA) Methods NFPA 805 is a PB fire protection standard that allows engineering analyses to be used to show that FPP features and systems provide sufficient capability to meet the regulatory requirements.
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| Specifically, Section 2.4, "Engineering Analyses," states the following:
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| Engineering analysis is an acceptable means of evaluating a FPP 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.
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| NFPA 805, Chapter 1, defines the nuclear safety goal, objectives, and performance criteria that the FPP must meet as follows:
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 53 Nuclear Safety Goal 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.
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| Nuclear Safety Objectives In the event of a fire during any operational mode and plant configuration, the plant shall be as follows:
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| (1) Reactivity Control. Capable of rapidly achieving and maintaining subcritical conditions.
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| (2) Fuel Cooling. Capable of achieving and maintaining decay heat removal (DHR) and inventory control functions.
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| (3) Fission Product Boundary. Capable of preventing fuel clad damage so that the primary containment boundary is not challenged.
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| Nuclear Safety Performance Criteria 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.
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| (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.
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| (b) Inventory and Pressure Control. With fuel in the reactor vessel, head on and tensioned, inventory and pressure control shall be capable of controlling coolant level such that subcooling is maintained for a [pressurized water reactor] (PWR) and shall be capable of maintaining or rapidly restoring reactor water level above top of active fuel for a [boiling water reactor] (BWR) such that fuel clad damage as a result of a fire is prevented.
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| (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.
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| (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.
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| (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.
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| OFFICIAL USE O~JLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 54 3.2.1. Compliance with Nuclear Safety Capability Assessment Methods NFPA 805, Section 2.4.2, "Nuclear Safety Capability Assessment," states the following:
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| The purpose of this section is to define the methodology for performing a nuclear safety capability assessment. The following steps shall be performed:
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| (1) Selection of systems and equipment, and their inter-relationships necessary to achieve the nuclear safety performance criteria in Chapter 1.
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| (2) Selection of cables necessary to achieve the nuclear safety performance criteria in Chapter 1.
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| (3) Identification of the location of nuclear safety equipment and cables.
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| (4) Assessment of the ability to achieve the nuclear safety performance criteria given a fire in each fire area.
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| This section of the SE evaluates the conformance of licensee's methodology to the first three of the above-listed topics. SE Section 3.5 addresses the assessment of the fourth topic.
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| RG 1.205, Revision 1 (Reference 14) endorses with exceptions and clarifications, NEI 04-02, Revision 2 (Reference 21). In addition, when applied in conjunction with RG 1.205, Chapter 3 of industry guidance document NEI 00-01, "Guidance for Post-Fire SSD Circuit Analysis" Revision 1, (Reference 56) and Revision 2 (Reference 28) provides an acceptable deterministic methodology. This NRC-endorsed method documents in a table format (Le., NEI 04-02, Table B-2, "NFPA 805 Chapter 2 - Nuclear Safety Transition - Methodology Review") the licensee's comparison of its post-fire SSD analyses from its existing fire protection licensing basis to the guidance in NEI-00-01, Chapter 3. The NRC staff reviewed LAR Section 4.2.1, "Nuclear Safety Capability Assessment Methodology Review," and Attachment B, "NEI 04-02 Table B NSCA Methodology Review," against these guidelines.
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| The licensee states that NSCA were performed on a fire area basis. Once the systems needed to achieve and maintain safe and stable conditions in accordance with NFPA 805 were identified, components needed to ensure the capability of these systems to achieve the nuclear safety performance criteria of NFPA 805 were identified and a comprehensive list of equipment, referred to by the licensee as the SSEL was developed. In addition to components required to achieve the nuclear safety performance criteria of NFPA 805, the SSD equipment list (SSEL) includes components in system flow paths that require operation or repositioning to allow the system to function, and components that could spuriously operate and impair SSD. Based on the cables and components present in the fire area of concern, an assured success path is determined. The fire area analysis methodology considers the occurrence of multiple fire induced failures and multiple spurious actuations.
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| For the majority of the NEI 00-01 attributes listed in LAR Attachment B, the licensee stated that the approach used to conduct the post-fire SSD analyses aligns with the NEI 00-01, Revision 1 (Reference 56) guidance. However, there were several attributes for which the licensee stated that they either align with the intent of the NEI 00-01 guidance or do not align with the NEI 00-01 guidance. Table 3.2-1, "Nuclear Safety Capability Assessment Method Review," in Attachment B to this SE, identifies each applicable NEI 00-01 guidance section, documents whether the licensee stated that it met the NEI 00-01 guidance, or provided justification for meeting the OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 55 intent of that guidance or not meeting that guidance, and presents the NRC staffs evaluation of the acceptability of the licensee's justification.
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| The NEI 00-01 guidance, as endorsed in RG 1.205, is only one acceptable means to demonstrate compliance. Therefore, the NRC staff reviewed the instances where the licensee deviated from NFPA 805 standard. The NRC staff determined that, in all cases in which the licensee stated to have met the intent of the NEI 00-01 guidance, the alternative methodology used by the licensee was an acceptable means to meet the NFPA 805 requirement. For instance, Section 3.1.1.4 of NEI 00-01 states that cables and equipment required for alternative shutdown must be independent of the fire area of concern. The licensee aligns with the intent of this guidance through the transfer of control to the SSF, which isolates required systems and equipment from the effects of a fire for the fire areas of concern. Following transfer of control to the SSF, the equipment credited for an SSF shutdown meets the intent of the guidance provided in NEI 00-01. Also, although Section 3.1.2.2 of NEI 00-01 states that RCS pressure should be controlled by controlling the rate of charging/makeup to the RCS, the licensee states that pressure control may be accomplished utilizing reactor makeup from the SSF or injection from HPI in conjunction with pressurizer heaters, safety relief valves, pressure operated relief valve (s) (paRV's), RCS loop high point vent valves, or reactor head vent valves and controlling decay heat removal rates. Since an assured RCS pressure success path has been determined for each fire area, the intent of the guidance provided in NEI 00-01 is met.
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| Specific cases where the licensee states that it does not align with the NEI 00-01 guidance include the following:
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| : 1. Sections 3.3.1.7,3.3.3.3, and 3.5.2.4 of NEI 00 Common Power Supplies If the upstream protection device (Le., feeder breaker or fuse) of a required power supply is not properly coordinated with downstream (load) protection devices, a fire induced fault (e.g., short to ground) affecting any of the load circuits could cause the upstream (feeder) breaker to trip prior to causing a trip of the individual load protection device. This scenario, which is referred to as the "Common Power Supply Concern," would result in a loss of electrical power to all circuits connected to the affected power supply, including those circuits relied on to achieve the nuclear safety performance criteria.
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| Section 2.4.2.2.2 of NFPA 805 states that circuits whose fire induced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. The evaluation performed to demonstrate conformance to this criterion is referred to as a "Coordination Study." Specific issues to be considered in the study are discussed in Sections 3.3.1.7,3.3.3.3, and 3.5.2.4 of NEI 00-01. In the licensee's October 31,2008 (Reference 4) submittal the licensee stated that it had assumed that electrical protection devices were properly coordinated and did not consider the impact of inadequate breaker coordination when selecting cables. The NRC staff also noted that the licensee's existing coordination study does not include all power supplies required to achieve the nuclear safety objectives of NFPA 805 during power and non-power operations.
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| By letter dated August 3, 2009 (Reference 8), the licensee stated that the existing coordination study currently relied on in the LAR would require further enhancement to meet Section 2.4.2.2.2 of NFPA 805 and that a revised breaker coordination study was underway. By letter dated September 13, 2010 (Reference 12) the licensee stated that the revised breaker coordination study had been completed and identified modifications to four breakers that have an overall risk increase due to their lack of coordination with the upstream protective device.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 56 The four breakers are being modified to maintain the Fire PRA risk profile reported in the LAR (see SE Section 3.4 a more detailed discussion). The plant modifications are described in SE Section 2.8.1 and the licensee stated that appropriate compensatory actions will be implemented until the item is fully resolved.
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| In the cited supplementary information, the licensee also states that the revised coordination study meets the requirements of NFPA 80S, Section 2.4.2.2.2(a), for circuits that share a common power supply with circuits required to achieve the nuclear safety performance criteria.
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| The results of the coordination study will be documented in the NSCA, NPO Pinch Point Analysis, and the Fire PRA. Incorporating all related non-coordinated information in the NSCA and NPO Pinch Point Analysis, and updating the Fire PRA model, to include the results of the breaker coordination study is an implementation item (SE Section 2.9, Table 2.9-1, Item 33).
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| Updating the breaker coordination study to include all new NFPA 805 SSEL-related power supplies (i.e., PSW) for power and non-power operations, and defining additional plant modification if necessary to ensure that the assumptions of the Fire PRA and NSCA remain valid, is an implementation item (SE Section 2.9, Table 2.9-1, Item 44).
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| Based on the information provided in the LAR, as supplemented, the NRC staff finds that the licensee's approach has adequately addressed the issue of inadequate breaker coordination and that the licensee's approach will provide the required electrical protection under NFPA 80S, Section 2.4.2.2.2.(a) upon completion of the identified plant modifications and implementation items.
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| : 2. Section 3.5.1.3 of NEI 00 Timing of Fire-induced Failures For plants transitioning to NFPA 80S, Section 2.2.1 of RG 1.205 states that for cases where the NRC did not specifically approve certain aspects of the plant's current FPP (e.g., through an approved request under 10 CFR 50.12, "Specific Exemptions") licensees can submit elements of their plant's FPP, if they want explicit approval of these elements under 10 CFR 50.48(c).
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| In LAR Attachment T, "Prior-Approval Clarification Request" the licensee requests the staff to document that it had previously approved several assumptions related to the timing of fire damage. Specifically, LAR Attachment T states:
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| "As part of this LAR submittal and transition to NFPA 805, it is requested that the NRC formally document as a "prior approval" recognition that during the 10 minutes required to activate the SSF, fire growth will not have reached a point where fire damage will preclude operator actions from the Control Room nor will any spurious operations or loss of offsite power conditions occur within the first 10 minutes following the identification of a confirmed active fire" (emphasis added).
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| In summary, the licensee requested the NRC staff to document that the NRC previously approved the following limitations of fire damage for a 10 minute period of time needed to activate the SSF:
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| * no damage circuits and controls located in the control room
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| * no spurious operations
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| * no loss of offsite power Section 3.5.1.3 of NEI 00-01 states that circuit failure types resulting in spurious operations should be assumed to exist until action has been taken to isolate the given circuit from the fire OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 57 area. The licensee states it does not align with this guidance because its pre-transition license basis did not assume spurious actuations or hot shorts due to a fire for the first 10 minutes of the event (i.e., no spurious operations or hot shorts are assumed to occur for 10 minutes after confirmation of an active fire).
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| In supplementary information provided in a letter dated November 19, 2010 (Reference 52) the licensee agreed to eliminate the "10 minute free offire damage" assumptions outlined above and to perform an evaluation using NFPA 805 risk-informed processes. Specifically, the licensee states that it will utilize a risk-informed approach to evaluate scenarios that previously relied upon the 10-minute prior approval. This will involve a thorough review of existing analyses to identify new variances. Changes to the FPP, as a result of these variances, will be resolved using the change evaluation process. Upon completion of this activity, all applicable FRE(s) will be updated and compliance will be demonstrated consistent with NFPA 805, Section 4.2.4.2. Completion of these activities is an implementation item (SE Section 2.9, Table 2.9-1, Item 46).
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| For a variance to be supported by the risk-informed process, its risk differential in terms of core damage frequency must be calculated. As part of its response, the licensee provided the estimated LlCDF and LlLERF for three VFDRs it has identified thus far. Specifically, the licensee states that it performed a sensitivity study to estimate the delta risk for four valves, which were excluded from the FRE process due to deterministic application of the 1O-minute assumption.
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| The licensee states that the cumulative delta risk from these potential VFDRs is within the available PSW risk offset margin for all fire areas where the SSF is credited (i.e., AB Fire Area).
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| The licensee states that elimination of the 1O-minute assumption will prohibit ONS from deterministically complying with ONS UFSAR, Section 3.1.11 which requires that following the loss of the control room function, the reactor must be able to be shutdown and maintained in a safe and stable condition.
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| As discussed by the NRC staff in closure memorandum for FAQ 07-0032 (ADAMS Accession No. ML081400292), conformance to 10 CFR 50.48(c) meets or exceeds the requirements of 10 CFR 50.48(a) and GDC 3. Therefore, the NRC staff agrees that conformance to 10 CFR 48(c) also satisfies ONS specific UFSAR Criterion 11 for fire response.
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| As a result of its elimination of the "10 minute free offire damage assumption," the licensing basis of the SSF following a fire will be dictated by the NFPA 805 risk-informed process. As such, the time allowance for performing certain SSF actions during a fire will be established by analyses required to support the risk-informed operation of the SSF.
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| Based on the information provided in the licensee's November 19, 2010 (Reference 52) supplement the NRC staff finds that the licensee's process for eliminating the "10 minute free of fire damage assumption" provides reasonable assurance that the safety objectives of NFPA 805 are satisfied.
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| : 3. Section 3.5.1.5 of NEI 00 Multiple Spurious Operations Section 3.5.1.5 of NEI 00-01 provides guidance for the analysis of multiple spurious actuations (MSOs). The LAR states that it does not align with this guidance because MSOs were not addressed in its pre-transition SSD analysis. The licensee further states that MSOs are resolved through transition to I\IFPA 805.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 58 The PB approach taken by the licensee utilizes FREs in accordance with NFPA 805, Section 4.2.4.2, "Use of Fire Risk Evaluation." To ensure that all potentially significant fire scenarios have been evaluated, this approach requires that potential MSO combination be identified and included. The licensee states that the fire area analysis methodology assumes multiple fire induced failures and multiple spurious actuations, based on the SSD cables and components present in the fire area of concern. All postulated SSD cable and component failures were identified and a resolution provided at the cable or component level for the credited train.
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| The NRC staff was concerned that the MSO analysis was limited to only SSD cables and components. In its October 14, 2010 response (Reference 54), the licensee states that the methodology assumed multiple fire induced failures and multiple spurious actuations, based on the cables and components present in the fire area of concern, and was not limited to SSD cables and components.
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| Based on the information provided in the October 14, 2010 (Reference 54), the NRC staff finds that the licensee's basis for alignment to Section 3.5.1.5 of NEI 00-01 acceptable. SE Section 3.2.3 provides the NRC staff evaluation of the licensee's MSO assessment process.
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| : 4. Section 3.5.2.5 of NEI 00 Common Enclosure Circuits Section 3.5.2.5 of NEI 00-01 provides guidance for addressing the Common Enclosure Associated Circuit concern described in Section 2.4.2.2.2 of NFPA 805. One element of this concern is that fire-induced electrical faults on cables that lack appropriately sized fuses or circuit breakers may cause secondary fires outside of the immediate fire area.
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| In the LAR, the licensee credits its original breaker coordination study to address Common Enclosure concerns. However, as discussed above, the original ONS coordination study does not satisfy applicable NFPA 805 or NEI 00-01 criteria. By letter dated September 13, 2010 (Reference 12) the licensee stated that a revised breaker coordination study had been completed. In Enclosure 3 of the LAR (Reference11) the licensee states that the second phase of the revised coordination study includes a review of the cable damage curves to determine if the electrical circuit design provides proper protection. The results of this review were entered into the ARTRAK database and analyzed in the Fire Area/Fire Zone impacts. All power supplies required by the NSCA, PRA, and NPO, as identified on the associated equipment list, were included in the breaker coordination study scope of "SSD related" power supplies. The licensee further states that the coordination study meets the requirements of NFPA 805, Section 2.4.2.2.2.(b), for circuits that share a common enclosure with circuits required to achieve nuclear safety performance criteria. In addition, the licensee states that a review of recent modifications confirms that adequate electrical circuit protection has been maintained as part of the design change process. In addition, the licensee states that the results of the coordination study will be documented in the NSCA, NPO Pinch Point Analysis, and the Fire PRA.
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| Incorporating all related non-coordinated information in the NSCA and NPO Pinch Point Analysis, and updating the Fire PRA model, to include the results of the breaker coordination study is an implementation item (SE Section 2.9, Table 2.9-1, Item 33).
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| Based on the information provided in the LAR, as supplemented, the NRC staff finds that the licensee's approach has adequately addressed common enclosure associated circuit concern at ONS and that the licensee will provide the required electrical protection under NFPA 805, Section 2.4.2.2.2.(b) upon completion of the identified implementation items.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 59
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| : 5. Section 3.1.1.9 of NEI 00 72-hour Coping Duration The nuclear safety goal, objectives, and performance criteria of NFPA 805 allow more flexibility than the previous deterministic FPPs based on Appendix R to 10 CFR Part 50, as well as, in part, NEI 00-01 Chapter 3, 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 the LAR, the licensee states that the NFPA 805 licensing basis for ONS is to achieve and maintain safe and stable conditions in the plant, which is defined in NFPA 805, Section 1.6.56, as follows:
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| For fuel in the reactor vessel, head on and tensioned, safe and stable conditions are defined as the ability to maintain Keff <0.99, with a reactor coolant temperature at or below the requirements for hot shutdown for a boiling water reactor and hot standby for a pressurized water reactor. For all other configurations, safe and stable conditions are defined as maintaining Keff <0.99 and fuel coolant temperature below boiling.
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| However, the licensee further states, in the LAR, that the nuclear safety goal of NFPA 805 was accomplished at ONS by developing and analyzing a comprehensive list of systems and equipment to identify those critical components required to achieve and maintain hot standby for 72 hours following a fire from at-power conditions. The licensee also states that long-term actions would be required to maintain hot standby beyond the proposed 72-hour "mission time."
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| NFPA 805 does not define a time period in which a safe and stable condition should be evaluated. Therefore, demonstrating the ability to maintain hot standby for only the first 72 hours following a fire does not, by itself, provide an adequate level of assurance that the nuclear safety goal of NFPA 805 is satisfied.
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| By letter dated November 19, 2010, (Reference 52), the licensee states that following stabilization at hot standby, assessment and repair activities, would commence to restore plant equipment needed to enable an RCS cool down in a safe and controlled manner. ONS does not anticipate a need to maintain a unit in hot standby for greater than 72 hours. Following stabilization at hot standby, assessment and repair activities would commence to restore plant equipment needed to enable PCS cool down in a safe and controlled manner. For the most limiting fire scenarios, it is anticipated that the end state of the cooldown would be an RCS temperature of approximately 250°F with a long-term strategy for reactivity, decay heat removal and inventory/pressure control. Long-term subcooled natural circulation decay heat removal is provided by supplying lake water to the steam generators and steaming to atmosphere. The extended coping period at these conditions is based on the significant volume of water available for decay heat removal and reduced need for primary makeup to only match nominal system losses. Since the scope of repair activities needed to maintain safe and stable conditions beyond 72 hours is dependent upon the magnitude and location of a potential fire, the licensee states that the mitigation strategy, damage assessment procedures, and repair equipment for both main control room (MCR) and SSF Shutdown scenarios will be established as part of a
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| "/ong term safe shutdown program" which is to be included in the scope of the NFPA 805 program. The licensee states that this program will be completed as part of its implementation of NFPA 805 and will include mitigation strategies, damage assessments, required equipment, and procedural guidance. Any changes that need to be made to the "long term safe shutdown program" during implementation will be resolved using the change evaluation process.
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| The licensee states that based on the following factors, the qualitative risk associated with the recovery of long term SSD equipment is expected to be insignificant based on the following factors:
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| OrrlCIAL USE ONLY SECURITY RELATED INrORMATION 60
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| * The number of required recovery actions is limited.
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| * Procedures will be in place for each recovery action.
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| * The staff will be trained in the use of the recovery procedures.
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| * Required tools and replacement parts will be maintained on site.
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| * The 72 hour coping period provides a reasonable assurance that adequate time is provided to augment plant staffing and complete the recovery actions.
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| The predetermined strategy with supporting procedures and repair equipment to prepare for transition from the initial hot standby condition to long-term decay heat removal provides assurance that the fuel will be maintained in a safe and stable condition. The completion of required activities described above is an implementation item (SE Section 2.9, Table 2.9-1, Item 39).
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| Based on the information provided in the LAR, as supplemented by the licensee's November 19, 2010 (Reference 52) response, the NRC staff finds that the licensee approach has adequately demonstrated the capability to achieve and maintain the fuel in a safe and stable condition for an indefinite period following a fire.
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| A. Current Transformer Circuit Analysis Attachment B of the LAR states that ONS does not align with the guidance contained in Section 3.5.2.1 of NEI 00-01 because it disagreed that an open circuit in the secondary winding of current transformers (CTs) could cause secondary fires. In addition to not meeting NFPA 805 or NEI 00-01 expectations, the staff noted that the licensee's assumption and alignment basis statements were not consistent with the ONS Design Basis Document (DBD) for fire protection, which states:
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| * CTs may induce secondary fires through the fire-induced opening of circuitry associated with the secondary side windings of the CT
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| * The impact of the fire-induced opening of CT secondary-side circuits will be considered.
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| Resolution will be provided through proper CT qualification or the performance of a fire hazards analysis to determine if a secondary fire ignition will be a concern.
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| By letter dated September 13, 2010 (Reference 12) the licensee states:
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| * The internal NRC memorandum referenced in the LAR is not part of the ONS fire protection licensing basis. Accordingly, the following statement will be removed from the B-2 Table, Section 3.5.2.1:
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| The NRC disagreed with the conclusion formed by Brookhaven National Lab that this was a credible event. Based on EPRI [Electric Power Research Institute]
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| data and documented in NRC internal correspondence, this was determined to be an "overly conservative" position and "lacked substantiation."
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| * The assumption associated with the secondary CT circuits is being removed to ensure that ONS has properly evaluated the effects of an open secondary CT as prescribed in NFPA 805, Section 2.4.2 and guided in NEI 00-01, Section 3.5.2.1.
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| * The analysis for secondary CT circuits will utilize the methodology in the ONS DBD. All CT cables are analyzed and any consequences will be included in the NSCA.
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| OrrlCIAL USE ONLY SECURITY RELATED INrORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 61
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| * The B-2 Table alignment basis statement will be revised to state that ONS aligns with the guidance provided in NEI 00-01.
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| * The NSCA shall be revised to ensure the analysis of secondary CT circuits is carried forward.
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| Completion activities needed to assure that fire-induced open secondary circuits of current transformers will not impact the ability to achieve and maintain the fuel in a safe and stable condition is identified as an Implementation Item (SE Section 2.9, Table 2.9-1, Item 40).
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| Based on the information contained in the LAR, as supplemented, the NRC staff finds the licensee's proposed method for addressing fire-induced open secondary circuits in current transformers acceptable.
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| B. Monitoring and Diagnostic Instrumentation By letter dated April 14, 2010, (Reference 11), the licensee provided its response to the NRC staff's concerns regarding the adequacy of process monitoring and diagnostic instrumentation assured to remain available in the event of fire. For shutdown from the main control room, the licensee states that the following process monitoring and diagnostic instrumentation are available:
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| * RCS Temperature
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| * moisture separator (MS) Pressure
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| * Pressurizer Level
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| * SG Level'
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| * BWST Level
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| * Source Range Flux
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| * PSW Flow to A & B SG
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| * HPI Header Flow
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| * HPI Seal Injection Flow
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| * Letdown Storage Tank Level
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| * Letdown Storage Tank Temperature
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| * RB Pressure For shutdown from the SSF, the following process monitoring and diagnostic instrumentation are identified:
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| * RCS Temperature
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| * Pressurizer Pressure
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| * Pressurizer level
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| * SSF auxiliary service water (ASW) Flow
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| * SSF RC Makeup Flow
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| * SSF RC Makeup Suction Pressure
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| * SSF RC Makeup Discharge Pressure The specific process monitoring and diagnostic instrumentation assured to remain available for each fire area, has been identified by the licensee in calculation OSC-9695, "Oconee Nuclear Safety Capability Assessment for Units 1, 2, and 3," April 8, 2010. The licensee states that the OFFICIAL USE O~JLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED I~JFORMATION 62 LAR B-2 Table will be revised to show that OSC-9695 is the reference for the NSCA that identifies the above process instrumentation. Revision of the B-2 Table is an implementation item (SE Section 2.9, Table 2.9-1, Item 38).
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| Both neutron instrumentation and SG pressure indication are available in the MCR. For SSF shutdown, however, the licensee states the use of boron sampling in lieu of neutron source range monitoring instrumentation and the lack of SG pressure instruments have been previously accepted by the staff and documented in an exemption. Attachment K of the LAR states that NRC acceptance of this configuration (i.e., the lack of neutron instrumentation and SG pressure indication at the SSF) is documented by the NRC staff in an SE dated August 31,1983 (ADAMS Accession No. ML091310038) (Reference 40).
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| With regard to diagnostic instrumentation, the licensee states that the process monitoring instrumentation described in OSC-9695, provides the operator with adequate indication to quickly recognize and mitigate abnormal plant transients. In addition, the post-fire SSD procedure will contain a list of all credited instrumentation for each affected fire area.
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| With regard to the effects of fire to instrument sense lines, the licensee's April 14, 2010 letter (Reference 11) states that the potential for inaccurate instrument indications and/or spurious equipment actuations that could occur as a result of an instrument sensing line being exposed to a fire and increased temperatures has been considered in the analysis. Instrument sensing lines that could prevent the fulfillment of the SSD performance criteria have been identified and included in the fire area compliance assessment for review. Based on an evaluation of the materials used in the primary sensing line pipes and fittings, the licensee states that fire will not impact the sensing line fluid boundary. However, exposure of the copper tubing used on the secondary plant systems is evaluated as a loss of the instrumentation function (i.e., indication or interlock).
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| C. Offsite Power The April 2010 LAR (Reference 11) Alignment Basis indicates that the availability of offsite power has been analyzed. By letter dated September 13, 2010 (Reference 12), the licensee states that the credited power supplies are the Keowee Hydro Station (KHS) and the SSF DG and neither the KHS nor the SSF DG requires offsite power. The licensee also states that the adverse consequences of offsite power being available are considered in the NSCA.
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| The licensee has created an action item to revise calculation OSC- 9291, NFPA 805 Transition B-2 Table, Section 3.1.1.7 to reword the alignment basis to clearly state that offsite power is not credited for the deterministic analysis and therefore not analyzed for its availability in the deterministic analysis. The licensee also states that the alignment statement will also be revised to ensure the proper relationship with the alignment basis. This is an implementation item (SE Section 2.9, Table 2.9-1, Item 47).
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| D. Alignment Basis Level of Detail The licensee submitted its initial LAR for the transition to NFPA 805 on May 30, 2008 (Reference 2). As a result of changes needed to comply with RG 1.205, Revision 1, this LAR was superseded by an LAR submitted on April 14, 2010 (Reference 11). A review of the 2010 LAR identified several sections of Table B-2 that had been modified to the extent that the level of detail provided was below the level needed to readily confirm alignment with the NEI guidance. In a letter dated September 13, 2010 (Reference 12), the licensee identifies a total of OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 63 57 sections of the LAR B-2 Table that had been modified. Of these, the licensee determined that fifteen sections had insufficient detail to clearly demonstrate conformance to the applicable sections of NEI 00-01. The licensee states that the LAR B-2 Table will be revised to include additional clarification of alignment with the NEI guidance. Revision of the LAR B-2 Table will be tracked in the corrective action program and include these additional clarifications as an implementation item (SE Section 2.9, Table 2.9-1, Item 38).
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| E. Armor Jacketed Cable Grounding In a letter dated November 30,2009 (Reference 10), the licensee states that the interlocked armor on the cables at ONS are terminated and grounded as required by ONS Engineering Design Criteria DC-4.11 which states that the armor of interlocked armor cable be electrically continuous and grounded to equipment enclosure at each end of the cable. The licensee also states that a sample of plant design changes have been reviewed to ensure the original design criteria is being referenced in the change modifications with regards to grounding the cable armor. Based on its review of drawings, cable specifications, and modifications, the licensee states that it has a high degree of confidence that the as-installed configuration of the armor cable grounding scheme is consistent with the original plant design.
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| Based on the information provided in the LAR, as supplemented, the NRC staff finds that the licensee has adequately addressed the issue of grounding of armored cable to preclude inter cable shorts.
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| F. Section 3.2.1 Conclusion Based on the information provided in the licensee's submittal, the NRC staff finds the method the licensee used to perform the NSCA with respect to selection of systems and equipment, selection of cables, and identifying the location of equipment and cables acceptable, because the method used either met the NRC-endorsed guidance directly or met the intent of the endorsed guidance with adequate justification as documented in Table 3.2-1 (see SE Attachment B).
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| 3.2.2. Applicability of Feed and Bleed As stated below, 10 CFR 50.48(c)(2)(iii) limits the use of feed and bleed:
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| 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 SSD path for maintaining reactor coolant inventory, pressure control, and decay heat removal capability (i.e., feed-and-bleed) for pressurized-water reactors (PWRs) is not permitted.
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| The NRC staff reviewed LAR Table 5-3, "10 CFR 50.48(c) - Applicability/Compliance References," and Attachment C, "NEI 04-02 Table B-3, Fire Area Transition," to evaluate whether the licensee meets the feed and bleed requirements. The licensee stated in LAR Table 5-3 that feed and bleed is not utilized as the sole fire protected SSD path at ONS for any scenario. The NRC staff verified this by reviewing the designated SSD path listed in LAR Attachment C for each fire area. Although loss of pressurizer heaters was considered possible, this review confirmed that all fire area analyses include the SSD equipment necessary to provide decay heat removal without relying on feed and bleed and the PORV is not the only means of pressure control for potential solid water operation. In addition, all fire areas either OFFICIAL USE O~JLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 64 met the deterministic requirements of NFPA 805, Section 4.2.3; or the PB evaluation performed in accordance with NFPA 805, Section 4.2.4, demonstrated that the integrated assessment of risk, DID, and SMs for the fire area was acceptable.
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| The NRC staff determined that based on the information provided in LAR Table 5-3, as well as the fire area analyses documented in LAR Attachment C, 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 at ONS.
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| 3.2.3. Assessment of Multiple Spurious Operations (MSOs)
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| NFPA 805, Section 2.4.2.2.1, "Circuits Required in Nuclear Safety Functions," states that:
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| Circuits required for the nuclear safety functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1, ["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.
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| In addition, NFPA 805, Section 2.4.3.2, states that the probabilistic safety assessment (PSA) evaluation shall address the risk contribution associated with all potentially risk-significant fire scenarios. Because the PB approach taken at ONS was to utilize FREs in accordance with NFPA 805, Section 4.2.4.2, "Use of Fire Risk Evaluation," adequately identifying and including potential multiple spurious operation (MSO) combinations is required to ensure that all potentially risk-significant fire scenarios have been evaluated.
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| Accordingly, the NRC staff reviewed LAR Section 4.2.1.4, "Evaluation of Multiple Spurious Operations," and Attachment F, "Fire-Induced Multiple Spurious Operations Resolution," to determine whether the licensee has adequately addressed MSO concerns at ONS. The licensee's chosen approach used an expert panel to identify potential MSO combinations, which needed to be considered in the NSCA, as well as to assess the plant-specific vulnerabilities associated with these MSO combinations.
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| The expert panel was a diversified group of subject matter experts in the following areas:
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| * Operations
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| * Fire Protection Engineering
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| * Post-Fire SSD Analysis
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| * Probabilistic Risk Assessment (PRA)
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| * Fire Protection/Post-Fire SSD Consultant The expert panel utilized guidance provided in NEI 00-01, Revision 1, Appendix F, Section F.4.2, and "Expert Panel Review" (ADAMS Accession No. ML050310295), (Reference 56). The expert panel, which was conducted in February 2006, considered: the post-fire SSD analysis for ONS, the self-assessment process identified in NEI 04-06 (Reference 44), insights provided by the internal events PRA for ONS, industry and plant-specific operating experience, and a review of the ONS Simplified Flow Diagram (OSFD) of the reactor coolant system (RCS). The OFFICIAL USE O~JLY SECURITY RELATED INFORMATION
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| O~~ICIAL USE ONLY SECURITY RELATED IN~ORMATION 65 expert panel generated a list of MSO scenarios in an effort to reflect the intent of the guidance provided in NRC Regulatory Information Summary (RIS) 2004-03, Revision 1 (Reference 45).
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| Subsequent to the expert panel meeting, the generic list of fire-induced MSO scenarios provided by the PWR Owners Group (PWROG) as part of the update process for NEI 00-01, Revision 2, were compared to the MSO scenarios identified by the expert panel and, if not already considered, were added to the ONS-specific list of MSO scenarios. The results of both the expert panel review and the review of the PWROG MSO scenarios were incorporated into the NSCA as well as the Fire PRA.
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| The MSO combinations included in the NSCA were evaluated with respect to compliance to the deterministic requirements of NFPA 805 Section 4.2.3, "Deterministic Approach." For those situations in which the MSO combination did not meet the deterministic requirements of NFPA 805, the components and associated cables were added to the scope of the FREs performed for the associated fire area.
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| The licensee's alignment basis for Section 3.5.1.5(B) of NEI 00-01 stated that circuit failures were considered to occur on each conductor of each SSD cable. However, the alignment basis did not specifically state if the failures were considered to occur individually (e.g., conductor A shorts to B; conductor A shorts to C) or concurrently (e.g., conductor A shorts to B and C). The NRC staff requested the licensee to provide additional clarification with regard to its evaluation of intra-cable circuit failures within a single multi-conductor cable. In its September 13, 2010 (Reference 12), response to a staff request for additional information, the licensee states that any and all potential spurious actuations that may result from intra-cable shorting were considered. Such failures were considered to occur concurrently, regardless of number, in accordance with the guidance provided in NEI 00-01, Section 3.5.1.5(B).
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| The NRC staff reviewed the licensee's expert panel process for identifying circuits susceptible to multiple spurious operations, as described above, and concludes that the licensee adopted a systematic and comprehensive process for identifying MSO scenarios to be analyzed utilizing available industry guidance. Furthermore, the process used provides reasonable assurance that the FRE appropriately identifies and includes risk-significant MSO combinations. Based on these conclusions, the NRC staff finds the licensee's approach for assessing the potential for MSO combinations acceptable for use at ONS.
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| 3.2.4. Transition of Operator Manual Actions to Recovery Actions NFPA 805, Section 1.6.52, "Recovery Action," defines a recovery action as follows:
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| Activities to achieve the nuclear safety performance criteria that take place outside the MCR or outside the primary control station(s) for the equipment being operated, including the replacement or modification of components.
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| NFPA 805, Section 4.2.3.1, states that:
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| 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 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.
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| O~~ICIAL USE ONLY SECURITY RELATED IN~ORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 66 NFPA 805, Section 4.2.4, "Performance-Based Approach," states the following:
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| When the use of recovery actions has resulted in the use of this approach, the additional risk presented by their use shall be evaluated.
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| The NRC staff reviewed LAR Section 4.2.1.3, "Transition of Operator Manual Actions to Recovery Actions," and Attachment G, "Operator Manual Actions Transition," to evaluate whether the licensee meets the associated requirements for the use of recovery actions per NFPA 805.
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| The licensee based its approach for transitioning operator manual actions (OMAs) into the 10 CFR 50.48(c) RI/PB FPP as recovery actions on NEI 04-02, Revision 2, Section 4.6, "Regulatory Submittal and Transition Documentation," as endorsed with exceptions by RG 1.205, Revision 1 (Reference 14).
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| The population of OMAs addressed during the NFPA 805 transition process at ONS included the existing OMAs in the deterministic FPP, as well as those being added during the NFPA 805 transition to address VFDRs of NFPA 805 Section 4.2.3. OMAs meeting the definition of a recovery action are required to comply with the NFPA 805 requirements outlined above. The licensee states that all pre-transition OMAs at ONS are actions that take place at the primary control stations and are therefore not recovery actions per NFPA 805, Section 1.6.52.
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| However, during the resolution of VFDRs, the licensee identified 12 recovery actions to satisfy the DID requirements of NFPA 80S, Section 1.2. The licensee stated that all of these recovery actions were reviewed to verify that they could not have an adverse impact that would increase risk and none were found to have an adverse impact on the Fire PRA results.
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| The licensee stated that it subjected all recovery actions to a feasibility review. In accordance with the NRC-endorsed guidance in NEI 04-02, the feasibility criteria used were based on the nine attributes provided in Section B.5.2, "Methodology Success Path Resolution Considerations," of Appendix B, "Nuclear Safety Analysis," to NFPA 805. LAR Attachment G includes Table G-1, "Feasibility Criteria - Recovery Actions and DID Actions (Based on NFPA 805, Appendix B.5.2(e) and NEI 04-02, Revision 2)," that lists the nine attributes used to assess recovery action feasibility. A feasibility evaluation was performed for each identified recovery action on a fire-area basis. Based on the results of these evaluations, the licensee determined that the recovery actions met all except the following feasibility criteria:
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| * Communication. Hand-held radios are relied upon to ensure communication between the field operator and the main and SSF control rooms. An evaluation will be performed to ensure the radios operate in the locations of the recovery actions, either with or without repeaters.
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| * Procedures. Procedures do not exist for the non-SSF fire areas. SSD procedures will be developed for Fire Areas RB1, RB2, and RB3.
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| * Training. No training is provided for fire areas utilizing the MCR for unit shutdown.
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| Training will be provided to the operators on the new SSD procedures developed for Fire Areas RB1, RB2, and RB3.
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| * Drills. Emergency drills will be conducted on the new SSD procedures developed for Fire Areas RB1, RB2, and RB3.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 67 Completion of these activities is considered an implementation item (SE Section 2.9, Table 2.9-1, Item 14). The NRC staff finds that the licensee's application of feasibility criteria for recovery actions, is consistent with the endorsed guidance found in NEI 04-02, Revision 2, and is therefore acceptable.
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| The licensee stated that no specific recovery actions were modeled in the Fire PRA. Instead, the fire-induced failure of the cables that prompted the associated recovery actions was included as VFDRs. Therefore, the risk of the VFDRs includes or bounds the risk of the identified recovery actions even if the recovery actions were modeled and assumed to be perfectly reliable. The NRC staff finds this bounding approach to demonstrating the reliability of the identified recovery actions acceptable.
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| While performing the review of the licensee's treatment of the transition of OMAs to recovery actions, the NRC staff identified several issues that required the licensee to provide additional information in order to adequately demonstrate compliance with specific portions of the applicable NFPA 805 requirements. In response to RAls (References 42 and 43), the licensee stated that it defined primary control station (PCS) actions in the LAR as follows:
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| * Actions inside the MCRs,
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| * Actions inside the SSF control room,
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| * Actions inside the SSF facility to transfer control from the MCR to the SSF,
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| * Actions inside the SSF facility to operate manual valves, and "Deployment and operation of the SSF submersible pump" was also defined as a PCS action in the LAR. This action includes actions to retrieve, assemble, and deploy the portable submersible pump, including making necessary hose(s) and electrical power connections that are not predominantly conducted in the SSF or deployed during the initial transfer of control from the control room. By letter dated November 19, 2010 (Reference 52), the licensee stated that deployment and operation of the SSF submersible pump is reclassified as a Fire Area AB VFDR and is being transitioned as a recovery action. The licensee further stated that this recovery action would only be required to support long-term SSF operation in very specific scenarios. These scenarios are modeled in the flooding PRA, but are not modeled in the Fire PRA. The licensee provided characteristics of the actions (e.g., proceduralized and periodically evaluated) based on which action meets the feasibility criteria. The NRC staff finds that the licensee's application of feasibility criteria for this recovery action is consistent with the endorsed guidance found in NEI 04-02, Revision 2, and is therefore acceptable.
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| As discussed previously, the NRC staff requested the licensee provide additional information regarding the assumption that no spurious operations or loss-of-offsite power would occur within the first 10 minutes following confirmation of an active fire event during which time the operators are transferring control to the SSF. In its letters dated August 3,2009 (Reference 8), April 14, 2010 (Reference 11), and November 19, 2010 (Reference 52), the licensee stated that the 10 minute delay was not credited in the Fire PRA, but that the Fire PRA assumed, for each associated fire scenario, applicable spurious operations and failure of all equipment determined from the fire analysis. In addition, component failures or spurious actuations caused by any fire induced damage were not subsequently credited as recovered by an OMA in the Fire PRA. The NRC staff finds this approach acceptable because fire-induced functional failures were not recovered in the Fire PRA, which resulted in a conservative assessment of the fire risk.
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| Based on the above considerations, the NRC staff finds that the licensee has followed the endorsed guidance of NEI 04-02 and RG 1.205 regarding the transition of OMAs to recovery OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 68 actions in accordance with NFPA 805, thereby meeting the regulatory requirements of 10 CFR 50.48(c). The NRC staff concludes that the feasibility criteria applied to recovery actions are acceptable based on conformance with the endorsed guidance contained in NEI 04-02.
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| 3.2.5. Conclusion for Section 3.2 The NRC staff reviewed the licensee's LAR, as supplemented, for conformity with the requirements contained in NFPA 805, Section 2.4.2, regarding NSCA at ONS. The NRC staff found that the licensee's process is adequate to appropriately identify and locate the systems, equipment, and cables required to provide reasonable assurance of achieving and maintaining the fuel in a safe and stable condition, as well as to meet the nuclear safety performance criteria of NFPA 805, Section 1.5.
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| The NRC staff verified, through review of the documentation provided in the LAR, that feed and bleed was not the sale fire-protected SSD path for maintaining reactor coolant inventory, pressure control, and decay heat removal capability, in accordance with 10 CFR 50.48(c)(2)(iii).
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| The NRC staff reviewed the licensee's process to identify and analyze MSOs. Based on the information provided in the LAR, as supplemented, the process used to identify and analyze MSOs at ONS is considered comprehensive and thorough. Through the use of an expert panel, potential MSO combinations were identified and included as necessary into the NSCAs as well as the applicable FREs. The NRC staff also considers the licensee's approach for assessing the potential for MSO combinations to be acceptable because it was performed in accordance with NRC-endorsed guidance.
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| The NRC staff found that, based on the information provided in the LAR, as supplemented, the process used by the licensee to review, categorize, and address recovery actions during the transition from the existing deterministic fire protection licensing basis to an RI/PB FPP is consistent with the NRC-endorsed guidance contained in NEI 04-02 and RG 1.205 regarding the transition of OMAs to recovery actions and other actions required to be taken at a PCS.
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| Therefore, this process meets the regulatory requirements of 10 CFR 50.48(c) and the guidelines of NFPA 805.
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| 3.3. Fire Modeling NFPA 805 allows the use of fire modeling as a PB alternative to the deterministic approach outlined in the standard. NFPA 805, Section 1.6.18, defines a fire model as a "mathematical prediction of fire growth, environmental conditions, and potential effects on structures, systems, or components based on the conservation equations or empirical data."
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| NFPA 805, Section 2.4.1, "Fire Modeling Calculations," specifically addresses the application requirements for using PB fire models as follows:
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| * NFPA 805, Section 2.4.1.2.1, "Acceptable Models," states the following:
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| Only fire models that are acceptable to the authority having jurisdiction shall be used in fire modeling calculations.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 69
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| * NFPA 805, Section 2.4.1.2.2, "Limitations of Use," states the following:
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| Fire models shall only be applied within the limitations of that fire model.
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| * NFPA 805, Section 2.4.1.2.3, "Validation of Models," states the following:
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| The fire models shall be verified and validated.
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| NFPA 805, Section 4.2.4.1, "Use of Fire Modeling," identifies the specific approach for use of fire modeling as a PB method, including the following required aspects: identify targets, establish damage thresholds, determine limiting condition(s), establish fire scenarios, protection of required nuclear safety success path(s), and operations guidance.
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| In addition, RG 1.205, Revision 1 (Reference 14), Regulatory Position C.4.2, and NEI 04-02, Revision 2 (Reference 21), Section 5.1.2, "Fire Modeling Considerations," provide guidance by identifying fire models that are considered acceptable for use by the NRC for plants transitioning to a RI/PB FPP in accordance with NFPA 805 and 10 CFR 50.48(c).
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| The NRC staff reviewed LAR Section 4.5.2, "Fire Modeling," which describes how the licensee used fire modeling as a part of the transition to NFPA 805 at ONS, and LAR Section 4.7.3, "Compliance with Quality Requirements in Section 2.7.3 of NFPA 805," which describes how the licensee complied with the NFPA 805 quality requirements for fire protection systems and features at ONS.
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| In LAR Section 4.5.2, the licensee stated that fire modeling analyses were used only to support development of the Fire PRA for use in performing FREs (Le., in accordance with NFPA 805, Section 4.2.4.2, "Use of Fire Risk Evaluations"). The fire modeling PB method of NFPA 805, Section 4.2.4.1 was not used. The NRC staff reviewed the technical adequacy of the ONS Fire PRA, including the supporting fire modeling analyses, as documented in SE Section 3.4.3.
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| Because the fire modeling PB method of NFPA 805, Section 4.2.4.1, was not used, the NRC staff has not reviewed any such methods for acceptability in that context. Since the NRC staff has not reviewed any such fire modeling methods, the staff does not find any plant-specific fire modeling methods acceptable for use to support compliance with NFPA 805, Section 4.2.4.1, as a part of this licensing action supporting transition to RIIPB FPP.
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| 3.4. Fire Risk Assessments This section addresses the licensee's FRE PB method, which is based on NFPA 805, Section 4.2.4.2.
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| NFPA 805, Section 4.2.4.2, "Use of Fire Risk Evaluations," states the following:
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| Use of fire risk evaluation for the performance-based approach shall consist of an integrated assessment of the acceptability of risk, DID, and SMs.
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| 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 2.4.4.1 ["Risk Acceptance Criteria"]. The fire risk shall be calculated using the approach described in 2.4.3 ["Fire Risk Evaluations"].
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 70 DID and SMs are discussed in SE Section 3.4.1. The acceptability of risk is discussed in SE Sections 3.4.2 through 3.4.6.
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| For those fire areas where the licensee used a PB approach to meet the nuclear safety performance criteria, the licensee used FREs in accordance with NFPA 805 Section 4.2.4.2 to demonstrate the acceptability of the plant configuration. Some VFDRs were resolved with plant modifications. Each remaining VFDR was evaluated for risk impact and maintenance of the philosophy of DID and maintenance of SMs associated with bringing the VFDR into the licensing basis instead of removing the VFDR by bringing the plant into compliance with the deterministic requirements.
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| 3.4.1. Maintaining Defense-in-Depth and Safety Margins When implementing the PB approach, the licensee followed the guidance contained in Section 5.3.5, "Acceptance Criteria," of NEI 04-02, which includes a detailed consideration of DID and SMs as part of the risk evaluation process. FREs were performed for each fire area, which includes a risk evaluation of each VFDR and a composite risk evaluation of the entire fire area.
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| Each fire area FRE includes an assessment of DID systems and features and an assessment of how adequate the SM is maintained. The results of this assessment are summarized for each VFDR by fire area, in the LAR Attachment C Table B-3.
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| Defense-in-Depth (DID)
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| NFPA 805, Section 1.2, states the following:
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| DID shall be achieved when an adequate balance of each of the following elements is provided:
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| * Preventing fires from starting.
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| * Rapidly detecting fires and controlling and extinguishing promptly those fires that do occur, thereby limiting fire damage.
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| * 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.
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| The NRC staff reviewed LAR Sections 4.2.4, "Fire Area-by-Fire Area Transition," Section 4.5.3, "NFPA 805 Fire Risk Evaluation Process," Table 4-3 "Fire Risk Evaluation Guidance Summary Table," and Table 4-4, "Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features," 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.
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| Each fire area FRE includes an assessment of DID systems and features credited to maintain a balance amongst the DID attributes and identification of DID enhancements needed to disposition VFDRs and restore the balance among the DID attributes. LAR Attachment C Table B-3 (1) documents the existing balance of DID, (2) indicates whether or not the element needs to be strengthened by modifications (such as the installation of fire detection systems or other OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 71 fire protection modification), and (3) documents the presence of automatic fire detection and suppression systems.
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| The licensee's process for evaluating fire suppression and detection systems also included a review of those systems credited to meet the NFPA 805 deterministic requirements. This review included the identification of automatic suppression and detection systems credited in NRC staff approved exemptions from the existing fire protection licensing basis and being carried forward into the RI/PB FPP as well as those credited by the licensee in EEEEs.
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| The NRC staff review finds 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 changes as proposed in its LAR adequately maintains DID against fires as required by NFPA 805.
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| Safety Margin (SM)
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| Although the appendices to NFPA 805 are not incorporated into the regulation, they may provide insight into what the authors of that Standard intended. Section A.2.4.4.3 of Appendix A to NFPA 805 provides the following background related to the meaning of the term "SMs:"
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| An example of maintaining sufficient safety margins occurs when the existing calculated margin between the analysis and the performance criteria compensates for the uncertainties associated with the analysis and data.
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| Another way that safety margins are maintained is through the application of codes and standards. Consensus codes and standards are typically designed to ensure such margins exist.
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| LAR Section 4.2.4, Section 4.5.3, and Table 4-3 stated that SMs were considered as part of the FRE process. Specifically, LAR Section 4.5.3.2 stated that the licensee evaluated each VFDR against the SM criteria of NEI 04-02.
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| NEI 04-02, Section 5.3.5.3, "Safety Margins," lists two specific criteria that should be addressed when considering the impact of plant changes on SMs:
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| * Codes and Standards or their alternatives accepted for use by the NRC are met, and
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| * Safety analyses acceptance criteria in the licensing basis (e.g., FSAR and supporting analyses) are met, or provides sufficient margin to account for analysis and data uncertainty.
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| As described in SE Section 3.1, the licensee meets various codes and standards associated with fire protection. The licensee listed the SM attributes considered during its FRE for each fire area in LAR Attachment C, Table B-3. The SM attributes listed include the following:
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| * Fire modeling performed in support of the transition has been performed within the Fire PRA utilizing codes and standards developed by industry and endorsed by NRC.
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| * Plant system performance parameters were not modified as a result of the FREs.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 72
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| * The bases for the application of the Fire PRA codes and standards were not altered in support of the FREs.
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| * In accordance with the requirements of 10 CFR 50.48(c)(2)(iii), the Fire PRA results, including cutsets for the scenarios of concern, were reviewed by the licensee and the licensee verified that the results presented do not rely solely on feed-and-bleed as the fire-protected SSD path for maintaining reactor coolant inventory, pressure control, and decay heat removal capability for each fire area.
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| In addition to the attributes listed above, the installation of the PSW and its associated risk reduction provides additional margin. The criteria described in NEI 04-02, Section 5.3.5.3 and the LAR are consistent with the criteria as described in RG 1.174 and therefore acceptable.
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| The NRC staff finds that the licensee's approach has adequately addressed the issue of SM in the implementation process because the licensee used appropriate codes and standards (or NRC approved alternatives) and met the safety analyses acceptance criteria in the licensing basis (or, through the application of its Fire PRA in its FREs provided sufficient margin to account for analysis and data uncertainty).
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| Defense-in-Depth and Safety Margin Conclusions The licensee's FRE process included a detailed review of fire protection DID and SMs. The individual FREs and LAR Table 4-4 and Attachment C Table B-3 document the results of the DID and SM review. The NRC staff finds the licensee's evaluation in regard to DID and SMs to be 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 and RG 1.174, Revision 1.
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| 3.4.2. Fire Risk Evaluation In accordance with the guidance in RG 1.205, Section C.2.2.4, "Risk Evaluations," risk increases or decreases for each fire area using FREs and for the overall plant should be provided. In LAR Attachment C, Table B-3, the licensee provided the results of the FRE for each individual VFDR that was not resolved with a modification. The risk increases and decreases associated with the VFDRs for each fire area and the total fire risk for each unit are provided in LAR Attachment W, Tables W-2, W-3, and W-4.
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| The tables in LAR Attachment W provide the estimated risk increase in each fire area associated with not modifying the facility to remove the VFDR as permitted by the RI/PB implementation of 10 CFR 50.48(c). Consistent with RG 1.174 (combined change request) and as stated in RG 1.205, it is acceptable for transition to credit selected non-fire related modifications to reduce the risk associated with retaining VFDRs. The licensee credited the risk reduction from the proposed PSW modification and provided the amount of risk reduction from PSW for each fire area. The licensee stated that the installation of the PSW system is expected to decrease the risk associated with hazards other than fire because the functions provided can be used to help mitigate other initiating events just as they are used to mitigate fire-initiated events. However, the licensee also indicated that they did not credit this additional risk reduction capability of the PSW for other hazards, but rather only estimated and provided the risk impact of the PSW installation on fire risk. The NRC staff has reviewed this process for FREs and the licensee's documented results and found it to be acceptable because it estimates the change in risk required by NFPA 805, as incorporated by reference in 10 CFR 50.48(c).
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 73 3.4.3. Quality of the Probabilistic Risk Assessment NFPA 805 Section 2.4.3.2 states that the PSA approach, methods, and data shall be acceptable to the AHJ. In reviewing a RJ LAR, the NRC staff evaluates the acceptability of the plant specific PRA analyses and their proposed application using guidance from RG 1.174. RG 1.174 addresses PRA approach methods, and data and also provides additional guidance clarifying how acceptability should be determined. In RG 1.174, 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 scope, level of detail, and technical adequacy of the PRA are to be commensurate with the application for which it is intended and the role the PRA results play in the integrated decision process. The more emphasis that is put on the risk insights and on PRA results in the decisionmaking process, the more requirements that have to be placed on the PRA, in terms of both scope and how well the risk and the change in risk is assessed. Conversely, emphasis on the PRA scope, level of detail, and technical adequacy can be reduced if a proposed change results in a risk decrease or is very small.
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| In its LAR, the licensee estimates that the total change in risk associated with its proposed transition to 10 CFR 50.48(c) will be a substantial decrease in CDF and LERF. A RI application that can clearly be shown to result in a decrease in risk is considered to have satisfied the relevant risk-related principle of RI regulation (i.e., Principle 4 of RG 1.174). Therefore, the NRC staff's review of the quality of the ONS PRA described below focused on whether the PRA is adequate to support the conclusion that granting the proposed amendment is expected to result in an overall decrease in risk.
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| The licensee performed a Fire PRA to support this application. The scope and level of detail of the Fire PRA is consistent with the requested licensing action, which changes the FPP and includes FREs. The licensee has no seismic PRA, but the NRC staff has concluded that seismic-fire interaction is adequately addressed with the licensee's qualitative analysis. Other external events, such as external floods and high winds, are expected to be insignificant causes and contributors to fire risk based on the relatively low frequency of these other external events occurring with a coincidental or consequential fire. Consistent with RG 1.174, since the licensee's application is a risk decrease, there is no requirement to calculate the total CDF or LERF contribution from all hazards when considering risk acceptability. Therefore, the NRC staff concludes that the scope and level of detail of the Fire PRA analysis is acceptable for this application.
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| As described in RG 1.174 and RG 1.200, one approach the NRC staff uses to assess the technical adequacy of the licensee's base PRA is to consider the industry peer review process and the licensee's resolution of the findings from this process for the specific application. In accordance with RG 1.200, the performance of an industry peer review of a licensee's base PRA that meets NRC-endorsed PRA standards obviates the need for the NRC staff to perform a detailed review of the licensee's base PRA. The PRA standards identify major elements of a PRA and provide numerous supporting requirements (SRs) for each element. SRs identify individual evaluation steps and describe the technical attributes of the analysis or documentation required to properly meet each SR. The results of a review against the PRA standard are referred to as findings, observations, or facts and observations (F&Os). The NRC staff's assessment of the quality of a licensee's PRA uses these F&Os as the starting point for its review of the technical adequacy of the licensee's base PRA model being used for an application.
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| OFFICIAL USE O~JLY SECURITY RELATED INFORMATION 74 In this application, the industry peer review used the ASME RA-Sb-2005 PRA standard, "Addenda to ASME RA-Sb-2002 Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications" (Reference 47), as the basis for the review of the licensee's internal events PRA. The licensee's application is an NFPA 805 pilot application. Consistent with Regulatory Position C.4.3 of the initial issuance of RG 1.205, the NRC staff performed a pre-LAR review of the licensee's Fire PRA model in lieu of an industry peer review of the model as part of the pilot process. The NRC staff's review of the Fire PRA model is discussed further under "Fire PRA Model," below.
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| The licensee identified administrative controls and processes used to maintain the Fire PRA model current with plant changes and to evaluate any outstanding changes not incorporated into the Fire PRA model for potential risk impact as part of the change evaluation process.
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| Further, as described in SE Section 3.8.3, the licensee has a program for ensuring the developers and users of these models are appropriately trained and qualified. The licensee did not identify any of the following 1) known outstanding or imminent plant changes that would require a change to the Fire PRA model, or 2) any planned plant changes during the 10 CFR 50.48(c) implementation period that would significantly impact the Fire PRA model, beyond those identified and scheduled to be implemented as part of the transition to the RI/PB FPP, as set forth in Section 2.8 of this SE. Therefore, the NRC staff finds that the PRA program for developing, maintaining, and using the Fire PRA provides confidence that the licensee has satisfied the guidance in RG 1.174, RG 1.200, and RG 1.205 that the Fire PRA model will appropriately represent the as-built, as-operated and maintained plant for this specific application, within the limitations established in the license condition and once the committed modifications, SE Section 2.8, and the implementation items, SE Section 2.9 are completed.
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| Internal Events PRA Model Revision 2 of the licensee's PRA was completed in December 1996 and peer reviewed using NEI 00-02, "Industry PRA Peer Review Process," by the B&W Owners Group in May 2001. By letter dated August 3, 2009 (Reference 8), the licensee summarized the changes made in November 2006 to update Revision 2 to Revision 3. An independent contractor for the licensee reviewed Revision 3 of the PRA using the ASME RA-Sb-2005 PRA standard (Reference 47) in June 2006. The contractor identified and commented on SRs in the ASME PRA Standard that were either "not met" or that did not meet the Capability Category (CC) II; the CC that RG 1.200 deemed as adequate for most applications. Using the results of the contractor review, the licensee made changes to the PRA, updating Revision 3 to Revision 3a in June 2008. In October 2008, the licensee reported that it performed a self-assessment on Revision 3a against the ASME PRA Standard, as modified by Revision 1 of RG 1.200.
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| In its LAR, the licensee stated that its PRA fully meets 242 of the 306 ASME PRA SRs. The licensee determined that 24 of the remaining 64 SRs were not applicable or did not need to meet a CC II to support this application. The LAR and responses to an NRC staff RAI (Reference 8) briefly described the 40 F&Os on the remaining SRs and assessed the impact of resolving these remaining observations on the reported risk of the transition to 10 CFR 50.48(c).
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| The 40 internal events PRA F&Os are described in Attachment C.1, Table 3.4-1, along with the licensee's resolution for this application and the NRC staff's conclusions regarding the acceptability of the licensee's resolution. These F&Os cover numerous aspects of the internal events PRA (e.g., human reliability modeling, internal flooding modeling, and large early release frequency calculations). A number of the F&O resolutions have not been implemented and will involve PRA modifications. These modifications of the PRA are not expected to change this application from a risk decrease to a risk increase due to the significant risk reduction attributed OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 75 to the PSW modification, and the licensee has committed to confirm a risk reduction after completion of all implementation items in Section 2.9 Table 2.9-1.
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| The NRC staff recognizes that the Fire PRA is developed from the internal events PRA model, and as such, the issues identified with the internal events PRA can impact the Fire PRA results.
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| However, the NRC staff concludes that there would have to be major errors or inaccuracies in the ONS internal events PRA in order to change the substantial estimated fire risk decrease from the PSW modification into a risk increase. Previous reviews of the licensee's internal events PRA did not identify any such major errors or inaccuracies beyond the human reliability analysis weaknesses for which sensitivity studies have been completed demonstrating a minimal expected impact on risk evaluations supporting the LAR. The NRC staff evaluated the peer review results and the licensee's responses as summarized in Attachment C.1 of this SE, and concludes that changes to the PRA after resolving all the F&Os are unlikely to result in an increase in risk due to the significant risk reduction attributable to the PSW modification.
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| Completion of all implementation items in Section 2.9 (including updating and revising the HRA) will further confirm that implementation of the RI/PB FPP will result in a risk decrease.
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| Therefore, the NRC staff finds that the internal events PRA has sufficient technical adequacy that the results can be relied upon to support the determination that the transition to NFPA805 will result in a decrease in risk.
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| Fire PRA Model The licensee developed its Fire PRA model using the guidance of NUREG/CR-6850/EPRI 1011989 (Reference 37). The model addresses both Level 1 (core damage frequency) and partial Level 2 (i.e., large early release frequency only) PRA during at-power operations. The licensee modified the internal events PRA 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 and new human actions necessary as the result of a fire. The Fire PRA was initially developed for Unit 3. A second model for Unit 2 was developed from the Unit 3 model. The licensee reported that a comparative analysis of the failures and ignition frequencies for comparable fire compartments between Units 1 and 2 was performed. The comparative analysis indicated that a separate Unit 1 fault tree and Fire PRA quantification file were not necessary because Units 1 and 2 are sufficiently similar. For the limited number of cases where the Unit 2 results were not considered to be bounding for Unit 1, the licensee aqjusted the Unit 2 model to yield results applicable to Unit 1. The adaption of a single PRA model to each unit at a multi-unit site with reasonable symmetrical designs is a common method to support risk-informed applications.
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| The NRC staff finds that the licensee's recognition and reporting on its process resulting in one PRA for Unit 3 and a common PRA for the symmetrical Units 1 and 2 is sufficient to conclude that the licensee has evaluated the impact of unit-specific differences and that the results are sufficiently unit-specific to support the LAR.
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| As stated above, the NRC staff performed a review of the licensee's Fire PRA model to determine the technical adequacy of the model because an industry peer review of the ONS Fire PRA had not been performed. The NRC staff conducted the review of the ONS Fire PRA model in March 2008 (Reference 33).
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| The NRC staff's review compared the licensee's Unit 3 Fire PRA characteristics against the SRs of the draft PRA standard ASME/ANS RA-Sa-2009, Part 4, "Fire PRA Technical Elements and Requirements" (Reference 34). The review also used the industry guidance set forth in draft NEI 07-12, "Fire Probabilistic Risk Assessment Peer Review Process Guidelines" (Reference OFFICIAL USE ONLY SECURITY RELATED INFORMATIO~J
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 76 35). The NRC review identified that the ONS Unit 3 Fire PRA, representative of all three units, was incomplete, although all tasks but one had been started and many of the tasks had been completed. Therefore, the NRC staff's audit report concluded that a focused-scope peer review of those portions of the Fire PRA that changed substantially in the time between the NRC staff's review and the submittal of the plant's 10 CFR 50.48(c) LAR may be necessary. The licensee did not have a focused-scope peer review performed. Instead the licensee provided descriptions of its resolutions to the F&Os with respect to the application to transition to NFPA 805 to the NRC staff for review. These F&Os resolutions are provided in Table V-1 of Attachment V of the LAR. The Fire PRA F&Os are described in Attachment C. Table 3.4-2, along with the licensee's resolution for this application and the NRC staff's finding regarding the acceptability of the licensee's resolution.
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| Similar to the internal events PRA review, the NRC staff concludes that there would have to be major errors or inaccuracies in the ONS Fire PRA in order to change the substantial estimated fire risk decrease from the PSW modification into a risk increase. The NRC staffs audit of the Fire PRA and review of the current resolution of all F&Os from the audit (summarized in Attachment C.of this SE) did not identify any such major errors or inaccuracies. Completion of all implementation items in Section 2.9 (including completing a peer review of the fire PRA, resolving the findings from the review, and re-evaluating the change in risk from transition) will further confirm that implementation of the RI/PB FPP will result in a risk decrease. Therefore, the NRC staff finds that the Fire PRA has sufficient technical adequacy that the results can be relied upon to support the determination that the transition to NFPA-805 will result in a decrease in risk.
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| Fire Modeling in Support of Development of Fire PRA Typically, the technical adequacy of the fire modeling that supports development of the base Fire PRA for a RI license application is determined by the PRA standards and associated peer review activities, with the NRC staff's review focused primarily on the licensee's resolution of peer review findings and the actual use of (i.e., changes made to) the PRA to address the risk impacts of the proposed LAR. However, since this LAR is a pilot application of the new 10 CFR 50.48(c) requirements, the NRC staff performed additional detailed reviews of the specific fire modeling used to support specific aspects of the Fire PRA in order to gain further assurance that these methods and approaches used for the application to transition to 10 CFR 50.48(c) are technically adequate. The follOWing paragraphs discuss the NRC staffs additional review of these aspects of the licensee's fire modeling.
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| In LAR Section 4.5.2, "Fire Modeling," the application of fire modeling was intended 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 fire modeling treatment in the Fire PRA. By letter dated August 3, 2009 (Reference 8), the licensee stated that only the generic fire modeling methodology discussed below was used for the Fire PRA and no additional fire models (e.g., Consolidate Model of Fire Growth and Smoke Transport (CFAST)) or detailed fire modeling was performed.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 77 The licensee's lOI approach applied a generic fire modeling methodology to distinguish between fire scenarios that required further evaluation and those that did not require further evaluation. In general, this methodology developed conservative lOis for each type of ignition source by assuming that the maximum heat release rates (HRR) develops at time zero and extends for up to 60 minutes. The licensee assumed their armored cables were of limited combustibility and did not include an HRR contribution from ignition and combustion of adjacent cables while developing the lOI. The licensee stated that since the poly-vinyl chloride (PVC) coating on the armored cable will not sustain propagation of 'fire along the armored cable for a significant distance, any horizontal propagation along cables is adequately captured within the target set of each scenario. In addition, in many scenarios the lOI extended vertically to the ceiling.
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| The licensee also developed screening approaches to the potential for the generation of an HGL in the compartment or fire area being analyzed. These screening HGL approaches were used in the Fire PRA to further screen scenarios and compartments that would not be expected to generate an HGL. If it was determined that HGL formation was possible, the time to HGL formation was estimated and compared to the time required for fire brigade response. The licensee has committed to install detectors in certain areas to support HGL assumptions.
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| However, because the licensee assumed their armored cables were of limited combustibility, the licensee did not include an HRR contribution from the cable material burning to their HGL formation. This resulted in lengthening the time for the HGL formation and screening out some target sets in fire compartments where HGL formation is a potential concern. While ONS has few "enclosed" compartments where HGL formation would be a concern, the NRC staff finds insufficient justification for limiting the combustibility of the armored cables and not accounting for the explicit contribution of their potential combustion to the lOI development and HGL development. However, in its November 19, 2010, (Reference 52) submittal, the licensee reported HGL formation times of at least 33 minutes using these assumptions which can be compared to an expected brigade response time of no more than 20 minutes based on observed fire drills as described in the response to RAI 5-27. The NRC staff finds that including the contribution of the combustion of armored cables is not likely to expand the lOI or accelerate the HGL formation to precede the fire brigade response to the extent that the substantive estimated fire risk decrease associated with the proposed transition to NFPA 805 will become a risk increase.
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| Qualified personnel performed a plant walk-down to identify ignition sources and surrounding targets or SSCs in compartments and applied the pre-solved empirical correlation screening tool to assess whether the SSCs were within the lOI of the ignition source. Based on the fire hazard present, these generalized lOis were used to screen from further consideration those ONS-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 98th percentile fire HRR from the NUREG/CR-6850 methodology.
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| The detailed Fire PRA used in support of the licensee's application further evaluated the ignition sources determined to adversely affect the operation of credited SSCs. The licensee adjusted the HRR values for a limited number of ignition source types (e.g., cabinets) based on fire modeling insights. For all transient fire HRRs, the 75th percentile HRR was used (Reference 8).
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| Ignition sources determined to adversely affect the operation of credited SSCs were further evaluated in the detailed Fire PRA to support the LAR. In Reference 8, the licensee clarified that the only combustible fluids that require inclusion in the Fire PRA were lubricants.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 78 NUREG-1824, "Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications" (Reference 29), documents the verification and validation (V&V) of fire models used to support applications of RI/PB FPP at nuclear power plants. 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).
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| By letter dated August 3, 2009 (Reference 8), the licensee identified the use of several empirical correlations that are not addressed in NUREG-1824. The NRC staff reviewed the empirical correlation screening tool methodology, 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.
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| In addition, the licensee also responded with a detailed listing of the fire models and empirical correlations used in the screening tool including the specific versions of the software packages used. The response also provided detailed information regarding the correlations and fire models used to support transition, as well as a cross reference between major sections of American Society for Testing and Materials (ASTM) guidance document ASTM E 1355-05a, "Standard Guide for Evaluating Predictive Capability of Deterministic Fire Models" (Reference
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| : 30) and the correlations in terms of their applicability and validation. Included in the discussion was a summary of the treatment of ZOI of electrical panels.
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| For the fire modeling screening tool, documented in the LAR and associated RAI responses, the NRC staff reviewed the quality assurance process requirements of NFPA 805 Section 2.7.3 for performing V&V, limiting the application of acceptable methods and models to within prescribed boundaries, ensuring that personnel applying acceptable methods are qualified, and performing uncertainty analysis. The NRC staff assessed the acceptability of the application of each empirical correlation based on the adequacy of the V&V documentation and the correlation's applicability within its limits. Specifically, the NRC staff used the following criteria in assessing the acceptability of each correlation:
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| * the empirical correlation is included in a fire model for which V&V has been completed and documented in NUREG-1824, and the correlation is applied within the limits of its applicability; or
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| * the empirical correlation is widely accepted and utilized by fire protection engineering professionals, is documented in an authoritative publication of the Society of Fire Protection Engineers (SFPE) (e.g., The SFPE Handbook of Fire Protection Engineering), and is applied within the limits of its applicability; or
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| * the empirical correlation has been subjected to a peer review, is published in a widely recognized peer-reviewed journal article or in a conference report (e.g.,
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| Fire Safety Journal), and is applied within the limits of its applicability.
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| Based on these criteria, the NRC staff found the application of each of the empirical correlations in the Fire PRA application acceptable. Table 3.4-3 in SE Attachment C provides a summary of the correlations used, how each was applied in the Fire PRA, the V&V basis, and the NRC staff's evaluation.
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| The licensee indicated that, in general, the criteria and modeling techniques referenced in NUREG/CR-6850 and the empirical correlation screening tool have been the primary tools used OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 79 for fire modeling in the development of the Fire PRA analysis. However, the licensee's fire modeling used for determining the lOI of postulated fire scenarios and f.or the determination of the critical fire size needed for HGL formation in the compartments of interest were different from those referenced in NUREG/CR-6850. Reviews of those deviations from NUREG/CR 6850 are also addressed in Table 3.4-3, in SE Attachment C.
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| The NRC staff's evaluation finds the fire modeling employed by the licensee in the development of the Fire PRA utilized empirical correlations that provide bounding solutions for the lOI or utilized conservative input parameters in the application of the correlations resulting in conservative results for the lOI assuming timely manual suppression is successful. Although the assumption that the fire does not ignite any additional combustible material beyond the original ignition source can be non-conservative, the NRC staff concludes that there are only a few configurations that might be affected by refining the analysis to include this fire propagation and that these few configurations could not increase the change in risk estimates to change the substantial estimated risk decrease into a risk increase. Completion of all implementation items in Section 2.9 (including completing a peer review of the fire PRA, resolving the findings from the review, and revaluating the change in risk from transition) will further confirm that implementation of the RI/PB FPP will result in a risk decrease. Therefore, the NRC staff finds that this approach provides reasonable assurance that these aspects of the fire modeling used in the development of the fire scenarios in the Fire PRA is acceptable for use in this application.
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| PRA Quality Conclusions The PRA models (internal events and fire) have been reviewed against the applicable PRA standards. All F&Os from the reviews have been investigated and addressed by the licensee for this application. Based on the NRC staff's review of the peer review results and the licensee's responses as summarized above and in Attachment C of this SE, the NRC staff concludes that changes to the PRA to resolve F&Os from the internal events and fire PRA reviews are not expected to change the substantial estimated risk decrease into a risk increase.
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| Completion of all implementation items in Section 2.9 (including an industry fire PRA peer review, resolution of all peer review comments, and recalculation of the change in risk estimates) will further confirm that implementation of the RI/PB FPP will result in a risk decrease. Therefore, the NRC staff finds that the fire PRA has sufficient technical adequacy that the results can be relied upon to support the determination that the transition to NFPA 805 will result in a decrease in risk.
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| 3.4.4. Additional Risk Presented by Recovery Actions The NRC staff reviewed LAR Attachment C, "NEI 04-02 Table B Transition," Attachment G, "Operator Manual Actions Transition," and Attachment K, "Licensing Action Transition."
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| SE Section 3.2.4 describes the evaluation and transition of Operator Manual Actions (OMAs) to recovery actions. Each VFDR was evaluated to determine if a new recovery action would be relied on to disposition the VFDR.
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| For fire areas that utilized a previously approved SSF strategy, the licensee used the guidance in RG 1.205 Revision 1 to identify recovery actions. This included consideration of Primary Control Station (PCS) and the definition of recovery action as clarified in RG 1.205, Revision 1.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 80 Based on the definition provided in RG 1.205, the ONS PCS actions are defined as:
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| * Actions inside the main control rooms,
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| * Actions inside the SSF control room,
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| * Actions inside the SSF facility to transfer control from the MCR to the SSF, and
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| * Actions inside the SSF facility to operate manual valves.
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| Any actions required to transfer control to, or operate equipment from the PCS, while required as part of the RI/PB FPP, were not considered recovery actions per the RG 1.205 guidance and any additional risk associated with these recovery actions need not be calculated. The only recovery actions the licensee identified as previously approved OMAs, were actions taken to deploy and operate the SSF submersible pump. By letter dated November 19, 2010, (Reference 52) the licensee estimated the increase in CDF associated with the fire induced loss of equipment requiring this recovery action as 2.3E-8/year. The staff finds that the guidance in RG 1.205 on how the risk of recovery actions should be evaluated has been met, and that this CDF increase is sufficiently small that the risk acceptance guidelines associated with pre approved recovery actions have all been met. All other previously approved OMAs are associated with the main control room, the PCS, or transfer of control to a PCS.
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| The licensee established 12 new recovery actions that are relied on as part of the resolution of VFDRs. The three fire areas where new recovery actions were established are RB Unit 1, RB Unit 2, and RB Unit 3. These actions are described in LAR Table G-2.
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| As described in LAR Section G.4.2, the additional risk of a recovery action is conservatively taken as the CDF and LERF associated with the VFDR that resulted in the need for the recovery action. The additional risk of the new recovery actions is estimated to be negligible as reported in LAR Tables W-2, W-3, and W-4 (References 12). The licensee reviewed all of the recovery actions for adverse impact and dispositioned each action as stated in LAR Attachment G Section G.3.2. None of the OMA's listed in LAR Table G-2 that were identified as recovery actions were found to have an adverse impact on the Fire PRA.
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| The NRC staff has reviewed the licensee's evaluations for the additional risk of recovery actions and finds that the approach applied is acceptable because it utilizes the definition of recovery actions in NFPA 805 and RG 1.205, conservatively estimates the risk of previously approved and new recovery actions, and the risk associated with the new and the previously approved recovery actions are included appropriately in the change in risk estimates.
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| 3.4.5. Risk-Informed or Performance-Based Alternatives to NFPA 805 Alternatives to Compliance With NFPA 805, Section 50.48(c)(4):
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| The final rule provides licensees the flexibility of requesting, via a license amendment, to use risk-informed or performance-based alternatives that deviate from compliance with NFPA 805. The NRC recognizes that licensees may propose acceptable approaches that are not encompassed by the criteria in NFPA 805. Therefore, the NRC is including a provision for requesting such approaches in the rule. However, to ensure adequate protection of public health and safety, the NRC is requiring that licensees obtain NRC OFFICIAL USE ONLY SECURITY RELATED I~JFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 81 review and approval to use those methods, and is providing criteria in Section 50.48(c)(4) for review of their acceptability.
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| Final Rule, Voluntary Fire Protection Requirements for Light Water Reactors; Adoption of NFPA 805 as a Risk-Informed, Performance-Based Alternative, (69 FR 33,543) (June 16, 2004).
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| The licensee made no requests under 50.48(c)(4).
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| 3.4.6. Cumulative Risk and Combined Changes NFPA 805, Section 2.4.4.1, "Risk Acceptance Criteria," states the following:
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| The change in public health risk from any plant change shall be acceptable to the AHJ.
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| CDF and LERF shall be used to determine the acceptability of the change. When more than one change is proposed, additional requirements shall apply. If previous changes have increased risk but have met the acceptance criteria, the cumulative effect of those changes shall be evaluated. If more than one plant change is combined into a group for the purposes of evaluating acceptable risk, the evaluation of each individual change shall be performed along with the evaluation of combined changes.
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| The acceptability guidelines for changes to plant risk are described in RG 1.174. RG 1.205 further clarifies that changes in risk are to be judged on a fire-area by fire area-basis, as well as the total change in risk. RG 1.205 also clarifies that the additional risk from previously approved recovery actions may be reported separately and treated differently than changes in risk from other plant changes required to be estimated during transition to 10 CFR 50.48(c).
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| As allowed by RG 1.174 (combined change request) and RG 1.205 for transition, credit for selected non-fire related modifications (e.g., PSW modifications) that affect the Fire PRA results can be considered. [[
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 82 OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY REU\TED INFORMATION 83 II D. Cumulative Changes in Risk During transition, the cumulative risk is addressed by including all RI changes in the risk estimates. The licensee reported that its estimate of the internal events CDF and LERF for Unit 3 are [[ ]] per year and[[ ]] per year, respectively. Summing the internal events and Fire PRA risk estimates, crediting the PSW modification, yields CDF and LERF estimates for Unit 3 of [[ ]] per year and [[ ]] per year, respectively. As described above, implementation of NFPA-805 and installation of the PSW is expected to reduce risk. As stated in RG 1.174, if the application is clearly shown to result in a decrease in risk (Le., CDF and LERF), then the change is considered to have satisfied the relevant principle of risk-informed regulation (Principle 4 of RG 1.174) regardless of the total plant risk. Therefore, the licensee did not need to provide estimates for the contribution from other external events such as seismic, external floods, high winds, and tornados.
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| The licensee reported the total change in fire CDF and fire LERF for each fire area with and without credit for the PSW in LAR Tables W-2, W-3, and W-4. Credit for the fire detection modifications is already included in all these results. The reported results include the risk increases associated with accepting the VFDRs without credit for the PSW in all fire areas and the risk decreases from credit for the PSW in all fire areas. By letter dated November 19, 2010, (Reference 52), the licensee identified a number of additional VFDRs in the AB that require an evaluation of their risk impact but that are not included in the LAR. These VFDRs occur if the licensee assumes that damage from fires can occur at any time, not only after 10-minutes as assumed in the current licensing basis. The licensee stated that these additional VFDRs would arise from possible spurious actions within the first 10-minutes of confirmation of an active fire.
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| The licensee reported that the initial results of its evaluation of potential risk increases from these additional VFDRs in Unit 3 are [[ ]] and [[ ]] for CDF and LERF respectively. These values are reflected in Tables 3.4 and 3.5 of this SE by increasing the appropriate estimates in the Units 1 and 2 and Unit 3 AB results. Due to the significant risk reduction attributed to the PSW modification, and based on the preliminary sensitivity analyses performed by the licensee, none of these additional VFDRs/recovery actions are expected to change this application from a risk decrease to a risk increase. In addition, the licensee has identified implementation items in Section 2.9 of this SE to ensure their resolutions will maintain this application as a risk decrease including, if necessary, to make plant modifications to address these VFDRs.
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| The reported results include the risk increases associated with accepting the VFDRs without credit for the PSW in all fire areas, and the risk decreases from credit for the PSW in all fire areas. The net change in risk including both VFDRs and the new PSW for each area except the reactor building meet the "very small" increase acceptance guidelines for CDF and LERF of
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| [[ ]] and [[ ]], respectively. The increases in LERF from accepting VFDRs in the each unit's RB are about [[ ]] with or without credit for the PSW. These increases in LERF values[[ ]] the 1E-7/year "very small" LERF increase acceptance guideline in RG 1.174, but [[ ]] "small" LERF increase acceptance guideline (1 E 6/year). The NRC staff finds that the licensee has reported the individual results required by OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 84 NFPA 805, RG 1.200, and RG 1.174 Section 2.1.1 on combined change requests, and finds the individual results acceptable in light of the overall risk decrease associated with the PSW modification.
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| The changes in total fire risk for each unit are provided in the following table. RG 1.174 directs that the total change in risk from all hazard events (internal, fires, external floods, etc.)
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| associated with the proposed change be evaluated and compared to the acceptance guidelines.
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| As shown in the table, there is an overall decrease in fire risk associated with the changes proposed in this LAR due to the significant decrease achieved by the installation of the PSW.
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| Table 3.4: Fire CDF and LERF for ONS Unit 1 Unit 2 Unit 3 Fire CDF Fire LERF Fire CDF Fire LERF Fire CDF Fire LERF Risk increase [[ ]] [[ ]] [[ ]] [[ ]] [[ ]] [[ ]]
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| from accepting VFDRs Risk decrease from PSW installation Total Change in Risk Final Total Fire t 7 Risk including psw V V V V V V Therefore, the NRC staff finds that the combined change request is acceptable because of the overall fire risk decrease associated with the installation of the PSW modification.
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| 3.4.7. Conclusion for Section 3.4 Based on the NRC staff's review of the information provided by the licensee in the LAR, Transition Report, and associated RAI responses, the NRC staff review finds:
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| : 1. The licensee's PRA used to perform the risk assessments in accordance with NFPA 805 Section 2.4.4 (plant change evaluations) and Section 4.2.4.2 (fire risk evaluation) is of sufficient quality to support the application to transition to NFPA-805, because the NRC staff concludes that the weaknesses and limitations, discussed in SE Attachment C, are not expected to change the substantial estimated risk decrease into a risk increase.
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| : 2. The licensee's resolution of numerous PRA review F&Os, discussed in SE Attachment C, are directed toward determining that resolving the issues would not change the substantial estimated risk decrease associated with transitioning to 10 CFR 50.48(c) into a risk increase. However, given the number of F&O resolutions that are not fully complete or have not been implemented and will involve PRA method and model changes, the licensee has committed to complete several implementation items identified in Sections 2.9 to further confirm that implementation of the RI/PB FPP will OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 85 result in a risk decrease. The following items are identified as implementation items in SE Section 2.9:
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| : a. the licensee updates or upgrades its HRA methodology and estimates of human failure probability and, if needed, completes a peer review of this analysis, and resolves all the findings from this peer review,
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| : b. the licensee upgrades its Fire PRA, completes a peer review of its upgraded Fire PRA, and resolves all the findings from this peer review,
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| : c. The licensee includes in its PRA the as-built PSW system and completes the modeling of any additional VFDRs that are caused by assuming that damage from fires can occur at any time, not only after 10 minutes, and
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| : d. the licensee confirms that there was a reduction in risk associated with transition to NFPA after completing the improvements to the PRA, modeling any additional VFDRs, and modeling the as-built PSW system in the PRA.
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| : 3. The plant change process included a detailed review of fire protection DID and SM. The evaluations provided by the licensee are acceptable because the licensee's process followed the endorsed guidance in NEI 04-02, Revision 2 and is consistent with the approved NRC staff guidance in RG 1.205, Revision 1.
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| : 4. The additional risk presented by the use of recovery actions was determined and provided in accordance with the guidance in RG 1.205 Revision 1 and NFPA 805 Section 4.2.4. The risk of those recovery actions was found to be acceptable since they were below the acceptance guidelines in RG 1.205, Revision 1, and RG 1.174.
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| : 5. The licensee did not request approval of any risk informed or performance-based alternatives to compliance to NFPA 805.
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| : 6. The licensee's application is a combined change, as defined by RG 1.205, Revision 1, which combines risk increases identified in the FREs with risk decreases due to other modifications (e.g., PSW). The combined change process is consistent with RG 1.174 and RG 1.205 and is acceptable.
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| : 7. The changes in risk (Le., LlCDF and LlLERF) associated with the proposed alternatives to compliance with the deterministic criteria of NFPA 805 (plant change evaluations and FREs) are consistent with RG 1.205 and RG 1.174 guidelines and are acceptable.
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| 3.5. Nuclear Safety Capability Assessment Results NFPA 805, Section 2.2.3, "Evaluating Performance Criteria" states the following:
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| 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.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 86 NFPA 805, Section 2.2.4, "Performance Criteria" states the following:
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| 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.
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| NFPA 805, Section 2.2.7, "Existing Engineering Equivalency Evaluations" states:
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| 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.
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| 3.5.1. Nuclear Safety Capability Assessment Results by Fire Area NFPA 805, Section 2.4.2, "Nuclear Safety Capability Assessment," states the following:
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| The purpose of this section is to define the methodology for performing a nuclear safety capability assessment. The following steps shall be performed:
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| (1) Selection of systems and equipment and their interrelationships necessary to achieve the nuclear safety performance criteria in Chapter 1.
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| (2) Selection of cables necessary to achieve the nuclear safety performance criteria in Chapter 1.
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| (3) Identification of the location of nuclear safety equipment and cables.
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| (4) Assessment of the ability to achieve the nuclear safety performance criteria given a fire in each fire area.
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| This section of the SE addresses the last topic regarding the ability of each fire area to meet the nuclear safety performance criteria of NFPA 805. SE Section 3.2.1 addressed the first three topics.
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| NFPA 805, Section 2.4.2.4, "Fire Area Assessment," also states the following:
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| 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.
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| In accordance with the above, the process defined in NFPA 80S, Chapter 4 provides a framework to select either a deterministic or a PB approach to meeting the nuclear safety performance criteria. Within each of these approaches, additional requirements and guidance provide the information necessary for the licensee to perform the engineering analyses needed OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATIO~J 87 to determine which fire protection systems and features are required to meet the nuclear safety performance criteria of I\lFPA 805.
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| NFPA 805, Section 4.2.2, "Selection of Approach," states the following:
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| 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 utilize deterministic methods for simplifying assumptions within the fire area.
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| This section of the SE evaluates the approach used to meet the nuclear safety performance criteria on a fire area basis, as well as what fire protection features and systems are required to meet the nuclear safety performance criteria.
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| The NRC staff reviewed LAR Section 4.2.4, "Fire Area-by-Fire Area Transition," Attachment C, "NEI 04-02 Table B Fire Area Transition," Attachment G, "Operator Manual Actions Transition to Recovery Actions," Attachment S, "Plant Modifications" and Attachment W, "Fire PRA Insights" (Reference 11) during its evaluation of the ability of each 'fire area to meet the nuclear safety performance criteria of NFPA 805.
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| The ONS is a three-unit plant and is divided into 15 fire areas. Based on the information provided by the licensee in the LAR, as supplemented, the licensee performed the NSCA on a fire area basis for each of those fire areas. LAR Attachment C provides the results of these analyses on a fire area basis. For each fire area, the licensee documented the following:
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| * The approach used in accordance with NFPA 805 (Le., the deterministic approach in accordance with NFPA 805, Section 4.2.3, or the PB approach in accordance with NFPA 805, Section 4.2.4).
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| * The SSCs required in order to meet the nuclear safety performance criteria.
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| * An evaluation of the effects of fire suppression activities on the ability to achieve the nuclear safety performance criteria.
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| The disposition of each VFDR using either modifications (completed or committed) or the performance of a FRE in accordance with NFPA 805, Section 4.2.4.2.
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| For a fire in the fire areas of the Units 1 and 2 Blockhouse (BH1/2), the Unit 3 Blockhouse (BH3), the RBs (RB1, RB2, and RB3), the SSF, the TB, the west penetration rooms (WP1, WP2, and WP3), or yard, safe and stable plant conditions are achieved utilizing the PSW pump and credited systems powered from the PSW power supply and controlled from the main control rooms. For a fire in Fire Areas CT4 Blockhouse (CT-4), Keowee Hydro Station (KEO), or PSW building, normal shutdown systems are not affected by the fire and the plant will be shutdown, if desired, using normal shutdown systems and operating procedures.
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| For a fire in the fire area AB the SSF will perform as a dedicated shutdown facility used to establish safe and stable plant conditions.
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| The licensee also performed a detailed analysis of fire protection DID with respect to fire detection and fire suppression systems for each fire area. LAR Section 4.8 includes a detailed listing of the fire areas, fire zones, and fire protection features necessary to meet the OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 88 requirements of NFPA 805. LAR Table 4-4, "Summary of f\lFPA 805 Compliance Basis and Required Fire Protection Systems and Features," provides a detailed listing of the fire areas and fire zones at ONS, as well as an indication of whether automatic fire suppression and detection systems are required in these areas. This table identifies those fire areas/zones where automatic suppression and detection system modifications are required and list the regulatory and/or technical issue that makes the system required.
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| SE Table 3.5 identifies and briefly describes each fire area at ONS. This table is based on LAR Table 4-4, "Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features," which was provided by the licensee in LAR Section 4.8, "Summary of Results."
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| SE Table 3.5 also identifies the NFPA 805 compliance basis for each fire area, as well as the change in risk associated with CDF and LERF, as calculated by the licensee. The change in risk is broken down into three categories: (1) the risk increase due to VFDRs, (2) the risk decrease due to implementation of the PSW modification, and (3) the total change in risk resulting from summing the first two categories. The detailed discussion for each fire area, including the NRC staff's evaluation of the licensee's compliance with the applicable requirements, is contained in SE Attachment D, "Nuclear Safety Capability Assessment Results by Fire Area."
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 89 Table 3.5: ONS Fire Area and Compliance Strategy Summary Fire Risk Evaluation PSW Modification Delta Total Licensing NFPA 805 Risk Delta Risk Delta Risk Fire Area Fire Area Description Actions Compliance Credited? Basis ACDF ALERF ACDF ALERF Unit 1
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| [[ ]] [[ 11 [[ ]] [[ ]] [[ 11 [[ 11 AB Auxiliary Building Yes 4.2.4.2 BH12 Units 1 & 2 Block House Yes 4.2.4.2 BH3 Unit 3 Block House Yes 4.2.4.2 RB1 Unit 1 Reactor Buildin Yes 4.2.4.2 RB2 Unit 2 Reactor Buildin Yes 4.2.4.2 RB3 Unit 3 Reactor Buildin Yes 4.2.4.2 SSF Standby Shutdown Facilit Yes 4.2.4.2 TB Turbine Buildin Yes 4.2.4.2 WP1 Unit 1 West Penetration Room Yes 4.2.4.2 WP2 Unit 2 West Penetration Room Yes 4.2.4.2 WP3 Unit 3 West Penetration Room Yes 4.2.4.2 YARD Yard Yes 4.2.4.2 Unit 2
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| [[ 11 [[ ]] [[ ]] [[ 11 [[ ]] [[ ]]
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| AB Auxiliary Building Yes 4.2.4.2 BH12 Units 1 & 2 Block House Yes 4.2.4.2 BH3 Unit 3 Block House Yes 4.2.4.2 RB1 Unit 1 Reactor Buildin Yes 4.2.4.2 RB2 Unit 2 Reactor Buildin Yes 4.2.4.2 RB3 Unit 3 Reactor Buildin Yes 4.2.4.2 SSF StandbY Shutdown Facilit Yes 4.2.4.2 TB Turbine Buildin Yes 4.2.4.2 WP1 Unit 1 West Penetration Room Yes 4.2.4.2 WP2 Unit 2 West Penetration Room Yes 4.2.4.2 WP3 Unit 3 West Penetration Room Yes 4.2.4.2 YARD Unit 3 OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 90 Table 3.5: ONS Fire Area and Compliance Strategy Summary Fire Risk Evaluation PSW Modification Delta Total Licensing NFPA 805 Delta Risk Risk Delta Risk Fire Area Fire Area Description Actions Compliance Credited? Basis dCDF dLERF dCDF dLERF dCDF dLERF AB Auxiliary Building Yes 4.2.4.2 BH12 Units 1 & 2 Block House Yes 4.2.4.2 BH3 Unit 3 Block House Yes 4.2.4.2 RB1 Unit 1 Reactor Buildinq Yes 4.2.4.2 RB2 Unit 2 Reactor Building Yes 4.2.4.2 RB3 Unit 3 Reactor Building Yes 4.2.4.2 SSF Standby Shutdown Facility Yes 4.2.4.2 TB Turbine Building Yes 4.2.4.2 WP1 Unit 1 West Penetration Room Yes 4.2.4.2 WP2 Unit 2 West Penetration Room Yes 4.2.4.2 WP3 Unit 3 West Penetration Room Yes 4.2.4.2 YARD Other CT-4 CT-4 Block House KEO Keowee Hydro Station PSW Protected Service Water Note: N/A - Not Applicable, applies to those fire areas that are deterministically compliant in accordance with NFPA 805, Section 4.2.3.
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| NI- Not Included, applies to the planned PSW structure, which was not included as a fire compartment in the Fire PRA; the licensee states that the additional risk from a PSW fire is insignificant.
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| £ - The delta risk is epsilon or negligible. Plant equipment associated with the VFDRs in the fire area were evaluated and not modeled in the Fire PRA and are not expected by the licensee to introduce any significant risk contributors to the risk being evaluated for the FREs.
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| OFFICIAL USE ONLY SECURITY REU\TED INFORMATION 91 SE Attachment D is broken down into those fire areas that were analyzed using the deterministic approach in accordance with NFPA 805, Section 4.2.3, and those using the PB approach in accordance with NFPA 805, Section 4.2.4. Each fire area includes a discussion of how the licensee met the requirement to evaluate the fire suppression effects on the ability to meet the nuclear safety performance criteria.
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| SE Attachment D also addresses NRC staff-approved exemptions and* other licensing actions that exempt the licensee from the existing deterministic licensing basis that the licensee desires to bring into the RIIPB FPP as allowed by NFPA 805, Section 2.2.7. The attachment includes a description of the previously approved exemption or other licensing action that exempts the licensee from the deterministic requirements, the basis and continuing validity of the exemption or other licensing action, and the NRC staff's evaluation of that exemption or licensing action.
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| The licensee stated in LAR Section 4.2.3, "Licensing Action Transition," that the review of these existing licensing actions included a determination of the basis of acceptability and a determination that the basis of acceptability was still valid.
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| A primary purpose of NFPA 805, Chapter 4 is to determine, by analysis, what fire protection features and systems need to be credited to meet the nuclear safety performance criteria.
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| Four sections of NFPA 805, Chapter 3, have requirements dependent upon the results of the engineering analyses performed in accordance with NFPA 805, Chapter 4: (1) fire detection systems, in accordance with Section 3.8.2; (2) automatic water-based fire suppression systems, in accordance with Section 3.9.1; (3) gaseous fire suppression systems, in accordance with Section 3.10.1; and (4) passive fire protection features, in accordance with Section 3.11. The features and systems addressed in these sections are only required when the analyses performed in accordance with NFPA 805, Chapter 4, indicate that the features and systems are required to meet the nuclear safety performance criteria.
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| Passive fire protection features address the fire barriers used to form fire area boundaries (and barriers separating SSD trains) that were previously reviewed and approved in accordance with the licensee's existing deterministic FPP. For its transition to NFPA 805, the licensee decided to retain most of the previously approved fire area boundaries as part of the RIIPB FPP.
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| The fire barrier fire resistance rating required for separation between fire areas under NFPA 805 (3 hours) is the same as that required under Appendix R (3 hours). Accordingly, based on the previously approved fire area boundaries continuing to meet the NFPA 805 fire barrier acceptance criteria, the NRC staff finds retaining these passive fire protection features acceptable.
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| For its transition to NFPA 805, the licensee has also decided to create two new fire areas, the TB and AB fire areas, from the previously approved balance-of-plant fire area, as part of the RI/PB FPP. The licensee plans to upgrade the fire barriers separating the TB from the AB and separating the AB from the west penetration room to have a 3-hour fire resistance rating as described in LAR Attachments A, C, and S. The licensee has also committed to upgrade the fire barriers between the purge inlet rooms and the spent fuel pool (SFP) area (AB Fire Area) to have a 3-hour fire resistance rating. Based on the licensee's commitment to upgrade these fire barriers to meet the NFPA 805 fire barrier acceptance criteria, the NRC staff finds these new passive fire protection features acceptable. The fire barriers being upgraded are described in SE Section 2.8.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 92 The licensee's FREs identified the need to improve general area and/or hazard detection in several fire areas, either to support assumptions made in the Fire PRA or to provide DID. The licensee plans to upgrade and/or install new automatic fire detection systems in many fire zones throughout the plant, which are identified by fire area in SE Attachment D. In response to an NRC staff RAI (Reference 12), the licensee has stated that the upgraded and newly installed fire detection systems will be installed in accordance with NFPA 72, Nationa/ Fire Alarm Code, as required by NFPA 805. Based on the licensee's commitment to upgrade existing fire detection systems and install new fire detection systems to meet the NFPA 805 criteria, the NRC staff finds these upgraded/new fire detection systems acceptable. SE Section 2.8 describes the fire detection system modification.
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| In addition to the above, SE Attachment 0 provides an evaluation of the credited recovery actions for each applicable fire area. As discussed in SE Section 3.2.4, the licensee credited recovery actions to satisfy the DID requirements of NFPA 805, Section 1.2, but are not needed to maintain the availability of a success path and do not adversely impact risk. Because the licensee has identified these recovery actions as being necessary to provide adequate DID, the NRC staff has evaluated them as a part of the RIIPB FPP. As such, future removal of these recovery actions would require a plant change evaluation in accordance with NFPA 805, Section 2.4.4.
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| Finally, as a part of the NSCA, the licensee evaluated fire detection and suppression systems on a fire zone basis. In SE Attachment 0, the evaluation of each fire area includes a table that documents the licensee's review of these fire detection and suppression systems, as well as the NRC staff's evaluation of the review and its results.
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| As documented in SE Attachment 0, for those fire areas that utilized a deterministic approach in accordance with NFPA 805, Section 4.2.3, the NRC staff finds that each of the fire areas analyzed using the deterministic approach meets the associated criteria of NFPA 805, Section 4.2.3.2. This conclusion is based on (1) the licensee's documented compliance with NFPA 805, Section 4.2.3.2; (2) the licensee's assertion that the success path will be free of fire damage without reliance on recovery actions; (3) an assessment that the suppression systems in the fire area will have no impact on the ability to meet the nuclear safety performance criteria; and, (4) the licensee's appropriate determination of the automatic fire suppression and detection systems required to meet the nuclear safety performance criteria.
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| In addition, for those fire areas that utilized the PB approach in accordance with NFPA 805, Section 4.2.4, the NRC staff finds that each fire area has been properly analyzed, and compliance with the NFPA 805 requirements demonstrated as follows:
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| * Exemptions and other licensing actions that exempt the licensee from the existing fire protection licensing basis were reviewed for applicability, as well as continued validity, and found acceptable.
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| * VFDRs were either evaluated and found to be acceptable based on an integrated assessment of risk, DID, and SMs, or modifications were planned/implemented to address the issue.
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| * Recovery actions used to demonstrate the availability of a success path to achieve the nuclear safety performance criteria, or to provide DID, were evaluated and the additional risk of their use determined, reported, and found to be acceptable.
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| OrrlCIAL USE ONLY SECURITY RELATED It-WORMATION 93
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| * The licensee's analysis appropriately identified the fire protection SSCs required to meet the nuclear safety performance criteria, including:
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| Fire suppression and detection systems.
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| Fire area boundaries (ceilings, walls, and floors), such as fire barriers, fire barrier penetrations, and through penetration fire stops.
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| Accordingly, each fire area utilizing the PB approach was able to achieve and maintain the nuclear safety performance criteria, and the associated FREs meet the applicable NFPA 805 requirements for risk, DID, and SMs.
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| 3.5.2. Fire Protection During Non-Power Operational Modes NFPA 805 Section 1.1, "Scope," states:
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| 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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| NFPA 805 Section 1.3.1, "Nuclear Safety Goal," states the following:
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| 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.
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| Thus, the nuclear safety goal of NFPA 805 requires the evaluation of the effects of a fire during any operational mode and plant configuration, including non-power operation (NPO) modes. In general, the underlying concerns are the differences between the functional requirements (i.e., a different (or additional) set of systems and components) and time dependencies on decay heat removal system operation during NPOs and full-power operations.
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| The NRC staff reviewed LAR Section 4.3, "Non-Power Operational Modes" and Attachment 0, NEI 04-02, Table F-1, "Non-Power Operational Modes Compliance," to evaluate the licensee's treatment of potential fire impacts during NPOs. The licensee used the process from NEI 04-02 (Reference 21), for demonstrating that the nuclear safety performance criteria are met for Higher Risk Evolutions (HREs) during NPO modes.
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| To clarify the guidance from NEI 04-02, on providing "reasonable assurance that a fire during non-power operations will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition." the NRC staff issued interim guidance in FAQ 07-0040, "Non-Power Operations Clarification," Revision 4. Specifically, FAQ 07-0040 clarifies the following:
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| * The process for selecting equipment and cabling to evaluate during NPO modes.
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| * Evaluation of HREs during NPO modes.
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| * The process for analyzing key safety functions (KSFs) in different plant operating states (POSs).
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| OrrlCIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY REU\TED INFORMATION 94
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| * The actions taken beyond the normal FPP DID actions when a specific KSF could be lost as a direct result of fire damage.
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| As discussed in FAQ 07-0040, protection of equipment during NPO modes includes a combination of the normal FPP DID actions and additional RI steps based on the availability of systems and equipment needed to support KSFs, and whether or not the plant is in an HRE.
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| The licensee states that its strategy for control and protection of equipment during NPO modes includes a combination of normal fire protection DID actions, additional RI steps based on the availability of systems and equipment needed to support KSFs, and whether or not the plant is in an HRE.
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| The licensee defines KSFs as:
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| * decay heat removal,
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| * reactor coolant system (ReS) inventory control,
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| * reactivity control, and
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| * power availability, including support functions.
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| The licensee determined that containment closure was not a KSF since it does not directly support the nuclear safety goals of NFPA 805. However, the licensee identified the importance of demonstrating with high confidence that the equipment hatch can be closed prior to a release that would exceed the NFPA 805 radiological release criteria. Establishing high confidence would require the implementation of additional fire protection DID actions. Developing a process to evaluate the potential effects of fire on habitability and the impact of additional DID actions is an implementation Item 16.
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| As discussed in FAQ 07-0040, each plant may have a unique definition of what constitutes a higher risk evolution. However, the definition should consider the following:
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| * time to boil
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| * reactor coolant system and fuel pool inventory
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| * decay heat removal capability In LAR Attachment 0, the licensee defines an HRE as outage activities, plant configurations or conditions during shutdown where the plant is more susceptible to an event causing the loss of a KSF. The licensee further states that HREs include:
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| * Draining to reduced inventory when reactor coolant level is at or below the reactor vessel flange.
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| * Reactor coolant system at or below reduced inventory
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| * Midloop operation
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| * Any specific evolution determined by station management OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 95 Reduced Inventory is further defined by the licensee as a configuration with fuel in the reactor vessel and level less than 50" above the centerline of the reactor vessel hot leg. The licensee states that decay heat removal capability and time to boiling isaddressed in its shutdown risk management procedure (NSD-403) with the term Thermal Margin, which is the time to core boiling upon loss of decay heat removal.
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| The licensee states that it used the process from NEI 04-02, as clarified by FAa 07-0040, to demonstrate that nuclear safety performance criteria of NFPA 805 are met for HREs during NPO modes. This process includes the following steps:
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| * Review the existing outage management processes.
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| * Identify equipment/cables.
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| * Review plant systems to determine success paths that support each of the DID KSFs.
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| * Identify cables required for the selected components and determine their routing.
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| * Perform fire area assessments (identify pinch points).
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| * Manage risk associated with fire-induced vulnerabilities during the outage.
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| The same basic methodology utilized for the nuclear capability safety assessment is used when assessing the impact of fire on nuclear safety during NPO modes. The licensee states that KSF are identified in Shutdown Risk Management Procedure NSD 403 and additional detail is provided in SD 1.3.5, Shutdown Protection Plan. Thus, the licensee's evaluation focused on those sets of systems, components and equipment that are required to ensure that the KSF's defined in these procedures can be maintained during potential HRE's. The licensee is revising Fleet Directive NSD-403 and SD 1.3.5, definition of "high risk evolution," to address NPO criteria and to reconcile thermal margin criteria with the criteria in FAa 07-0040. These revisions are an implementation item (SE Section 2.9, Table 2.9-1, Item 15).
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| The process used by the licensee to identify the systems and equipment to be included in the NPO review began with the identification of the POSs that need to be considered. The POSs identified are those provided in FAa 07-0040 for PWRs, which are consistent with those contained in Attachment 2 to Appendix G, "Phase 2 Significance Determination Process Template for PWR during Shutdown," of NRC Inspection Manual Chapter (IMC) 0609, "Significance Determination Process" (Reference 46). For other non-power conditions (e.g.,
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| defueled reactor vessel), normal FPP controls, processes and procedures will be used.
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| After identifying the plant-specific POSs that require additional equipment to be included in the NPO review, the licensee states that it performed the following:
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| : 1. Determined the KSFs that support the POS of concern.
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| : 2. Identified the equipment relied upon to provide the KSFs, including support functions, during the POS to be evaluated. This information was then entered into the Appendix R Database Management System (ARTRAK) SSD database to facilitate sorting of the component and cable information on a fire zone by fire zone basis. For those components not already in ARTRAK, cable selection and routing was performed as per the nuclear safety OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATIO~J 96 methodology. The nuclear safety capability analysis methodology identified all required cables associated with a component and does not perform a circuit analysis until the area by-area compliance assessment. As a result, according to the licensee, a conservative population of components and cables was identified for NPO. The resulting information was entered into ARTRAK and a series of NPO reports were developed within the software to allow the 'pinch point' analysis to be performed.
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| : 3. Utilizing the fire zone cable routing and equipment location information from ARTRAK, the licensee analyzed the KSF success paths on a fire zone by fire zone basis to assess the impact of a single fire.
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| : 4. Analyzed circuits of equipment not already credited (or credited in a different way, such as on versus off, open versus closed, etc.), in accordance with the nuclear safety methodology and identified additional cables to be included in the NPO review.
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| The licensee states that the current outage management procedures and site directives do not include all of the conditions applicable to the POSs reviewed in the NPO evaluation. To address this finding the licensee states the following activities will be performed:
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| * Develop procedural guidance to monitor BWST temperature before freezing occurs.
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| * Develop procedural controls to monitor lake levels and the availability of the reverse gravity condenser circulating water flow path during HREs.
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| * Develop procedural controls to use RCS wide-range pressure instruments, in lieu of reactor coolant (RC) low-range pressure, during HREs.
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| * Develop procedural controls to monitor the "A" Train BHUT level.
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| * Ensure capability for an operator to access motor-operated valves (MOVs) 1, 2, 3LP-21 (or 1, 2, 3LP-22) where 1, 2, 3DHR-GF1 &2 success paths are credited.
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| * Ensure capability for an operator to access manual valves 1, 2, 3HP-363 and 1, 2, 3HP-78, where 1,2, 31NVCTL3c success paths are credited.
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| Completion of each of these activities is an implementation item (SE Section 2.9, Table 2.9-1, Items 17 through 22, respectively).
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| Based on its review of the information provided in the LAR, the NRC staff concludes that the licensee used methods consistent with the interim guidance provided in FAQ 07-0040 and RG 1.205 to identify the equipment required to achieve and maintain the fuel in a safe and stable condition during NPO modes.
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| Components that were identified as needed to support an NPO KSF but were not included on the post-fire SSD equipment list required additional circuit analysis. The licensee loaded that information into the ARTRAK database, which allowed sorting of the component and cable information on a fire zone by fire zone basis. Utilizing the fire zone cable routing and equipment location information from ARTRAK, the licensee's evaluation of NPO fire impacts focused on analyzing the KSF success paths on a fire zone by fire zone basis in order to assess the impact of a single fire. Those fire zones with KSF success path impacts were identified and OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 97 categorized based on fire risk vulnerability. Recommendations to establish additional fire protection/fire prevention actions during HREs by fire zone were developed based on the assessed fire risk vulnerability. Due to the lack of rated fire barriers between all fire zones, additional fire prevention recommendations were made for fire zones where compartment to compartment interactions could potentially take place.
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| The licensee documented its analysis of the impact of a fire in each fire zone on the success paths for the KSFs, and recommendations of changes to fire risk and outage management procedures, in a site-specific calculation. Consistent with FAQ 07-0040, the recommendations of the site-specific NPO fire impact calculation apply only to those fire zones where fires could cause the complete loss of a KSF (pinch point). Fire modeling was not used to eliminate any fire zone from being a pinch point. Specific examples of recommendations include:
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| * Prohibition or limitation of hot work in fire zones during periods of increased vulnerability.
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| * Limitation of combustible materials in fire zones during periods of increased vulnerability.
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| * Plant configuration changes (e.g., removing power from equipment once it is placed in its desired position). The licensee states that it will develop procedures to realign and remove power from the MOVs in the unit-specific gravity feed flow paths prior to entering HREs to preclude spurious operation. Development of these procedures is an implementation item (SE Section 2.9, Table 2.9-1, Item 24).
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| * Provision of additional fire patrols at periodic intervals or other appropriate compensatory measures (such as surveillance cameras) during periods of increased vulnerability.
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| * Reschedule the work to a period with lower risk or higher DID.
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| * Crediting of committed modifications (e.g., PSW HPI System).
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| The licensee states that it will update the NPO evaluation of fire impacts on KSF success paths following the installation of the NFPA 805 committed modifications noted in SE Section 2.8.1.
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| Completion of this update is an implementation item (SE Section 2.9, Table 2.9-1, Item 23).
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| The licensee states that it does not currently rely on the use recovery actions to restore KSFs.
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| The licensee further states that recommendations resulting from its review will be incorporated into appropriate plant procedures prior to implementation of NFPA 805. Development of these procedure changes is an implementation item (SE Section 2.9, Table 2.9-1, Item 25).
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| In accordance with the method endorsed in NEI 04-02 and FAQ 07-0040, the primary mechanism being used to meet the nuclear safety performance criteria during NPO conditions is through the use of normal fire protection defense-in-depth (FP DID) actions to reduce the risk of fire. Specific examples include, but are not limited to, additional use of fire watch patrols or other appropriate measures (such as surveillance cameras) and establishing appropriate administrative controls to govern ignition sources, hot work and combustible materials. During HREs, this is achieved by implementing enhanced FP DID actions, specific examples of which are identified above, that reduce the frequency, severity or impact of fires such that the key pinch points are protected. During non-HREs, this is achieved by implementing the normal FP DID actions throughout the plant.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 98 The NRC staff reviewed the breaker coordination study. In letters dated August 3, 2009 (Reference 8) and April 14, 2010 (Reference 11), the licensee agreed that the coordination study currently relied on in the LAR, as supplemented, requires further enhancement to meet Section 2.4.2.2.2 of NFPA 805 and that a revised breaker coordination study was underway. By letter dated September 13, 2010 (Reference 12) the licensee stated that the revised breaker coordination study had been completed and identified modifications to four breakers that have an overall risk increase due to their lack of coordination with the upstream protective device.
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| The four breakers are being modified to maintain the Fire PRA risk profile reported in the LAR (see SE Section 3.4 for a more detailed discussion). The plant modifications are described in SE Section 2.8, Table 2.8.1-1.
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| In the cited supplementary information, the licensee stated that the revised study included the coordination of electrical protective devices associated with NPO/KSF power supplies and that any required modifications identified during the breaker coordination study were entered into its corrective action program and appropriate compensatory actions were implemented until the item is fully resolved. The licensee also stated that further analysis was performed for those feeders on the selected power supply, which were shown to be uncoordinated. The cables associated with these uncoordinated feeders were identified and routed by Fire Area in order to determine the impact to the associated Fire Areas/Scenarios. The cables associated with the uncoordinated feeders were documented in ARTRAK and will be utilized as an input to the NSCA, Fire PRA model, and the NPO Pinch Point Analysis.
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| In the LAR, the licensee also credited its original breaker coordination study to address common enclosure concerns. However, as discussed above, the original ONS coordination study does not satisfy applicable NFPA 805 or NEI 00-01 criteria. In Enclosure 3 of the LAR (Reference
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| : 11) the licensee states that the second phase of the revised coordination study included a review of the cable damage curves to determine if the electrical circuit design provides proper circuit protection in the form of circuit breakers, fuses and other devices that are designed to isolate cable faults before ignition temperature is reached. The results of this review were entered into the ARTRAK database and analyzed in the Fire Area/Fire Zone impacts. All power supplies required by the NSCA, Fire PRA, and !\IPO Pinch Point Analysis, as identified on the associated equipment list, were included in the breaker coordination study scope of "SSD related" power supplies. The licensee further stated that the coordination study meets the requirements of NFPA 805, Section 2.4.2.2.2, for circuits that share a common enclosure with circuits required to achieve nuclear safety performance criteria. In addition, the licensee states that a review of recent modifications confirms that adequate electrical circuit protection has been maintained as part of the design change process. In addition, the licensee states that the results of the coordination study will be documented in the NSCA, NPO Pinch Point Analysis, and the Fire PRA.
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| Incorporating information related to cables associated with uncoordinated feeder breakers of credited power supplies into the NSCA and NPO Pinch Point Analysis and updating the Fire PRA model to include the results of the breaker coordination study is an implementation item (SE Section 2.9, Table 2.9-1, Item 33). Updating the breaker coordination study to include all new NFPA 805 SSEL-related power supplies (Le., PSW) for power and non-power operations, and defining additional plant modification if necessary to ensure that the assumptions of the Fire PRA and NSCA remain valid, is an implementation item (SE Section 2.9, Table 2.9-1, Item 44).
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| The NRC staff also requested the licensee to provide an evaluation of spurious equipment actuations and/or mal-operations (including multiple spurious operations) during non-power operation modes. In its letter dated November 30, 2009 (Reference 10), the licensee states that OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 99 site-specific NPO calculations had been revised to provide a greater level of detail in explaining how spurious equipment actuations and/or mal-operations including multiple spurious operations have been analyzed in the evaluation of pinch points for NPO. Specifically, the component and cable selection process was revised to include all components with a potential for spurious operation including flow blockage and diversion and associated cables causing spurious operations. Any cable hit within a fire zone was considered an adverse impact on the component and any related KSF success path(s), which the licensee stated was conservative.
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| In its September 13, 2010 (Reference 12) letter, the licensee states that any and all potential spurious actuations that may result from intra-cable shorting were considered. Such failures were considered to occur concurrently, regardless of number, in accordance with the guidance provided in NEI 00-01, Section 3.5.1.5[B]. The NRC staff finds that the licensee's overall approach conforms with the endorsed guidance.
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| Conclusion for Section 3.5.2 NFPA 805 requires that the nuclear safety performance criteria be met during any operational mode or condition, including NPO. As described above, the licensee has performed the following engineering analyses to demonstrate that it meets this requirement:
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| * Identified the KSFs required to support the nuclear safety performance criteria during NPOs.
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| * Identified the POSs where further analysis is necessary during NPOs.
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| * Identified the equipment required to meet the KSFs during the POSs analyzed.
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| * Identified the location of this equipment and their associated cables.
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| * Performed analyses on a fire zone basis to identify pinch points where one or more KSFs could be lost as a direct result of fire-induced damage.
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| * Planned modifications to appropriate station procedures in order to employ one or more fire protection strategies for reducing risk at these pinch points during HREs.
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| In addition, normal FP DID actions are credited for addressing the risk impact of those fires which potentially affect one or more trains of equipment that provide a KSF required during NPO modes, but would not be expected to cause the total loss of that KSF. Accordingly, based on the information provided in the LAR as supplemented, the NRC staff concludes that the licensee has provided reasonable assurance that the nuclear safety performance criteria are met during NPO modes and HREs at ONS.
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| 3.5.3. Conclusion for SE Section 3.5 The NRC staff reviewed the licensee's RI/PB FPP, as described in the LAR and its supplements, to evaluate the NSCA results. The licensee used a combination of the deterministic approach in accordance with NFPA 805, Section 4.2.3, and the PB approach in accordance with NFPA 805, Section 4.2.4, to perform this assessment at ONS.
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| For those fire areas that utilized a deterministic approach, the NRC staff confirmed the following:
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 100
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| * None of the exemptions from the existing fire protection licensing basis were credited to meet the deterministic requirements in any of the deterministic fire areas.
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| * Fire suppression effects were evaluated and found to have no adverse impact on the ability to achieve and maintain the nuclear safety performance criteria for each fire area.
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| * No recovery actions were relied on to meet the deterministic requirements in any of the fire areas.
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| * The required automatic fire suppression and automatic fire detection systems were appropriately documented for each fire area.
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| The NRC staff found that each fire area utilizing the deterministic approach met the deterministic requirements of NFPA 805, Section 4.2.3.
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| For those fire areas that utilized the PB approach in accordance with NFPA 805 Section 4.2.4, the NRC staff confirmed that:
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| * Exemptions from the existing ONS fire protection licensing basis that were previously approved by the NRC, and which are being carried forward by the licensee into the RIIPB FPP, were evaluated and found to be valid and acceptable for meeting the deterministic requirements of NFPA 805 as allowed by NFPA 805, Section 2.2.7.
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| * Fire suppression effects were evaluated and found to have no adverse impact on the ability to achieve and maintain the nuclear safety performance criteria for each fire area.
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| * All VFDRs were evaluated using the FRE PB method (in accordance with NFPA 805, Section 4.2.4.2) to address risk impact, DID, and SMs and found to be acceptable.
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| * No recovery actions were necessary to demonstrate the availability of a success path.
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| * All recovery actions credited with providing DID were evaluated with respect to the additional risk presented by their use and found to be acceptable in accordance with NFPA 805, Section 4.2.4.
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| * The required automatic fire suppression and automatic fire detection systems were appropriately documented for each fire area.
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| Accordingly, the NRC staff has reasonable assurance that the nuclear safety performance criteria will be met for each fire area utilizing the PB approach, in accordance with NFPA 805, Section 4.2.4. Furthermore, the associated FREs meet the requirements for risk, DID, and SMs.
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| The NRC staff's review of the licensee's analysis for, and outage management process during, NPO modes found that the licensee provided reasonable assurance that the nuclear safety performance criteria will be met during NPO modes and HREs. The staff's review also found that the normal FPP DID actions are credited for addressing the risk impact of those fires which potentially affect one or more trains of equipment that provide a KSF required during NPO modes, but would not be expected to cause the total loss of that KSF. The NRC staff finds this overall approach for fire protection during NPO modes acceptable.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 101 3.6. Radioactive Release Performance Criteria I\JFPA 805, Chapter 1 defines the radioactive release goals, objectives, and performance criteria that must be met by the FPP in the event of a fire at a nuclear power plant:
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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.
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| Radioactive Release Objective Either of the following objectives shall be met during all operational modes and plant config urations.
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| (1) Containment integrity is capable of being maintained.
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| (2) The source term is capable of being limited.
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| 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.
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| The NRC staff reviewed LAR Section 4.4, "Radioactive Release Performance Criteria," and Attachment E, "NEI 04-02 Table G-1 Radioactive Release Transition," to evaluate the engineering and procedural controls credited by the licensee to limit potential radioactive releases to unrestricted areas associated with fire fighting activities.
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| In letter dated April 14, 2010, (Reference 11), the licensee stated that the current operating license for ONS, delineated in Technical Specification 5.5.5b, permits a liquid effluent release limit of 10 times that of 10 CFR Part 20, Appendix B, Table 2, Column 2, and that this NRC approved limit, (NRC Staff's SE dated January 6, 1993 (ADAMS Accession No. ML012040034)
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| (Reference 55), is the radioactive release performance criteria for liquid effluent releases for ONS. Per the introductory text to Part 20, Table 2, the concentration values given in Column 2 are equivalent to the radionuclide concentrations which, if inhaled or ingested continuously over the course of a year, would produce a total effective dose equivalent (TEDE) of 50 mrem (or 0.5 mSv). A 1-year release having radionuclide concentrations a factor of 10 times the Table 2, Column 2 values would produce a TEDE of 500 mrem (or 5 mSv), the maximum allowed radiation dose limit to individual members of the public permitted by 10 CFR Part 20, Section 1301 (d). Since the liquid effluent release limits permitted by the ONS operating license limits is equivalent to the 10 CFR Part 20 maximum permissible radiation dose to the general public, the NRC staff considers the licensee's liquid effluent release limit of 10 times the 10 CFR Part 20, Appendix B, Table 2, Column 2 equivalent to the NFPA 805 radioactive release performance criteria and therefore acceptable. In response to an NRC staff RAI, the licensee stated that the release limits for gaseous effluents at ONS, including smoke release, conform to the NFPA 805 radioactive release performance criteria (Reference 12).
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 102 In order to assess whether the ONS FPP to be implemented under NFPA 805 meets the above requirements, the licensee reviewed the existing ONS pre-fire plans and fire brigade training materials. Pre-fire plans that address fire areas where there is no possibility of radioactive materials being present were screened from further review. All other fire zone pre-fire plans were then evaluated to ensure that the locations that have the potential for radioactive release due to fire fighting activities are subject to specific steps for containment and monitoring of potentially contaminated smoke and fire suppression water. Engineering and procedural controls for water release and smoke were then reviewed to determine how effectively the specific steps in the pre-fire plans provide guidelines for the containment and monitoring for potentially contaminated smoke and fire suppression water.
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| The licensee's review determined the current FPP is compliant with the requirements of NFPA 805 and the guidance in RG 1.205, with the exception of the 10 CFR Part 20 limits for liquid effluent. As discussed above, the NRC staff considers the NRC-approved liquid effluent release limits in the ONS operating license to be in conformance with the maximum permissible dose limits of 10 CFR Part 20. In addition, the licensee stated that during the radioactive release review, a new 'fire brigade SOG-16 was developed to address smoke manqgement and potentially contaminated water runoff when a fire involves potentially contaminated areas that may not be identified on the pre-fire plans (Reference 11). These areas may include other radioactively contaminated areas that have been established for short-term periods, such as outages and maintenance evolutions.
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| Table 3.6-1, "ONS Fire Areas and Their Compliance with the NFPA 805 Radioactive Release Performance Criteria," in Attachment E to this SE summarizes, for each fire pre-plan, (1) the fire zone included in the pre-plan, (2) the engineered controls used to minimize radioactive releases generated from the combustion of radioactive materials or from fire suppression activities, and (3) the NRC staffs evaluation of the adequacy of the licensee's methods of controlling and monitoring contaminated suppression agent runoff and combustion smoke.
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| The licensee also reviewed the fire brigade training materials to ensure they are consistent with the pre-fire plans in terms of containment and monitoring of potentially contaminated smoke and fire suppression water. The licensee's review determined that the existing fire brigade training materials are adequate. In addition, the new SOG-16 described above has been fully implemented into the ONS fire brigade training program.
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| NFPA 805 requires the licensee to address the nuclear safety and radioactive release goals, objectives and performance criteria in any operational mode. The licensee stated that the radioactive release review was not performed based on plant operating modes, since fire suppression activities, as defined in the pre-fire plans and fire brigade fire fighting instruction operating guidelines, are written for any plant operating mode. During non-power operational modes, the licensee stated that the fire pre-fire plans conservatively assume an "at power" entry condition and do not differentiate between operating and shutdown conditions. In addition, as described previously, a new fire brigade SOG-16 was developed to address smoke management and potentially contaminated water runoff when a fire involves potentially contaminated areas that have been established for short-term periods such as outages.
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| In letter dated April 14, 2010, (Reference 11), the licensee stated that the existing pre-fire plans adequately address containing and monitoring products of combustion generated from a fire in the RB while the equipment hatch is open.
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| In letter dated April 14, 2010, (Reference 11), the license stated that the pre-fire plans are controlled documents under the licensee's procedures and within the scope of the configuration OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE O~JLY SECldRITY RELATED INFORMATION 103 management process. The licensee also stated that the results of the radioactive release reviews will be maintained post-transition by the established ONS Configuration Management Program as described in LAR Section 4.7. (Note: SE Section 3.8 contains the NRC staff's review of the licensee's configuration management processes.)
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| Based on (1) the information provided in the LAR as supplemented, (2) the licensee's use of pre-fire plans, (3) the results of the NRC staff's evaluation of the identified engineered controls used to control suppression water and combustion products, (4) the development and implementation of a new fire brigade operating guideline regarding control of radiological release, and (5) fire brigade training on monitoring and controlling suppression water runoff and combustion smoke, the NRC staff concludes that the licensee's RI/PB FPP provides reasonable assurance that radiation releases to any unrestricted area due to the direct effects of fire suppression activities at ONS are as low as reasonably achievable and are not expected to exceed the radiological dose limits in 10 CFR Part 20 and the NRC-approved liquid effluent release limit of 10 times that of 10 CFR Part 20, Appendix B, Table 2, Column 2, allowed for in the ONS operating license. In conclusion, the NRC staff finds that the licensee's RI/PB FPP approach aligns with the goals, objectives, and criteria specified in NFPA 805 Sections 1.3.2, 1.4.2, and 1.5.2 and is acceptable.
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| 3.7. Monitoring Program For this section of the SE, the following requirements from NFPA 805, Section 2.6, are applicable to the NRC staff's review of the licensee's amendment request:
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| Monitoring. A monitoring program shall be established to ensure that the availability and reliability of the fire protection systems and features are maintained and to assess the performance of the FPP in meeting the performance criteria. Monitoring shall ensure that the assumptions in the engineering analysis remain valid.
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| Availability, Reliability, and Performance Levels. Acceptable levels of availability, reliability, and performance shall be established.
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| Monitoring Availability, Reliability, and Performance. Methods to monitor availability, reliability, and performance shall be established. The methods shall consider the plant operating experience and industry operating experience.
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| Corrective Action. If the established levels of availability, reliability, or performance are not met, appropriate corrective actions to return to the established levels shall be implemented. Monitoring shall be continued to ensure that the corrective actions are effective.
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| The NRC staff reviewed the ONS NFPA 805 Monitoring Program described in LAR Section 4.6, "Monitoring Program" (Reference 11), that the licensee is developing to monitor availability, reliability, and performance of ONS FPP systems and features after the transition to NFPA 805.
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| While the program was still under development at the time the LAR was submitted for review, the focus of the NRC staff's evaluation involved identifying the critical elements related to the 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 program will occur on the same schedule as the NFPA 805 RI/PB FPP implementation. Completion of the ONS NFPA 805 Monitoring Program is an implementation item (SE Section 2.9, Table 2.9-1, Item 8).
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 104 The licensee is developing an ONS-specific calculation to document and describe the methodology and criteria used to select fire protection systems and features for inclusion in the ONS NFPA 805 Monitoring Program. By letter dated September 27,2010 (Reference 13), the licensee provided a detailed description of the methodology and criteria and stated that a multi disciplinary review team is being utilized to review and check the calculation. The licensee's review team includ'es representatives from operations, fire protection, PRA, and SSD, all of whom are experienced and qualified to the licensee's training program for their positions.
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| The scope of the licensee's monitoring program includes FPP Systems, Structures, and Components (SSCs) and FPP programmatic elements. Supporting engineering evaluations associated with the NFPA 805 transition effort were reviewed to identify SSCs and programmatic elements credited in these supporting evaluations for providing some functional role in reducing fire risk. All SSCs that perform functions or support assumptions credited in the NFPA 805 engineering evaluations to reduce fire risk are evaluated to determine if additional monitoring of the component is required.
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| The licensee stated that SSCs necessary to meet the NFPA 805 nuclear safety performance criteria are typically monitored as part of Maintenance Rule monitoring (as promulgated in 10 CFR 50.65). Accordingly, the NRC staff finds that the licensee may use the Maintenance Rule for the components covered by that program as a means to meet the requirements of the ONS NFPA 805 Monitoring Program. As such, these systems and equipment will not be included in the ONS NFPA 805 Monitoring Program. However, the licensee will review the NSCA SSCs credited in the NFPA 805 analyses to validate that availability and reliability is monitored as part of the Maintenance Rule and that the criteria is adequate to meet the needs of NFPA 805. If the criteria are not adequate, new Maintenance Rule functions will be created and applied to support NFPA 805 monitoring requirements (Reference 52). Those systems and equipment required to meet the nuclear safety performance criteria that are not included in the Maintenance Rule monitoring program will be reviewed for inclusion in the NFPA 805 Monitoring Program.
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| The SSCs and programmatic elements to be included in either monitoring program will be monitored for availability and reliability to ensure that the functions credited will be accomplished as assumed in the supporting engineering evaluations. Since a credited function may be performed by a number of individual components for a given fire area, the licensee is establishing Performance Monitoring Groups (PMGs) for each fire zone. PMGs are functional categories of fire protection systems and administrative controls. The table provided in the licensee's letter dated September 27, 2010 (Reference 13) provides the initial list of ONS PMGs.
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| The licensee has defined screening thresholds, which are being used to determine the most risk-significant fire compartments utilizing the results of the Fire PRA. Those fire compartments (and all PMGs within the compartments) that are determined to be risk significant will be brought into the scope of the ONS NFPA 805 Monitoring Program. In response to a supplemental NRC staff RAI (Reference 52), the licensee identified the following screening thresholds being used to determine either the fire compartments or components, or both, to be included in the scope of the ONS NFPA 805 Monitoring Program:
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| * CDF greater than or equal to 1.0E-07 per year (on a compartment basis)
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| * LERF greater than or equal to 1.0E-08 per year (on a compartment basis)
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| * risk achievement worth (RAW) greater than or equal to 2 (on a PMG)
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 105 The licensee has defined High Safety Significant (HSS) fire zones and SSCs as those that exceed the screening criteria. The licensee stated that all FPP SSCs that are in HSS fire zones, and all HSS FPP components, that are amenable to risk measurement will be included in the ONS NFPA 805 Monitoring Program.
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| The screening criteria being implemented at ONS in regard to the ONS NFPA 805 Monitoring Program are acceptable to the NRC staff based on the following: (1) the CDF and LERF criteria used to screen compartments into the program are consistent with the self approval limits under the RI/PB FPP license condition (see SE Section 4.0), and (2) the NRC staff has previously determined the RAW criteria used for screening individual PMG into the program to be acceptable for use in determining risk significant SSCs that must be monitored under the Maintenance Rule, as described in NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants" (Reference 32).
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| The licensee also stated that it will establish criteria for acceptable levels of availability, reliability, and performance, or appropriate action levels, for each PMG. The intent is to establish conservative values of availability and reliability, such that assumptions made in the applicable supporting analyses are bounded. Target and action levels for availability will be primarily based on site-specific data reflecting expected out-of-service times to support maintenance and inspection activities, such that planned impairments with appropriate functional compensatory measures will not be assessed against availability criteria (Reference 52). Target and action values for reliability will be based primarily on industry guidance in EPRI Technical Report (TR) 1006756, "Fire Protection Surveillance Optimization and Maintenance Guide for Fire Protection Systems and Features" (Reference 53), with adjustments to reflect site-specific operating experience, Fire PRA assumptions, and equipment types (and vendor data when available). However, in response to an NRC staff RAI (Reference 52), the licensee stated that availability and reliability targets for NFPA 805 monitored SSCs will be selected, reviewed, and maintained to ensure that the assumptions of the applicable supporting analyses (e.g., Fire PRA, NSCA, etc.) remain valid. Performance of programmatic elements such as fire brigade, fire-watches, and combustible controls will be evaluated using the existing ONS plant health process.
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| The NRC staff finds that establishing availability target and action levels using site-specific data, and reliability targets and action levels in accordance with EPRI TR 1006756, in conjunction with setting availability and reliability targets for NFPA 805 monitored SSCs to ensure that the assumptions made in the Fire PRA and other supporting analyses will remain valid, is acceptable. The method for establishing appropriate levels of availability, reliability, and performance because there will be margin between the value assumed in the Fire PRA for a given component or system and the action level used in the ONS NFPA 805 Monitoring Program to require corrective action.
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| The licensee further stated that inspection and test frequencies being used to gather data to assess the availability and reliability of PMGs will be those currently contained in the ONS Selected Licensee Commitments (SLC's), which are contained in ONS UFSAR Chapter 16.
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| The licensee stated that as more performance data is obtained, frequencies may be adjusted using the PB process described in EPRI 1006756, the Nuclear Electric Insurance Limited (NEIL) underwriting guidelines, and applicable NFPA codes such as NFPA 72, "National Fire Alarm Code."
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 106 The NRC staff finds that establishing monitoring frequencies initially using those contained in SLCs is acceptable since the licensee did not identify any changes to ONS UFSAR Chapter 16 in the LAR, as supplemented. The staff also finds the PB methods for establishing monitoring frequencies described in EPRI 1006756, NEIL underwriting guidelines, and applicable NFPA codes acceptable for this NFPA 805 RI/PB FPP.
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| In addition, inspection and test acceptance criteria will be developed for each PMG that will determine when a system has failed to perform its required function. Initially, these criteria will be based on the system design, manufacturer criteria, and NFPA code requirements. However, the criteria may require adjustment by the system/program engineer or multidisciplinary review team as the program becomes established and monitoring data is gathered over a period of time. However, the values of availability and reliability data will be reviewed to ensure they remain bounding for the assumptions made in the applicable supporting analyses.
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| The licensee stated that a software program is being developed to collect applicable reliability and availability data and will provide alerts if target values are approached. Developing instructions for the software program is an implementation item (SE Section 2.9, Table 2.9-1, Item 37). The data will be periodically evaluated by the appropriate system or program engineers. Failure to meet availability and/or reliability criteria results in the initiation of the ONS Problem Investigation Process (PIP) to establish performance goals and corrective actions to return the component or PMG into compliance with the established criteria.
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| As described above, NFPA 805, Section 2.6, requires that a monitoring program be established in order to ensure that the availability and reliability of fire protection systems and features are maintained, as well as to assess the overall effectiveness of the FPP in meeting the performance criteria. Monitoring should ensure that the assumptions in the associated engineering analysis remain valid. Based on the information provided in the LAR as supplemented, the NRC staff finds that the licensee's process provides reasonable assurance that the licensee will implement an effective program for monitoring risk-significant fire SSCs because the multi-disciplinary review team ensures that the ONS NFPA 805 Monitoring Program does the following:
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| * Establishes the appropriate performance monitoring groups to be monitored.
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| * Utilizes an acceptable screening process for determining the SSCs to be included in the PMGs.
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| * Establishes availability, reliability and performance criteria for the SSCs being monitored.
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| * Requires corrective actions when SSC availability, reliability, and performance criteria targets are exceeded in order bring performance back within the required range.
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| 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 program as of the date of this SE, completion of the ONS NFPA 805 Monitoring Program is an implementation item, as noted previously. Completion of the program will occur on the same schedule as the implementation of NFPA 805, which the NRC staff finds acceptable. Accordingly, the NRC staff concludes that, upon successful closure of this implementation item, there is reasonable assurance that the licensee will meet the requirements specified in NFPA 805, Sections 2.6.1, 2.6.2, and 2.6.3 regarding a monitoring program.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 107 3.8. Program Documentation, Configuration Control, and Quality Assurance This section of the SE documents the NRC staff's review in regard to the appropriate content, configuration control, and quality of the documentation used to support the transition to NFPA 805 at ONS.
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| 3.8.1. Documentation The NRC staff reviewed LAR Section 4.7.1 (Reference 11) to evaluate the appropriateness of the content of the ONS FPP DBD and supporting documentation.
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| ONS's FPP design basis is a compilation of multiple documents (such as analyses, calculations and engineering evaluations), databases, and drawings that are identified in Figure 4-8 of the LAR. ONS has documented analyses to support compliance with 10 CFR 50.48(c). The licensee stated that analyses performed to support the NFPA 805 transition were performed in accordance with the licensee's processes for ensuring assumptions are clearly defined, that results be easily understood,. clearly and consistently described, and that sufficient detail be provided to allow future review of the entire analyses, as required in NFPA 805, Section 2.7.1.
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| The ONS FPP DBD and necessary supporting documentation described in Section 2.7.1 of NFPA 805 will be revised as part of transition. Completion of the revisions to the ONS DBD and supporting documentation is an implementation item (SE Section 2.9, Table 2.9-1, Item 45).
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| The licensee stated in Section 4.7.1 of the LAR that documentation associated with the ONS RI/PB FPP will be maintained for the life of the plant and organized to facilitate review for accuracy and adequacy by independent reviewers and by the NRC staff. Based on the description of the content of the ONS FPP design basis and supporting documentation, and the licensee's plans to maintain this documentation throughout the life of the plant, the NRC staff finds that the licensee's approach meets the requirements of NFPA 805, Sections 2.7.1.1, 2.7.1.2, and 2.7.1.3 regarding adequate development and maintenance of the FPP DBD, and is therefore acceptable.
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| 3.8.2. Configuration Control The NRC staff reviewed LAR Section 4.7.2, (Reference 11). The licensee stated that program documentation established, revised, or utilized in support of compliance with 10 CFR 50.48(c) was subject to the licensee's configuration control processes that meet the requirements of Section 2.7.2 of NFPA 805. This includes the appropriate procedures and configuration control processes for ensuring that changes potentially impacting the FPP are reviewed.
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| In a letter dated September 27,2010 (Reference 13), the licensee stated that configuration control of RI/PB documents before and during the transition period is managed using EC procedures. These procedures were modified to include evaluation criteria for implementing design changes that specifically relate to attributes that may impact NFPA 805. More detailed reviews would be required if these evaluation criteria indicate impact. These reviews would be conducted by qualified fire protection, SSD, and PRA personnel involved in the ongoing transition work for NFPA 805.
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| The plant processes described above, will be in place during the NFPA 805 transition to identify changes that may impact the FPP. Additionally, a comprehensive update of the NFPA 805 OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 108 analyses is planned as part of the NFPA 805 implementation period to reflect the current plant configurations. The update will include review of plant configuration changes along with changes that may have occurred from RAI responses, updates from industry groups for MSO configurations, new or revised FAQ's, and development of the PSW modification. This final review will ensure current plant configurations are appropriately reflected and evaluated in the NFPA 805 documentation prior to full implementation of NFPA 805.
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| The licensee further stated that the ONS FPP change evaluation procedure will be updated during the implementation period to address the NRC-approved NFPA 805 change evaluation process and that configuration control processes and procedures will be updated during the transition period to manage configuration control of the NFPA 805 designllicensing-basis documents. Revision of these procedures is an implementation item (SE Section 2.9, Table 2.9-1, Item 27).
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| The NRC staff reviewed the licensee's description of the process for updating and maintaining the Fire PRA to reflect plant changes made after the transition to the NFPA has been completed in SE Section 3.4.1. Based on the licensee's description of the ONS configuration control process, that ONS RIIPB FPP design basis and supporting documentation are controlled documents, and that plant changes are reviewed for impact on the FPP, the NRC staff finds that the licensee has a configuration control process that aligns with the requirements of NFPA 805, Sections 2.7.2.1 and 2.7.2.2 for revising FPP DBDs, supporting documents, and applicable FPP documentation to reflect changes made to the RIIPB FPP after the NFPA 805 FPP has been implemented.
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| 3.8.3. Quality The NRC staff reviewed LAR Section 4.7.3, (Reference 11) to evaluate the quality of the engineering analyses used to support the transition to the NFPA 805 and to support post transition FPP activities at ONS. During the transition to 10 CFR 50.48(c), the licensee performed work in accordance with the quality requirements of Section 2.7.3 of NFPA 805.
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| Quality requirements from NFPA 805 that are not currently part of the licensee's processes will be revised to include any additional requirements. Revision of these quality requirements is an implementation item (SE Section 2.9, Table 2.9-1, Item 29).
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| NFPA 805 requires that each analysis, calculation, or evaluation performed be independently reviewed. The licensee stated that their analyses, calculations, and evaluations performed in support of compliance with 10 CFR 50.48(c) are performed in accordance with the licensee's procedures that require independent review. The licensee also stated that future changes to the FPP will follow the guidance outlined in RG 1.174 (Reference 15) which provides for the use of qualified individuals, procedures that require calculations 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.
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| Based on the licensee's description of the ONS process for performing independent reviews of analyses, calculations, and evaluations, the NRC staff finds the licensee's approach to meeting the requirements of NFPA 805, Section 2.7.3.1 acceptable.
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| OFFICIAL USE O~JLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 109 Verification and Validation NFPA 805 requires that each calculation model or numerical method used be verified and validated through comparison to test results or other acceptable models. The licensee stated in the LAR that calculation models and numerical methods used in support of compliance with 10 CFR 50.48(c) are verified and validated as required by Section 2.7.3.2 of NFPA 805.
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| The licensee also stated that it will revise the appropriate processes and procedures to include any additional NFPA 805 quality requirements that are not currently part of the licensee's processes for use during the performance of post-transition FPP changes. Revision of the applicable post transition processes and procedures to include the NFPA 805 requirements for verification and validation is an implementation item (SE Section, 2.9 Table 2.9-1, Item 29).
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| Based on the licensee's description of the ONS process for verification and validation of calculation models and numerical methods, the NRC staff finds the licensee's approach to meeting the requirements of NFPA 805, Section 2.7.3.2 acceptable.
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| Limitations of Use NFPA 805 requires that acceptable engineering methods and numerical models only be used for applications to the extent that these methods have been subject to verification and validation; and that they only be applied within the scope, limitations, and assumptions prescribed for that method. The licensee stated that the engineering methods and numerical models used in support of the transition to NFPA 805 were used subject to the limitations of use outlined in NFPA 805, Section 2.7.3.3, and that the engineering methods and numerical models used post-transition will be subject to these same limitations of use. As an example, in LAR Section 4.5.2, the licensee stated that the fire models developed to support the NFPA 805 transition at ONS fall within their verification and validation limitations. The licensee also stated that it will revise the appropriate processes and procedures to include any additional NFPA 805 quality requirements that are not currently part of the licensee's processes for use during the performance of post-transition FPP changes. Revision of the applicable post transition processes and procedures to include the NFPA 805 requirements for limitations of use is an implementation item (SE Section 2.9, Table 2.9-1, Item 29).
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| Based on the licensee's description of the ONS process for placing limitations on the use of engineering methods and numerical models, the NRC staff finds the licensee's approach to meeting the requirements of NFPA 805, Section 2.7.3.3 acceptable.
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| Qualification of Users NFPA 805 requires that personnel performing engineering analyses and numerical methods (e.g. fire modeling) 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. The licensee has stated that, during the transition to 10 CFR 50.48(c), work will be performed in accordance with the quality requirements of Section 2.7.3 of NFPA 805. It also stated that post-transition quality requirements from NFPA 805 that are not currently part of the licensee's processes will be revised to include any additional requirements. Revision of the applicable post transition processes and procedures to include any additional NFPA 805 requirements is an implementation item (SE Section 2.9, Table 2.9-1, Item 29). Also, the licensee stated that cognizant personnel who use and apply engineering analyses and OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 110 numerical methods in support of compliance with 10 CFR 50.48(c) are competent and experienced as required by Section 2.7.3.4 of NFPA 805.
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| For personnel performing fire modeling or fire PRA development and evaluation, the licensee has qualification requirements for individuals assigned various tasks. Position specific guides will be developed to identify and document required training and mentoring to ensure individuals, both employees of the licensee and subcontractors, are appropriately qualified per the requirements of NFPA 805, Section 2.7.3.4 to perform assigned work. Development of these position-specific guides is an implementation item (SE Section 2.9, Table 2.9-1, Item 28).
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| Based on the licensee's description of the ONS procedures for ensuring that the personnel who use and apply engineering analyses and numerical methods, including those who develop the ONS Fire PRA and perform fire modeling calculations, are competent and experienced, the NRC staff finds the licensee's approach for meeting the requirements of NFPA 805, Section 2.7.3.4, acceptable.
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| Uncertainty Analysis NFPA 805 requires that an uncertainty analysis be performed to provide reasonable assurance that the performance criteria have been met. (Note: 10 CFR 50.48(c)(2)(iv) states that an uncertainty analysis performed in accordance with NFPA 805, Section 2.7.3.5, is notrequired to support calculations used in conjunction with a deterministic approach.) When using the PB methods, the licensee stated that uncertainty analyses were performed for the analyses used in support of the transition to NFPA 805, and that uncertainty analyses will be performed for post transition analyses. Based on the licensee's description of the ONS process for performing uncertainty analyses, the NRC staff finds the licensee's approach to meeting the requirements of NFPA 805, Section 2.7.3.5 acceptable.
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| Conclusion for Section 3.8.3 Based on the above discussions, the NRC staff finds that the RIIPB FPP quality assurance process adequately addresses each of the requirements of NFPA 805, Section 2.7.3:
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| conducting independent reviews, performing verification and validation (V&V), limiting the application of acceptable methods and models to within prescribed boundaries, ensuring that personnel applying acceptable methods and models are qualified for performing uncertainty analyses. The NRC staffs evaluation of the application of the NFPA 805 quality assurance requirements in the licensee's LAR is provided in the individual sections of this SE, where appropriate.
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| 3.8.4. Fire Protection Quality Assurance Program GDC 1 of Appendix A to 10 CFR Part 50 requires:
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| 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.
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| The licensee's Fire Protection Quality Assurance Program was established in accordance with the guidelines of Appendix A to Branch Technical Position (BTP) APCSB 9.5-1, Section C, "Quality Assurance Program," (Reference 51) and associated NRC guidance. In addition, the guidance in Appendix C to NEJ 04-02 (Reference 21) suggests that the LAR include a OFFICI,I\L USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 111 description of how the existing fire protection quality assurance (QA) program will be transitioned to the new NFPA 805 RI/PB FPP, as discussed below.
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| The licensee stated in the LAR that it will maintain its current fire protection QA program after transition to the NFPA 805 RI/PB FPP. In response to an NRC staff RAI, the licensee further stated that it did not foresee any substantive changes to the existing FPP QA Program (Reference 10).
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| Based on the licensee's statement that the ONS FPP QA Program will be maintained after transition to the NFPA 805 RI/PB FPP, the NRC staff finds that the scope of the fire protection QA program will include the fire protection systems that are required by NFPA 805, Chapter 4, and is therefore acceptable.
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| 3.8.5. Conclusion for Section 3.8 The NRC staff reviewed the licensee's RIIPB FPP, as described in the LAR and its supplements, to evaluate the NFPA 805 program documentation content, the associated configuration control process, and the appropriate QA requirements. The NRC staff concludes that, upon completion of the implementation items related to these requirements, the licensee's approach meets the requirements specified in NFPA 805, Section 2.7, regarding program documentation, configuration control, and quality.
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| 4.0 FIRE PROTECTION LICENSE CONDITION In the April 14, 2010 LAR the licensee proposed a FPP license condition regarding transition to NFPA 805, in accordance with 10 CFR 50.48(c)(3)(i). The proposed license condition adopted parts of the standard fire protection license condition promulgated in RG 1.205, Revision 1, Regulatory Position C.3.1, (Reference 14). The licensee made plant-specific changes to the sample license condition. The proposed license condition also requested self-approval of quantitative risk-informed fire protection program changes. By letter dated December 22, 2010 (Reference 59), the licensee replaced the original proposed license condition with a new license condition. The new proposed license condition did not request self-approval of quantitative risk informed fire protection program changes. The new proposed license condition requires the licensee to request NRC review and approval in accordance with 10 CFR 50.90 prior to being allowed to self approve quantitative risk-informed fire protection program changes except for those associated with the implementation items listed in Table 2.9-1 needing a plant change evaluation provided the overall transition risk remains a decrease.
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| The new proposed plant-specific FPP license condition is consistent with the standard fire protection license condition promulgated in RG 1.205; Revision 1. The NRC staff has reviewed the proposed license condition and finds that it incorporates all of the relevant features of the license condition published in RG 1.205 to allow transition to NFPA 805 at ONS. The NRC staff therefore finds the licensee's proposed license condition acceptable.
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| Implementation of the RI/PB FPP under 10 CFR 50.48(c) will be through the application of a new FPP license condition. As part of the implementation of this license amendment, the licensee shall complete all commitments in Tables 2.8.1-1 and 2.9-1 listed in Sections 2.8 and 2.9, respectively, of this SE. The NRC staff considered the above item in the tables as part of its evaluation, and finds the commitments appropriate for transitioning to the RI/PB FPP. The NRC staff has conditioned the implementation of the proposed transition on completion of the commitments in Tables 2.8.1-1 and 2.9-1. The new license condition also establishes the date OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 112 by which full compliance with 10 CFR 50.48(c) will be achieved. In addition, the license condition also dictates the licensee's required actions and restrictions until the licensee is able to fully implement the new FPP in accordance with 10 CFR 50.48(c)(3)(i).
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| The new fire protection license condition will replace the existing fire protection license condition in each unit's license. As a result, the NRC will be reissuing license pages 2 through 11 in each unit's license because of pagination. The only changes to the license are the changes to the fire protection license condition.
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| 5.0
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| ==SUMMARY==
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| | |
| Based on the above evaluation of the licensee's application, as supplemented, the NRC staff finds, the transition to a risk-informed, performance-based FPP in accordance with the requirements established by 10 CFR 50.48(c) and NFPA 805 as incorporated therein is acceptable. The NRC staff concludes that the licensee's approach, methods, and data are acceptable to establish, implement, and maintain a RI/PB FPP in accordance with 10 CFR 50.48(c).
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| ==6.0 STATE CONSULTATION==
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| In accordance with the Commission's regulations, the South Carolina State official was notified of the proposed issuance of the amendments. The State official had no comments.
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| ==7.0 ENVIRONMENTAL CONSIDERATION==
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| The amendments change a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on October 28, 2010 (75 FR 66395). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
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| Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
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| ==8.0 CONCLUSION==
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| 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) 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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| OFFICIAL USE O~JLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 113
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| ==9.0 REFERENCES==
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| : 1. NFPA 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants", 2001 Edition, National Fire Protection Association, Quincy, MA.
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| : 2. Letter from Ronald A. Jones, Duke Energy, to the U. S. Nuclear Regulatory Commission Document Control Desk. May 30, 2008.
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| ==Subject:==
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| "Duke Energy Carolinas, LLC, Oconee Nuclear Site Units 1, 2, and 3, Docket Numbers 50-269,50-270 and 50-287, License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition), License Amendment Request (LAR) No. 2008-01." Charlotte, NC. Available at Agencywide Document Access and Management System (ADAMS) Accession No. ML081650475 and ML082041014.
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| : 3. Letter from Dave Baxter, Duke Energy, to the U. S. Nuclear Regulatory Commission Document Control Desk. June 30, 2008.
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| | |
| ==Subject:==
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| "Duke Energy Carolinas, LLC, Oconee Nuclear Site Units 1,2, and 3, Docket Numbers 50-269,50-270 and 50-287, Completion Schedule for Part 2 of License Amendment Request to Adopt NFPA 805 Performance Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition), License Amendment Request (LAR) No. 2008-01." Seneca, SC. Available at ADAMS Accession No. ML081890193.
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| : 4. Letter from Dave Baxter, Duke Energy, to the U. S. Nuclear Regulatory Commission Document Control Desk. October 21,2008.
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| | |
| ==Subject:==
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| "Duke Energy Carolinas, LLC, Oconee Nuclear Site Units 1, 2, and 3, Docket Numbers 50-269,50-270 and 50-287, License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition), License Amendment Request (LAR) No. 2008-01." Seneca, SC. Available at ADAMS Accession No.
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| ML091170546.
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| : 5. Letter from Dave Baxter, Duke Energy, to the U. S. Nuclear Regulatory Commission Document Control Desk. February 9, 2009.
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| | |
| ==Subject:==
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| "Duke Energy Carolinas, LLC, Oconee Nuclear Site Units 1, 2, and 3, Docket Numbers 50-269,50-270 and 50-287, License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition), License Amendment Request (LAR) No. 2008-01." Seneca, SC. Available at ADAMS Accession No. ML090480143.
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| : 6. Letter from Dave Baxter, Duke Energy, to the U. S. Nuclear Regulatory Commission Document Control Desk. February 23, 2009.
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| | |
| ==Subject:==
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| "Duke Energy Carolinas, LLC, Oconee Nuclear Site Units 1,2, and 3, Docket Numbers 50-269,50-270 and 50-287, License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition), License Amendment Request (LAR) No. 2008-01." Seneca, SC. Available at ADAMS Accession No.
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| ML090700134.
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| : 7. Letter from Dave Baxter, Duke Energy, to the U. S. Nuclear Regulatory Commission Document Control Desk. May 31,2009.
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| | |
| ==Subject:==
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| "Duke Energy Carolinas, LLC, Oconee Nuclear Site Units 1,2, and 3, Docket Numbers 50-269,50-270 and 50-287, Additional Information regarding Modifications in support of License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition), License Amendment Request (LAR) No. 2008-01."
| |
| Seneca, SC. Available at ADAMS Accession No. ML091590045.
| |
| OFFICIAL USE ONLY SECURITY RELJ\TED INFORMATION
| |
| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 114
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| : 8. Letter from Dave Baxter, Duke Energy, to the U. S. Nuclear Regulatory Commission Document Control Desk. August 3, 2009.
| |
| | |
| ==Subject:==
| |
| "Duke Energy Carolinas, LLC, Oconee Nuclear Site Units 1, 2, and 3, Docket Numbers 50-269, 50-270 and 50-287, Request for Additional Information regarding the License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition), License Amendment Request (LAR) No. 2008-01." Seneca, SC. Available at ADAMS Accession No. ML092190212.
| |
| : 9. Letter from Dave Baxter, Duke Energy, to the U. S. Nuclear Regulatory Commission Document Control Desk. September 29, 2009.
| |
| | |
| ==Subject:==
| |
| "Duke Energy Carolinas, LLC, Oconee Nuclear Site Units 1,2, and 3, Docket Numbers 50-269, 50-270 and 50-287, Request for Additional Information regarding the License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition), License Amendment Request (LAR) No. 2008-01."
| |
| Seneca, SC. Available at ADAMS Accession No. ML092740624.
| |
| : 10. Letter from Dave Baxter, Duke Energy, to the U. S. Nuclear Regulatory Commission Document Control Desk. November 30, 2009.
| |
| | |
| ==Subject:==
| |
| "Duke Energy Carolinas, LLC, Oconee Nuclear Site Units 1,2, and 3, Docket Numbers 50-269, 50-270 and 50-287, Request for Additional Information regarding the License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition), License Amendment Request (LAR) No. 2008-01."
| |
| Seneca, SC. Available at ADAMS Accession No. ML093410007.
| |
| : 11. Letter from Dave Baxter, Duke Energy, to the U. S. Nuclear Regulatory Commission Document Control Desk. April 14, 2010.
| |
| | |
| ==Subject:==
| |
| "Duke Energy Carolinas, LLC, Oconee Nuclear Site Units 1,2, and 3, Docket Numbers 50-269,50-270 and 50-287, License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition), License Amendment Request (LAR) No. 2008-01." Seneca, SC. Available at ADAMS Accession Package No.
| |
| ML101121042.
| |
| : 12. Letter from Dave Baxter, Duke Energy, to the U. S. Nuclear Regulatory Commission Document Control Desk. September 13, 2010.
| |
| | |
| ==Subject:==
| |
| "Duke Energy Carolinas, LLC, Oconee Nuclear Site Units 1, 2, and 3, Docket Numbers 50-269, 50-270 and 50-287, Request for additional Information regarding the License Amendment Request to adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition), License Amendment Request (LAR) No. 2008-01."
| |
| Seneca, SC. Available at ADAMS Accession No. ML102640110.
| |
| : 13. Letter from Dave Baxter, Duke Energy, to the U. S. Nuclear Regulatory Commission Document Control Desk. September 27,2010. Subject "Duke Energy Carolinas, LLC, Oconee Nuclear Site Units 1,2, and 3, Docket Numbers 50-269,50-270 and 50-287, Request for additional Information regarding the License Amendment Request to adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition), License Amendment Request (LAR) No. 2008-01."
| |
| Seneca, SC. Available at ADAMS Accession No. ML102720409 OFFICIAL USE ONLY SECURITY RELATED INFORMATION
| |
| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 115
| |
| : 14. Regulatory Guide (RG) 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," Revision 1, U. S. Nuclear Regulatory Commission, Washington, DC, December 2009. Available at ADAMS Accession No.
| |
| ML092730314.
| |
| : 15. Regulatory guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 1, U.
| |
| S. Nuclear Regulatory Commission, Washington, DC, November 2002. Available at ADAMS Accession No. ML023240437.
| |
| : 16. Regulatory Guide (RG) 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, U. S.
| |
| Nuclear Regulatory Commission, Washington, DC, March 2009. Available at ADAMS Accession No. ML090410014.
| |
| : 17. Regulatory Guide (RG) 1.189, "Fire Protection for Operating Nuclear Power Plants,"
| |
| Revision 2, U. S. Nuclear Regulatory Commission, Washington, DC, November 2009.
| |
| Available at ADAMS Accession No. ML092580550.
| |
| : 18. NUREG 0800, Standard Review Plan, Chapter 9.5.1.2, "Risk-Informed, Performance-Based Fire Protection," Revision 0, U. S. Nuclear Regulatory Commission, Washington, DC, December 2009. Available at ADAMS Accession No. ML092590527.
| |
| : 19. NUREG 0800, Standard Review Plan, Chapter 19.1, "Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, U. S.
| |
| Nuclear Regulatory Commission, Washington, DC, June 2007. Available at ADAMS Accession No. ML071700657.
| |
| : 20. NUREG 0800, Standard Review Plan, Chapter 19.2, "Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance,"
| |
| Revision 0, U. S. Nuclear Regulatory Commission, Washington, DC, June 2007. Available at ADAMS Accession No. ML071700658.
| |
| : 21. NEI 04-02, Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c), Revision 2, Nuclear Energy Institute (NEI),
| |
| Washington, DC, April 2008. ADAMS Accession No. ML081130188.
| |
| : 22. Regulatory Information Summary (RIS) 2007-19, "Process for Communicating Clarifications of Staff Positions Provided in Regulatory Guide 1.205 Concerning Issues Identified during the Pilot Application of National Fire Protection Association Standard 805," Revision 0, U. S.
| |
| Nuclear Regulatory Commission, Washington, DC, August 20,2007. Available at ADAMS Accession No. ML071590227.
| |
| : 23. Letter from Leonard N. Olshan, U.S. Nuclear Regulatory Commission, Bruce H. Hamilton, Duke Power Company LLC. August 16, 2007. Subject "Oconee Nuclear Stations, Units 1, 2, and 3 - Conforming License Amendments to Incorporate the Mitigation Strategies Required by Section B.5.b of Commission Order EA-02-026 (TAC NOS. MD4712, MD4713, and MD4714)." Washington, D.C. Available at ADAMS Accession No. ML072260290.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 116
| |
| : 24. Letter from J.F. Stolz, NRC to H.B. Tucker, Duke Power Company, April 28, 1983,
| |
| | |
| ==Subject:==
| |
| | |
| "Safety Evaluation by the Office of NRR of Standby Shutdown Facility." ADAMS Accession No. ML103370444.
| |
| : 25. Letter from Dave Baxter, Duke Energy, to the U. S. Nuclear Regulatory Commission Document Control Desk. May 18, 2010.
| |
| | |
| ==Subject:==
| |
| "Duke Energy Carolinas, LLC, Oconee Nuclear Station, Units 1, 2, and 3, Docket Numbers 50-269, 50-270 and 50-287, Renewed Operating Licenses DPR-38, DPR-47, and DPR-55, Revision to Tornado/HELB Mitigation Strategies and Regulatory Commitments." Seneca, SC. ADAMS Accession No.
| |
| ML101400144.
| |
| : 26. Letter from Robert W. Reid, U.S. Nuclear Regulatory Commission, to William O. Parker, Jr.,
| |
| Duke Power Company. August 11, 1978. "Fire Protection Safety Evaluation Report by the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, in the Matter of Duke Power Company Oconee Nuclear Station, Units 1, 2, & 3, Docket Nos. 50-269, -270, 287." Washington, D.C.
| |
| : 27. Letter from David B. Matthews, NRC, to H. B. Tucker, Duke Power Company. August 21, 1989.
| |
| | |
| ==Subject:==
| |
| Exemption from the Fire Protection Requirements of Section III.G of 10 CFR 50, Appendix R (TACs 52674/52675/52676). ADAMS Accession No. ML012000058.
| |
| : 28. NEI 00-01, Guidance for Post-Fire SSD Circuit Analysis, Revision 2, Nuclear Energy Institute (NEI), Washington, DC, December 2007. ADAMS Accession No. ML091770265.
| |
| : 29. NUREG-1824, "Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications," U.S. Nuclear Regulatory Commission, Washington, DC, May 2007.
| |
| : 30. ASTM E1355-05a, "Standard Guide for Evaluating the Predictive Capability of Deterministic Fire Models," American Society for Testing and Materials, West Conshohocken, PA, 2005.
| |
| : 31. Letter from Dave Baxter, Duke Energy to US NRC, June 29, 2009.
| |
| | |
| ==Subject:==
| |
| "Proposed License Amendment Request to Revise the Oconee Nuclear Station Current Licensing Basis for High Energy Line Break Events Outside of the Containment Building," License Amendment Request No. 2008-007. ADAMS Accession No. ML091870501.
| |
| : 32. NUMARC 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Revision 2, Nuclear Energy Institute (NEI), Washington, DC, April 1996.
| |
| : 33. NRC, "Oconee Nuclear Station Unit 3 Fire Probabilistic Risk Assessment Pre-Submittal Audit," June 24, 2008, ADAMS Accession No. ML080940603.
| |
| : 34. ASME/ANS RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008, Standard for Level1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," American Society of Mechanical Engineers and the American Nuclear Society, New York, NY, Draft.
| |
| : 35. NEI 07-12, "Fire Probabilistic Risk Assessment Peer Review Process Guidelines," (DRAFT),
| |
| December 1, 2008.
| |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
| |
| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 117
| |
| : 36. NRC, "Closure of NFPA-805 FAQ 08-0048 on Revised Fire Ignition Frequencies,"
| |
| September 1, 2009, ADAMS Accession No. ML092190457.
| |
| : 37. NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities,"
| |
| September 2005.
| |
| : 38. Letter from Harold R. Denton, NRC, to William O. Parker, Jr., Duke Power Company.
| |
| February 2, 1982. Oconee Units 1, 2, and 3, Exemption from 10 CFR 50.48 and Appendix R re Fixed Fire Suppression System. ADAMS Accession No. ML011990218.
| |
| : 39. Letter from Helen Nicolaras, U.S. Nuclear Regulatory Commission, to H. B. Tucker, Duke Power Company. December 27, 1984. Letter forwarding exemption pertaining to the requirement for emergency lighting, with at least an eight-hour battery power supply for the standby shutdown facility and the yard access route for Oconee, Units 1, 2, and 3. ADAMS Accession No. ML011990375.
| |
| : 40. Letter from John F. Stolz, U.S. Nuclear Regulatory Commission, to H. B. Tucker, Duke Power Company. August 31, 1983. Oconee Units 1,2, and 3 - Review of 07/15/1983 Request for Exemption. ADAMS Accession No. ML091310038.
| |
| : 41. Letter from John Stang, U.S. Nuclear Regulatory Commission, to Dave Baxter, Duke Energy Carolinas, LLC. November 18, 2009. "Oconee Nuclear Station, Units 1, 2, and 3 - Request for Additional Information Regarding License Amendment Request Transition to Title 10 of the Code of Federal Regulations, Section 50.48(c), National Fire Protection Association Standard NFPA 805 (TAC Nos. MD8822, MD8823, and MD8824)." Washington, D.C.
| |
| ADAMS Accession No. ML092920347.
| |
| : 42. Letter from John Stang, U.S. Nuclear Regulatory Commission, to Dave Baxter, Duke Energy Carolinas, LLC. March 8, 2010. "Oconee Nuclear Station, Units 1, 2, and 3 - Request for Additional Information Regarding License Amendment Request Transition to Title 10 of the Code of Federal Regulations, Section 50.48(c), National Fire Protection Association Standard NFPA 805 (TAC Nos. MD8822, IVID8823, and MD8824)." Washington, D.C.
| |
| ADAMS Accession No. ML100640646.
| |
| : 43. Letter from John Stang, U.S. Nuclear Regulatory Commission, to Dave Baxter, Duke Energy Carolinas, LLC. July 30, 2010. "Oconee Nuclear Station, Units 1, 2, and 3 (ONS)
| |
| Request for Additional Information (RAI) Regarding License Amendment Request, Transition to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.48(c), National Fire Protection Association Standard NFPA 805 (TAC Nos. ME3844, ME3845, and ME3846)." Washington, D.C. ADAMS Accession No. ML102110394.
| |
| : 44. NEI 04-06, "Guidance for Self-Assessment of Circuit Failure Issues," Revision L (DRAFT),
| |
| Nuclear Energy Institute, Washington, DC, March 2005. ADAMS Accession No. ML050760219.
| |
| : 45. Regulatory Issue Summary 2004-03, "Risk-Informed Approach for Post-Fire Safe-Shutdown Circuit Inspections," Revision 1, U. S. Nuclear Regulatory Commission, Washington, DC, dated December 29, 2004. ADAMS Accession No. ML042440791.
| |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 118
| |
| : 46. Inspection Manual Chapter 0609, "Significance Determination Process," Appendix G, "Shutdown Operations Significance Determination Process," Attachment 2, "Phase 2 Significance Determination Process Template for PWR During Shutdown," U. S. Nuclear Regulatory Commission, Washington, DC, February 2005. ADAMS Accession No. ML051400248.
| |
| : 47. ASIVIE RA-Sb-2005, "Addenda to ASIVIE RA-S-2002 Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," American Society of Mechanical Engineers, New York, NY, December 2005
| |
| : 48. ASME/ANS RA-S-2008, "Standard for Level1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," American Society of Mechanical Engineers and the American Nuclear Society, New York, NY, Draft.
| |
| : 49. Letter from Luis A. Reyes, U.S. Nuclear Regulatory Commission, to David A. Baxter, Duke Energy Carolinas, LLC. August 12, 2010.
| |
| | |
| ==Subject:==
| |
| "Final Significance Determination of One Yellow Finding and One White Finding and Notice of Violation (NRC Inspection Report 05000269/2010008, 05000269/2010007, 05000270/2010007, and 05000287/2010007, dated June 9,2010." Washington, D.C. ADAMS Accession No. ML102240588.
| |
| : 50. NEI 02-03, "Guidance for Performing a Regulatory Review of Proposed Changes to the Approved FPP," Revision 0, Nuclear Energy Institute, Washington, DC, June 2003. ADAMS Accession No. ML031780500.
| |
| : 51. Branch Technical Position (BTP) 9.5-1 APCSB Appendix A, "Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976," U. S. Nuclear Regulatory Commission, Washington, DC, August 1976. ADAMS Accession No. ML070660458.
| |
| : 52. Letter from T. Preston Gillespie, Jr., Duke Energy, to the U. S. Nuclear Regulatory Commission Document Control Desk. November 19,2010.
| |
| | |
| ==Subject:==
| |
| "Duke Energy Carolinas, LLC, Oconee Nuclear Station Units 1, 2, and 3, Docket Numbers 50-269, 50-270 and 50-287, Request for Additional Information regarding the License Amendment Request to adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition), License Amendment Request (LAR) No. 2008-01.
| |
| Seneca, SC. ADAMS Accession No. ML103300227.
| |
| : 53. EPRI Technical Report 1006756, "Fire Protection Surveillance Optimization and Maintenance Guide for Fire Protection Systems and Features," Electric Power Research Institute, Charlotte, NC, July 2003.
| |
| : 54. Letter from T. Preston Gillespie, Jr., Duke Energy, to the U. S. Nuclear Regulatory Commission Document Control Desk. October 14,2010.
| |
| | |
| ==Subject:==
| |
| "Duke Energy Carolinas, LLC, Oconee Nuclear Station Units 1, 2, and 3, Docket Numbers 50-269, 50-270 and 50 287, Request for Additional Information regarding the License Amendment Request to adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition), License Amendment Request (LAR) No. 2008-01.
| |
| Seneca, SC. ADAMS Accession No. ML102910093.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 119
| |
| : 55. Letter from Leonard A. Wiens, U.S. NRC to J.W. Hampton, Duke Power Company, LLC, January 6, 1993,
| |
| | |
| ==Subject:==
| |
| Issuance of Amendment - Oconee Nuclear Station, Units 1, 2, and 3 (TAC Nos. M84909, M84910, and M84911), Docket Nos. 50-269, 50-270, and 50 287. ADAMs Accession No. ML012040034.
| |
| : 56. NEI 00-01, Guidance for Post-Fire SSO Circuit Analysis, Revision 1, Nuclear Energy Institute (NEI), Washington, DC, January 2005. ADAMS Accession No. ML050310295.
| |
| : 57. NRC IN 92-18, Potential for Loss of Remote Shutdown Capability During a Control Room Fire, February 28, 1992. ADAMS Accession No. ML031200481.
| |
| : 58. Letter from Helen N. Pastis, U.S. NRC to, H. B. Tucker, Duke Power Company, LLC, June 7, 1988,
| |
| | |
| ==Subject:==
| |
| Issuance of Amendment Nos. 166, 166, and 163 to Facility Operating Licenses DPR-38, DPR-47, and DPR Oconee Nuclear Station, Units 1,2, and 3 (TAC Nos. 66090/66091/66092. ADAMs Accession No. ML012000214.
| |
| : 59. Letter from T. Preston Gillespie, Jr. Duke Energy, to the U. S. Nuclear Regulatory Commission Document Control Desk. December 22, 2010.
| |
| | |
| ==Subject:==
| |
| "Duke Energy Carolinas, LLC, Oconee Nuclear Station Units 1,2, and 3, Docket Nos. 50-269, 50-270 and 50-287, Request for Additional Information regarding the License Amendment Request to adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition), License Amendment Request (LAR) No. 2008-01."
| |
| Seneca, SC. ADAMS Accession NO.ML103620105.
| |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
| |
| | |
| OFFICIAL USE O~JLY SECURITY RELATED INFORMATION 120 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix This attachment contains Table 3.1-1, which provides the specific FPP elements and minimum design requirements from NFPA 805, Chapter 3, as appropriately modified by 10 CFR 50.48(c).
| |
| In addition, the table describes each fundamental FPP element from NFPA 805, Chapter 3, and identifies which of the methods listed below the licensee used as the means for achieving compliance with the requirement. Table 3.1-1 also provides the NRC staff's evaluation of the licensee's compliance statement for each FPP element. LAR Attachment A, "NEI 04-02 Table B-1, Transition of Fundamental FPP and Design Elements (NFPA 805, Chapter 3)," provides further details regarding the licensee's compliance strategy for specific NFPA 805, Chapter 3, requirements, including references to where compliance is documented.
| |
| As part of the assessment of its compliance with the NFPA 805, Chapter 3, elements, the licensee reviewed each section and subsection against the existing ONS FPP and provided specific compliance statements for each NFPA 805, Chapter 3, attribute that contained applicable requirements. The methods used by the licensee for achieving compliance with the NFPA 805, Chapter 3, fundamental FPP elements and minimum design requirements are as follows:
| |
| : 1. The existing FPP element directly complies with the requirement; noted in LAR Attachment A, also called the B-1 Table, as "Comply." In assessing these statements, the NRC staff reviewed the provided information to ensure that it presented a reasonable basis for concluding that the existing FPP element was adequate to meet the NFPA 805, Chapter 3, element.
| |
| : 2. The existing FPP element complies through the use of an explanation or clarification; noted in the B-1 Table as "Complies with Clarification." In assessing these statements, the NRC staff reviewed the provided information to ensure that it presented a reasonable basis for concluding that the FPP element, as clarified by the supplemental information, was adequate to meet the NFPA 805, Chapter 3, element.
| |
| : 3. The existing FPP element complies with the requirement based on prior NRC approval of an alternative to the fundamental FPP attribute and the bases for the NRC approval remain valid; noted in the B-1 Table as "Complies by Previous NRC Approval." In assessing these statements, the NRC staff reviewed the information provided to ensure that the basis was still valid for concluding that the alternative was adequate to meet the NFPA 805, Chapter 3, element.
| |
| : 4. The existing FPP element complies through the use of an EEEE; noted in the B-1 Table as "Complies with the Use of EEEE." In assessing these statements, the NRC staff reviewed the provided information to ensure that it presented a reasonable basis for concluding that the existing FPP element was adequate to meet the NFPA 805, Chapter 3, element.
| |
| : 5. The existing FPP element does not comply with the requirement, but the licensee is requesting approval for a PB method in accordance with 10 CFR 50.48(c)(2)(vii), noted as "Submit for NRC Approval."
| |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 121 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staff's Evaluation Statement General. This chapter contains the fundamental elements of the FPP and specifies the minimum design requirements for fire protection systems and features.
| |
| These FPP elements and minimum design requirements 3.1 shall not be subject to the performance-based methods Individual Elements Reviewed Below permitted elsewhere in this standard. Previously approved alternatives from the fundamental FPP attributes of this chapter by the AHJ take precedence over the requirements contained herein.
| |
| 3.2 Fire Protection Plan. Individual Elements Reviewed Below 3.2.1 Intent. A site-wide fire protection plan shall be established. This plan shall document management policy and program direction and shall define the responsibilities of those individuals responsible for the The NRC staff finds the licensee's statement of 3.2.1 Comply plan's implementation. This section establishes the compliance ~cceptable.
| |
| criteria for an integrated combination of components, procedures, and personnel to implement all FPP activities.
| |
| Management Policy Direction and Responsibility. A policy document shall be prepared that defines The NRC staff finds the licensee's statement of 3.2.2 Comply management authority and responsibilities and compliance acceptable.
| |
| establishes the general policy for the site FPP.
| |
| The policy document shall designate the senior The NRC staff finds the licensee's statement of 3.2.2.1 management position with immediate authority and Comply compliance acceptable.
| |
| responsibility for the FPP.
| |
| The policy document shall designate a position The NRC staff finds the licensee's statement of 3.2.2.2 responsible for the daily administration and coordination Comply compliance acceptable.
| |
| of the FPP and its implementation.
| |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
| |
| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 122 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement The policy document shall define the fire protection interfaces with other organizations and assign responsibilities for the coordination of activities. In The NRC staff finds the licensee's statement of 3.2.2.3 Comply addition, this policy document shall identify the various compliance acceptable.
| |
| plant positions having the authority for implementing the various areas of the FPP.
| |
| The NRC staff finds the licensee's statement of compliance acceptable.
| |
| IMPLEMENTATION ITEM - The licensee The policy document shall identify the appropriate AHJ 3.2.2.4 Comply identified an action to update the design basis for the various areas of the FPP.
| |
| specification (DBS) to include the statement the NRC is the AHJ for fire protection changes requiring approval (SE Section 2.9, Table 2.9-1, Item 1).
| |
| Procedures. Procedures shall be established for implementation of the FPP. In addition to procedures 3.2.3 that could be required by other sections of the standard, Individual Elements Reviewed Below the procedures to accomplish the following shall be established:
| |
| Inspection, testing, and maintenance for fire protection The NRC staff finds the licensee's statement of 3.2.3.(1) Comply systems and features credited by the FPP compliance acceptable.
| |
| Compensatory actions implemented when fire protection systems and other systems credited by the FPP and this The NRC staff finds the licensee's statement of 3.2.3.(2) Comply standard cannot perform their intended function and compliance acceptable.
| |
| limits on impairment duration OFFICIAL USE ONLY SECURITY RELATED INFORMATION
| |
| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 123 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement The NRC staff finds the licensee's statement of compliance acceptable.
| |
| IMPLEMENTATION ITEM - The licensee 3.2.3.(3) Reviews of FPP related perforsmance and trends Comply identified an action to complete the development of the NFPA 805 Monitoring Program including surveillance frequencies (SE Section 2.9, Table 2.9-1, Item 8).
| |
| Reviews of physical plant modifications and procedure The NRC staff finds the licensee's statement of 3.2.3.(4) Comply changes for impact on the FPP. compliance acceptable.
| |
| The NRC staff finds the licensee's statement of 3.2.3.(5) Long-term maintenance and configuration of the FPP. Comply compliance acceptable.
| |
| Emergency response procedures for the plant industrial The NRC staff finds the licensee's statement of 3.2.3.(6) Comply fire briQade. compliance acceptable.
| |
| Prevention. A fire prevention program with the goal of preventing a fire from starting shall be established, 3.3 documented, and implemented as part of the FPP. The Individual Elements Reviewed Below two basic components of the fire prevention program shall consist of both of the followin~:r Prevention of fires and fire spread by controls on The NRC staff finds the licensee's statement of 3.3.(1 ) Comply operational activities. compliance acceptable.
| |
| Design controls that restrict the use of combustible The NRC staff finds the licensee's statement of 3.3.(2) Comply materials. compliance acceptable.
| |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
| |
| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 124 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement Fire Prevention for Operational Activities. The fire prevention program activities shall consist of the necessary elements to address the control of ignition sources and the use of transient combustible materials The NRC staff finds the licensee's statement of 3.3.1 during all aspects of plant operations. The fire Comply compliance acceptable.
| |
| prevention program shall focus on the human and programmatic elements necessary to prevent fires from starting or, should a fire start, to keep the fire as small as possible.
| |
| General Fire Prevention Activities. The fire prevention The NRC staff finds the licensee's statement of 3.3.1.1 activities shall include but not be limited to the following Comply compliance acceptable.
| |
| proQram elements:
| |
| Training on fire safety information for all employees and contractors including, as a minimum, familiarization with The NRC staff finds the licensee's statement of 3.3.1.1.(1) Comply plant fire prevention procedures, fire reporting, and plant compliance acceptable.
| |
| emergency alarms.
| |
| Documented plant inspections including provisions for The NRC staff finds the licensee's statement of 3.3.1.1.(2) corrective actions for conditions where unanalyzed fire Comply compliance acceptable.
| |
| hazards are identified.
| |
| Administrative controls addressing the review of plant modifications and maintenance to ensure that both fire The NRC staff finds the licensee's statement of 3.3.1.1.(3) Comply hazards and the impact on plant fire protection systems compliance acceptable.
| |
| and features are minimized.
| |
| Control of Combustible Materials. Procedures for the control of general housekeeping practices and the The NRC staff finds the licensee's statement of 3.3.1.2 control of transient combustibles shall be developed and Comply compliance acceptable.
| |
| implemented. These procedures shall include but not be limited to the following program elements:
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 125 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement Wood used within the power block shall be listed pressure- impregnated or coated with a listed fire-Submit For The NRC staff finds the licensee's proposed PB retardant application.
| |
| 3.3.1.2.(1 ) NRC method to demonstrate compliance is acceptable Exception: Cribbing timbers 6 in. by 6 in. (15.2 em by Approval as described in SE Section 3.1.3.1.
| |
| 15.2 em) or larger shall not be required to be fire-retardant treated.
| |
| The NRC staff finds the licensee's statement of compliance acceptable.
| |
| Plastic sheeting materials used in the power block shall be fire-retardant types that have passed NFPA 701, 3.3.1.2.(2) Comply IMPLEMENTATION ITEM - The licensee Standard Methods of Fire Tests for Flame Propagation of identified an action to complete the development Textiles and Films, large-scale tests, or equivalent.
| |
| of procedural controls for plastic sheeting (SE Section 2.9, Table 2.9-1, Item 2).
| |
| Waste, debris, scrap, packing materials, or other combustibles shall be removed from an area immediately The NRC staff finds the licensee's statement of 3.3.1.2.(3) Comply following the completion of work or at the end of the shift, compliance acceptable.
| |
| whichever comes first.
| |
| Combustible storage or staging areas shall be The NRC staff finds the licensee's statement of 3.3.1.2.(4) designated, and limits shall be established on the types Comply compliance acceptable.
| |
| and quantities of stored materials.
| |
| The NRC staff finds that the licensee's Controls on use and storage of flammable and explanation of their method of compliance with combustible liquids shall be in accordance with NFPA 30, Complies with 3.3.1.2.(5) these requirements acceptable based on the Flammable and Combustible Liquids Code, or other Clarification information provided in the LAR B-1 Table applicable NFPA Standards.
| |
| element.
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| OFFICIAL USE O~JLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 126 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staff's Evaluation Statement The NRC staff finds that the licensee's explanation of their method of compliance with Controls on use and storage of flammable gases shall be Complies with 3.3.1.2.(6) these requirements acceptable based on the in accordance with applicable NFPA standards. Clarification information provided in the LAR B-1 Table element.
| |
| 3.3.1.3 Control of Ignition Sources. Individual Elements Reviewed Below A hot work safety procedure shall be developed, implemented, and periodically updated as necessary in accordance with NFPA 51 B, Standard for Fire Prevention The NRC staff finds the licensee's statement of 3.3.1.3.1 Comply During Welding, Cutting, and Other Hot Work, and NFPA compliance acceptable.
| |
| 241, Standard for Safeguarding Construction, Alteration, and Demolition Operations.
| |
| Smoking and other possible sources of ignition shall be The NRC staff finds the licensee's statement of 3.3.1.3.2 restricted to properly designated and supervised safe Comply compliance acceptable.
| |
| areas of the plant.
| |
| The NRC staff finds the licensee's statement of compliance acceptable.
| |
| IMPLEMENTATION ITEM - The licensee Open flames or combustion-generated smoke shall not 3.3.1.3.3 Comply identified an action to update appropriate station be permitted for leak or airflow testing.
| |
| procedure(s) for leak or air flow testing to preclude the use of open flames or combustion generated smoke (SE Section 2.9, Table 2.9-1, Item 3).
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 127 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement The NRC staff finds the licensee's statement of compliance acceptable.
| |
| Plant administrative procedure shall control the use of IMPLEMENTATION ITEM - The licensee portable electrical heaters in the plant. Portable fu.el.-fired identified an action to update the directives to 3.3.1.3.4 heaters shall not be permitted in plant areas containing Comply prohibit the use of fuel-fired heaters in plant a~eas equipment important to nuclear safety or where there is a with equipment important to nuclear safety or In potential for radiological releases resulting from a fire.
| |
| areas where there is the potential for radiological release due to fire (SE Section 2.9, Table 2.9-1, Item 35).
| |
| Structural. Walls, floors, and components required to maintain structural integrity shall be of noncombustible The NRC staff finds the licensee's statement of 3.3.2 Comply construction, as defined in NFPA 220, Standard on compliance acceptable.
| |
| Types of Buildinq Construction.
| |
| The NRC staff finds the licensee's statement of compliance acceptable.
| |
| Interior Finishes. Interior wall or ceiling finish classification shall be in accordance with NFPA 101, IMPLEMENTATION ITEM - The licensee 3.3.3 Life Safety Code, requirements for Class A materials. Comply identified an action to update the coatings Interior floor finishes shall be in accordance with NFPA program directives to include Class A and Class I 101 requirements for Class I interior floor finishes.
| |
| specifications for interior finishes (SE Section 2.9, Table 2.9-1, Item 4).
| |
| Insulation Materials. Thermal insulation materials, radiation shielding materials, ventilation duct materials, The NRC staff finds the licensee's statement of 3.3.4 Comply and soundproofing materials shall be noncombustible or compliance acceptable.
| |
| limited combustible.
| |
| 3.3.5 Electrical. Individual Elements Reviewed Below OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 128 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement Wiring above suspended ceiling shall be kept to a minimum. Where installed, electrical wiring shall be Submit For The NRC staff finds that the licensee's proposed 3.3.5.1 listed for plenum use, routed in armored cable, routed in NRC PB method to demonstrate compliance is metallic conduit, or routed in cable trays with solid metal Approval acceptable as described in SE Section 3.1.3.3.
| |
| top and bottom covers.
| |
| The NRC staff finds the licensee's statement of Only metal tray and metal conduits shall be used for compliance acceptable.
| |
| electrical raceways. Thin wall metallic tubing shall not be 3.3.5.2 used for power, instrumentation, or control cables. Comply IMPLEMENTATION ITEM - The licensee Flexible metallic conduits shall only be used in short identified an action to update documentation for lengths to connect components. the use of electrical raceway construction limits (SE Section 2.9, Table 2.9-1, Item 5).
| |
| Electric cable construction shall comply with a flame Submit For The NRC staff finds that the licensee's proposed propagation test as acceptable to the AHJ.
| |
| 3.3.5.3 NRC PB method to demonstrate compliance is Approval acceptable as described in SE Section 3.1.3.4.
| |
| rNote: This entry modified per 10 CFR 50A8( c)(2)(v)1 Roofs. Metal roof deck construction shall be designed and installed so the roofing system will not sustain a self-propagating fire on the underside of the deck when the The NRC staff finds the licensee's statement of 3.3.6 deck is heated by a fire inside the building. Roof Comply compliance acceptable.
| |
| coverings shall be Class A as determined by tests described in NFPA 256, Standard Methods of Fire Tests of Roof Coverinas.
| |
| Bulk Flammable Gas Storage. Bulk compressed or cryogenic flammable gas storage shall not be permitted The NRC staff finds the licensee's statement of 3.3.7 Comply inside structures housing systems, equipment, or compliance acceptable components important to nuclear safety.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 129 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staff's Evaluation Statement The licensee stated that compliance has been demonstrated through the use of an EEEE.
| |
| Complies with Based on the licensee's justification of continued Use of EEEE validity and evaluation quality, the NRC staff finds Storage of flammable gas shall be located outdoors, or in the licensee's statement of compliance separate detached buildings, so that a fire or explosion acceptable.
| |
| will not adversely impact systems, equipment, or 3.3.7.1 components important to nuclear safety. NFPA SOA, The NRC staff finds the licensee's statement of Standard for Gaseous Hydrogen Systems at Consumer Comply compliance acceptable.
| |
| Sites, shall be followed for hydrogen storage.
| |
| Submit For The NRC staff finds that the licensee's proposed NRC PB method to demonstrate compliance is Approval acceptable as described in SE Section 3.1.3.2.
| |
| Outdoor high-pressure flammable gas storage containers The NRC staff finds the licensee's statement of 3.3.7.2 shall be located so that the long axis is not pointed at Comply compliance acceptable.
| |
| buildinqs.
| |
| Flammable gas storage cylinders not required for normal The NRC staff finds the licensee's statement of 3.3.7.3 Comply operation shall be isolated from the system. compliance acceptable.
| |
| Bulk Storage of Flammable and Combustible Liquids. Bulk storage of flammable and combustible liquids shall not be permitted inside structures containing The NRC staff finds the licensee's statement of 3.3.8 systems, equipment, or components important to nuclear Comply compliance acceptable.
| |
| safety. As a minimum, storage and use shall comply with NFPA 30, Flammable and Combustible Liquids Code.
| |
| OFFICIAL USE O~JLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 130 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement The NRC staff finds the licensee's statement of compliance acceptable.
| |
| Transformers. Where provided, transformer oil collection basins and drain paths shall be periodically 3.3.9 Comply IMPLEMENTATION ITEM - The licensee inspected to ensure that they are free of debris and identified an action to update transformer deluge capable of performing their design function.
| |
| system flow test procedures (SE Section 2.9, Table 2.9-1, Item 6).
| |
| Hot Pipes and Surfaces. Combustible liquids, includ.ing high flashpoint lubricating oils, shall be .kept f~om coming The NRC staff finds the licensee's statement of 3.3.10 in contact with hot pipes and surfaces, Including Comply compliance acceptable.
| |
| insulated pipes and surfaces. Administrative controls shall require the prompt cleanup of oil on insulation.
| |
| Electrical Equipment. Adequate clearance, free of The NRC staff finds the licensee's statement of 3.3.11 combustible material, shall be maintained around Comply compliance acceptable.
| |
| energized electrical equipment.
| |
| Reactor Coolant Pumps. For facilities with non-inerted containments RCPs with an external lubrication system shall be provided with an oil collection system. The oil The NRC staff finds the licensee's statement of 3.3.12 collection system shall be designed and install~d such Comply compliance acceptable.
| |
| that leakage from the oil system is safely contained for off normal conditions such as accident conditions or earthquakes. All of the following shall apply.
| |
| The NRC staff finds the licensee's statement of The oil collection system for each RCP shall be capable Comply compliance acceptable.
| |
| of collecting lubricating oil from all potential pressurized 3.3.12.(1) Submit For The NRC staff finds that the licensee's proposed and nonpressurized leakage sites in each RCP oil NRC PB method to demonstrate compliance is system.
| |
| Approval acceptable as described in SE Section 3.1.3.
| |
| Leakage shall be collected and drained to a vented The NRC staff finds the licensee's statement of 3.3.12.(2) closed container that can hold the inventory of the RCP Comply compliance acceptable.
| |
| lubricating oil system.
| |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 131 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement A flame arrestor is required in the vent if the flash point The NRC staff finds the licensee's statement of 3.3.12.(3) characteristics of the oil present the hazard of a fire Comply compliance acceptable.
| |
| flashback.
| |
| Leakage points on an RCP motor to be protected shall include but not be limited to the lift pump and piping, The NRC staff finds the licensee's statement of 3.3.12.(4) overflow lines, oil cooler, oil fill and drain lines and plugs, Comply compliance acceptable.
| |
| flanged connections on oil lines, and the oil reservoirs, where such features exist on the RCPs.
| |
| The collection basin drain line to the collection tank shall The NRC staff finds the licensee's statement of 3.3.12.(5) be large enough to accommodate the largest potential oil Comply compliance acceptable.
| |
| leak such that oil leakage does not overflow the basin.
| |
| 3.4 Industrial Fire Brigade. Individual Elements Reviewed Below On-Site Fire-Fighting Capability. All of the following 3.4.1 Individual Elements Reviewed Below requirements shall apply.
| |
| The licensee stated that compliance has been A fully staffed, trained, and equipped fire-fighting force demonstrated through the use of an EEEE.
| |
| shall be available at all times to control and extinguish all Based on the licensee's justification of continued Complies with 3.4.1.(a) fires on site. This force shall have a minimum validity and evaluation quality, the NRC staff finds Use of EEEE complement of five persons on duty and shall conform the licensee's statement of compliance with the following NFPA standards as applicable: acceptable. Note that the licensee states that they will comply with NFPA 600, 2005 Edition.
| |
| The licensee stated that compliance has been demonstrated through the use of an EEEE.
| |
| Based on the licensee's justification of continued NFPA 600, Standard on Industrial Fire Brigades (interior Complies with 3.4.1.(a).(1 ) validity and evaluation quality, the NRC staff finds structural fire fighting) Use of EEEE the licensee's statement of compliance acceptable. Note that the licensee states that they will comply with NFPA 600, 2005 Edition.
| |
| NFPA 1500, Standard on Fire Department Occupational The licensee stated it complies with NFPA 600, 3.4.1.(a).(2)
| |
| Safety and Health Proqram 2005 Edition: see subsection 3.4.1.(a).(1) above.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 132 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement NFPA 1582, Standard on Medical Requirements for Fire The licensee stated it complies with NFPA 600, 3.4.1.(a).(3)
| |
| Fiqhters and Information for Fire Department Physicians 2005 Edition: see subsection 3.4.1.(a).(1) above.
| |
| Industrial fire brigade members shall have no other assigned normal plant duties that would prevent The NRC staff finds the licensee's statement of 3.4.1.(b) Comply immediate response to a fire or other emergency as compliance acceptable.
| |
| required.
| |
| During every shift, the brigade leader and at least two brigade members shall have sufficient training and knowledge of nuclear safety systems to understand the effects of fire and fire suppressants on nuclear safety The NRC staff finds the licensee's statement of 3.4.1.(c) Comply performance criteria. compliance acceptable.
| |
| Exception to (c): Sufficient training and knowledge shall be permitted to be provided by an operations advisor dedicated to industrial fire briqade support.
| |
| The industrial fire brigade shall be notified immediately The NRC staff finds the licensee's statement of 3.4.1.(d) Comply upon verification of a fire. compliance acceptable.
| |
| Each industrial fire brigade member shall pass an annual physical examination to determine that he or she can perform the strenuous activity required during manual The NRC staff finds the licensee's statement of 3.4.1.(e) Comply firefighting operations. The physical examination shall compliance acceptable.
| |
| determine the ability of each member to use respiratory protection equipment.
| |
| Pre-Fire Plans. Current and detailed pre-fire plans shall be available to the industrial fire brigade for all areas in The NRC staff finds the licensee's statement of 3.4.2 Comply which a fire could jeopardize the ability to meet the compliance acceptable.
| |
| performance criteria described in Section 1.5.
| |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 133 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staff's Evaluation Statement The NRC staff finds the licensee's statement of The plans shall detail the fire area configuration and fire compliance acceptable.
| |
| hazards to be encountered in the fire area, along with 3.4.2.1 Comply any nuclear safety components and fire protection IMPLEMENTATION ITEM - The licensee systems and features that are present. identified an action to update the pre-fire plans (SE Section 2.9, Table 2.9-1, Item 9).
| |
| Pre-fire plans shall be reviewed and updated as The NRC staff finds the licensee's statement of 3.4.2.2 Comply necessary. compliance acceptable.
| |
| The NRC staff finds the licensee's statement of compliance acceptable.
| |
| Pre-fire plans shall be available in the control room and 3.4.2.3 Comply made available to the plant industrial fire brigade. IMPLEMENTATION ITEM - The licensee identified an action to update the pre-fire plans (SE Section 2.9, Table 2.9-1, Item 10).
| |
| Pre-fire plans shall address coordination with other plant The NRC staff finds the licensee's statement of 3.4.2.4 Comply qroups durinq fire emerqencies. compliance acceptable.
| |
| Training and Drills. Industrial fire brigade members and other plant personnel who would respond to a fire in 3.4.3 conjunction with the brigade shall be provided with Individual elements reviewed below training commensurate with their emergency responsibilities.
| |
| Plant Industrial Fire Brigade Training. All of the following 3.4.3.(a) Individual elements reviewed below requirements shall apply.
| |
| The licensee stated that compliance has been Plant industrial fire brigade members shall receive demonstrated through the use of an EEEE.
| |
| training consistent with the requirements contained in Based on the licensee's justification of continued Complies with 3.4.3.(a).(1 ) NFPA 600, Standard on Industrial Fire Brigades, or validity and evaluation quality, the NRC staff finds Use of EEEE NFPA 1500, Standard on Fire Department Occupational the licensee's statement of compliance Safety and Health Program, as appropriate. acceptable. Note that the licensee states that they comply with NFPA 600, 2005 Edition.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 134 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement Industrial fire brigade members shall be given quarterly training and practice in fire fighting, including radioactivity The NRC staff finds the licensee's statement of 3.4.3.(a).(2) and health physics considerations, to ensure that each Comply compliance acceptable.
| |
| member is thoroughly familiar with the steps to be taken in the event of a fire.
| |
| A written program shall detail the industrial fire brigade The NRC staff finds the licensee's statement of 3.4.3.(a).(3) Comply training program. compliance acceptable.
| |
| Written records that include but are not limited to initial industrial fire brigade classroom and hands-on training, refresher training, special training schools attended, drill The NRC staff finds the licensee's statement of 3.4.3.(a).(4) Comply attendance records, and leadership training for industrial compliance acceptable.
| |
| fire brigades shall be maintained for each industrial fire brigade member.
| |
| Training for Non-Industrial Fire Brigade Personnel. Plant personnel who respond with the industrial fire brigade The NRC staff finds the licensee's statement of 3.4.3.(b) shall be trained as to their responsibilities, potential Comply compliance acceptable.
| |
| hazards to be encountered, and interfacing with the industrial fire brigade.
| |
| 3.4.3.(c) Drills. All of the followinQ requirements shall apply. Individual elements reviewed below Drills shall be conducted quarterly for each shift to test The NRC staff finds the licensee's statement of 3.4.3.(c).(1 ) Comply the response capability of the industrial fire brigade. compliance acceptable.
| |
| Industrial fire brigade drills shall be developed to test and challenge industrial fire brigade responses, including brigade performance as a team, proper use of equipment, effective use of pre-fire plans, and The NRC staff finds the licensee's statement of 3.4.3.(c).(2) coordination with other groups. These drills shall Comply compliance acceptable.
| |
| evaluate the industrial fire brigade's abilities to react, respond, and demonstrate proper fire-fighting techniques to control and extinguish the fire and smoke conditions being simulated by the drill scenario.
| |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 135 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement The NRC staff finds the licensee's statement of compliance acceptable.
| |
| Industrial fire brigade drills shall be conducted in various plant areas, especially in those areas id~nti~ied. ~o be .
| |
| 3.4.3.(c).(3) Comply IMPLEMENTATION ITEM - The licensee essential to plant operation and to contain significant fire identified an action to update fire brigade training hazards.
| |
| documentation (SE Section 2.9, Table 2.9-1, Item 7).
| |
| Drill records shall be maintained detailing the drill The NRC staff finds the licensee's statement of 3.4.3.(c).(4) scenario, industrial fire brigade member response, and Comply compliance acceptable.
| |
| ability of the industrial fire brigade to perform as a team.
| |
| The NRC staff finds the licensee's statement of 3.4.3.(c).(5) A critique shall be held and documented after each drill. Comply compliance acceptable.
| |
| Fire-Fighting Equipment. Protective clothing, respiratory protective equipment, radiation monitoring equipment, personal dosimeters, and fire suppression The NRC staff finds the licensee's statement of 3.4.4 equipment such as hoses, nozzles, fire extinguishers, Comply compliance acceptable.
| |
| and other needed equipment shall be provided for the industrial fire brigade. This equipment shall conform with the applicable NFPA standards.
| |
| 3.4.5 Off-Site Fire Department Interface. Individual elements reviewed below.
| |
| Mutual Aid Agreement. Off-site fire authorities shall be The NRC staff finds the licensee's statement of 3.4.5.1 offered a plan for their interface during fires and related Comply compliance acceptable.
| |
| emergencies on site.
| |
| Site-Specific Training. Fire fighters from the off-site fire authorities who are expected to respond to a fire at the The NRC staff finds the licensee's statement of 3.4.5.2 Comply plant shall be offered site-specific training and shall be compliance acceptable.
| |
| invited to participate in a drill at least annually.
| |
| Security and Radiation Protection. Plant security and The NRC staff finds the licensee's statement of 3.4.5.3 radiation protection plans shall address off-site fire Comply compliance acceptable.
| |
| authority response.
| |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 136 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staff's Evaluation Statement Communications. An effective emergency The NRC staff finds the licensee's statement of 3.4.6 communications capability shall be provided for the Comply compliance acceptable.
| |
| industrial fire brigade.
| |
| 3.5 Water Supply. Individual elements reviewed below A fire protection water supply of adequate reliability, quantity, and duration shall be provided by one of the two following methods.
| |
| (a) Provide a fire protection water supply of not less than two separate 300,OOO-gal (1,135,500-L) supplies.
| |
| (b) Calculate the fire flow rate for 2 hours. This fire flow The NRC staff finds that the licensee's rate shall be based on 500 gpm (1892.5 Llmin) for explanation of their PB method of compliance 3.5.1 manual hose streams plus the largest design demand of Comply using the HPSW system with these requirements any sprinkler or fixed water spray system(s) in the power acceptable to meet the intent of this subsection.
| |
| block as determined in accordance with NFPA 13, Standard for the Installation of Sprinkler Systems, or NFPA 15, Standard for Water Spray Fixed Systems for Fire Protection. The fire water supply shall be capable of delivering this design demand with the hydraulically least demanding portion of fire main loop out of service.
| |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
| |
| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 137 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement The tanks shall be interconnected such that fire pumps can take suction from either or both. A failure in one tank or its piping shall not allow both tanks to drain. The tanks shall be designed in accordance with NFPA 22, Standard for Water Tanks for Private Fire Protection.
| |
| Exception NO.1: Water storage tanks shall not be required when fire pumps are able to take suction from a The NRC staff finds the licensee's statement of 3.5.2 Comply large body of water (such as a lake), provided each fire compliance acceptable.
| |
| pump has its own suction and both suctions and pumps are adequately separated.
| |
| Exception NO.2: Cooling tower basins shall be an acceptable water source for fire pumps when the volume is sufficient for both purposes and water quality is consistent with the demands of the fire service.
| |
| The NRC staff has previously approved an alternative to this requirement that the licensee is carrying forward into the RI/PB FPP. The NRC Complies by staff has accepted the use of the HPSW pumps Previous as fire pumps in NRC SE, Section 4.3.1.2, dated Fire pumps, designed and installed in accordance with NRC August 11, 1978. Based on the licensee's NFPA 20, Standard for the Installation of Stationary Approval justification of continued validity, the NRC staff Pumps for Fire Protection, shall be provided to ensure 3.5.3 finds the licensee's statement of compliance that 100 percent of the req uired flow rate and pressure acceptable.
| |
| are available assuming failure of the largest pump or The licensee stated that compliance has been pump power source.
| |
| demonstrated through the use of an EEEE.
| |
| Complies with Based on the licensee's justification of continued Use of EEEE validity and evaluation quality, the NRC staff finds the licensee's statement of compliance acceptable.
| |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
| |
| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 138 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement The NRC staff finds that the licensee's proposed Submit For PB method to demonstrate compliance is NRC acceptable as described in SE Section 3.1.3.6, Approval 3.1.3.7, 3.1.3.8, and 3.1.3.11.
| |
| For use of the HPSW pumps as fire pumps to protect ONS Power Block: the NRC staff has At least one diesel engine-driven fire pump or two more previously approved an alternative to this Complies by seismic Category I Class IE electric motor-driven fire requirement that the licensee is carrying forward Previous 3.5.4 pumps connected to redundant Class IE emergency into the RIIPB FPP in NRC SE Section 4.3.1.2, NRC power buses capable of providing 100 percent of the dated August 11, 1978 (Reference 26). Based on Approval required flow rate and pressure shall be provided. the licensee's justification of continued validity, the NRC staff finds the licensee's statement of compliance acceptable.
| |
| For use of a single fire pump at Keowee Hydro Submit for Station, the NRC staff finds that the licensee's NRC proposed PB method to demonstrate compliance Approval is acceptable as described in SE Section 3.1.3.11.
| |
| The licensee stated that compliance has been demonstrated through the use of an EEEE.
| |
| Each pump and its driver and controls shall be separated Complies with Based on the licensee's justification of continued from the remaining fire pumps and from the rest of the Use of EEEE validity and evaluation quality, the NRC staff finds 3.5.5 plant by rated fire barriers.
| |
| the licensee's statement of compliance acceptable.
| |
| The NRC staff finds the licensee's statement of Comply compliance acceptable.
| |
| Submit For The NRC staff finds that the licensee's proposed Fire pumps shall be provided with automatic start and 3.5.6 NRC PB method to demonstrate compliance is manual stop only.
| |
| Approval acceptable as described in SE Section 3.1.3.8.
| |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 139 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement The NRC staff finds the licensee's statement of Comply Individual fire pump connections to the yard fire main compliance acceptable.
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| 3.5.7 loop shall be provided and separated with sectionalizing Submit For The NRC staff finds that the licensee's proposed valves between connections. NRC PB method to demonstrate compliance is Approval acceptable as described in SE Section 3.1.3.9.
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| A method of automatic pressure maintenance of the fire The NRC staff finds the licensee's statement of 3.5.8 protection water system shall be provided independent of Comply compliance acceptable.
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| the fire pumps.
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| Means shall be provided to immediately notify the control The NRC staff finds the licensee's statement of 3.5.9 room, or other suitable constantly attended location, of Comply compliance acceptable.
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| operation of fire pumps.
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| The licensee stated that compliance has been demonstrated through the use of an EEEE.
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| An underground yard fire main loop, designed and Complies with Based on the licensee's justification of continued installed in accordance with NFPA 24, Standard for the Use of EEEE validity and evaluation quality, the NRC staff finds 3.5.10 Installation of Private Fire Service Mains and Their the licensee's statement of compliance Appurtenances, shall be installed to furnish anticipated acceptable.
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| water requirements. Submit For The NRC staff finds that the licensee's proposed NRC PB method to demonstrate compliance is Approval acceptable as described in SE Section 3.1.3.9.
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| Means shall be provided to isolate portions of the yard fire main loop for maintenance or repair without simultaneously shutting off the supply to both fixed fire suppression systems and fire hose stations provided for manual backup. Sprinkler systems and manual hose The NRC staff finds the licensee's statement of 3.5.11 Comply station standpipes shall be connected to the plant fire compliance acceptable.
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| protection water main so that a single active failure or a crack to the water supply piping to these systems can be isolated so as not to impair both the primary and backup fire suppression systems.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 140 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement Threads compatible with those used by local fire departments shall be provided on all hydrants, hose couplings, and standpipe risers.
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| Exception: Fire departments shall be permitted to be The NRC staff finds the licensee's statement of 3.5.12 Comply provided with adapters that allow interconnection compliance acceptable.
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| between plant equipment and the fire department equipment if adequate training and procedures are provided.
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| Headers fed from each end shall be permitted inside buildings to supply both sprinkler and standpipe systems, provided steel piping and fittings meeting the requirements of ANSI 831.1, Code for Power Piping, are used for the headers (up to and including the first valve) supplying the sprinkler systems where such headers are The NRC staff finds the licensee's statement of 3.5.13 Comply part of the seismically analyzed hose standpipe system. compliance acceptable.
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| Where provided, such headers shall be considered an extension of the yard main system. Each sprinkler and standpipe system shall be equipped with an outside screw and yoke (OS&Y) gate valve or other approved shutoff valve.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 141 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement All fire protection water supply and fire suppression system control valves shall be under a periodic inspection program and shall be supervised by one of the following methods:
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| (a) Electrical supervision with audible and visual signals in the MCR or other suitable constantly attended The NRC staff finds the licensee's statement of 3.5.14 location. Comply compliance acceptable.
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| (b) Locking valves in their normal position. Keys shall be made available only to authorized personnel.
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| (c) Sealing valves in their normal positions. This option shall be utilized only where valves are located within fenced areas or under the direct control of the owner/operator.
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| The NRC staff has previously approved an Hydrants shall be installed approximately every 250 ft alternative to this requirement that the licensee is (76 m) apart on the yard main system. A hose house carrying forward into the RI/PB FPP. The NRC equipped with hose and combination nozzle and other Complies by staff has accepted the use of the installed auxiliary equipment specified in NFPA 24, Standard for Previous NRC hydrants in NRC SE, Section 4.3.1.3, dated the Installation of Private Fire Service Mains and Their Approval August 11, 1978 (Reference 26). Based on the Appurtenances, shall be provided at intervals of not more licensee's justification of continued validity, the 3.5.15 than 1000 ft (305 m) along the yard main system. NRC staff finds the licensee's statement of compliance acceptable.
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| Exception: Mobile means of providing hose and The licensee stated that compliance has been associated equipment, such as hose carts or trucks, shall demonstrated through the use of an EEEE.
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| be permitted in lieu of hose houses. Where provided, Complies with Based on the licensee's justification of continued such mobile equipment shall be equivalent to the Use of EEEE validity and evaluation quality, the NRC staff finds equipment supplied by three hose houses. the licensee's statement of compliance acceptable.
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| OFFICIAL USE ONLY SECURITY REL/\TED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 142 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement Submit For The NRC staff finds that the licensee's proposed NRC PB method to demonstrate compliance is Approval acceptable as described in SE Section 3.1.3.9.
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| The fire protection water supply system shall be dedicated for fire protection use only.
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| Exception NO.1: Fire protection water supply systems The NRC staff has previously approved an shall be permitted to be used to provide backup to alternative to this requirement that the licensee is nuclear safety systems, provided the fire protection water carrying forward into the RI/PB FPP. The NRC Complies by supply systems are designed and maintained to deliver staff has accepted the use of the water supply in Previous 3.5.16 the combined fire and nuclear safety flow demands for NRC SE, Sections 4.3.1.2 and 4.3.1.4, dated NRC the duration specified by the applicable analysis. August 11, 1978 (Reference 26). Based on the Approval Exception NO.2: Fire protection water storage can be licensee's justification of continued validity, the provided by plant systems serving other functions, NRC staff finds the licensee's statement of provided the storage has a dedicated capacity capable of compliance acceptable.
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| providing the maximum fire protection demand for the specified duration as determined in this section.
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| The licensee stated that compliance has been demonstrated through the use of an EEEE.
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| Complies with Based on the licensee's justification of continued Use of EEEE validity and evaluation quality, the NRC staff finds the licensee's statement of compliance acceptable.
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| The NRC staff finds that the licensee's proposed Submit For PB method to demonstrate compliance is NRC acceptable as described in SE Section 3.1.3.7 Approval and 3.1.3.10.
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| 3.6 3.6 Standpipe and Hose Stations. Individual elements reviewed below OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATIO~J 143 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staff's Evaluation Statement The NRC staff has previously approved an alternative to this requirement that the licensee is carrying forward into the RI/PB FPP. The NRC staff approved the design of the standpipe and For all power block buildings, Class III standpipe and fire hose systems including modification required Complies Via hose systems shall be installed in accordance with NFPA to the RB hose stations in the NRC SE, Section 3.6.1 Previous NRC 14, Standard for the Installation of Standpipe, Private 4.3.1.4, dated August 11, 1978 (Reference 26)
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| Approval Hydrant, and Hose Systems. and the NRC SER dated June 7, 1988 (Reference 57). Based on the licensee's justification of continued validity, the NRC staff finds the licensee's statement of compliance acceptable.
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| The licensee stated that compliance has been demonstrated through the use of an EEEE.
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| Complies with Based on the licensee's justification of continued Use of EEEE validity and evaluation quality, the NRC staff finds the licensee's statement of compliance acceptable.
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| Submit For The NRC staff finds that the licensee's proposed NRC PB method to demonstrate compliance is Approval acceptable as described in SE Section 3.1.3.7.
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| The licensee stated that compliance has been demonstrated through the use of an EEEE.
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| A capability shall be provided to ensure an adequate Complies with Based on the licensee's justification of continued water flow rate and nozzle pressure for all hose stations.
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| Use of EEEE validity and evaluation quality, the NRC staff finds This capability includes the provision of hose station 3.6.2 the licensee's statement of compliance pressure reducers where necessary for the safety of acceptable.
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| plant industrial fire brigade members and off-site fire Submit For The NRC staff finds that the licensee's proposed department personnel.
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| NRC PB method to demonstrate compliance is Approval acceptable as described in SE Section 3.1.3.7.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 144 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement The proper type of hose nozzle to be supplied to each power block area shall be based on the area fire hazards. The usual combination spray/straight stream nozzle shall not be used in areas where the straight stream can cause unacceptable damage or present an The NRC staff finds the licensee's statement of 3.6.3 Comply electrical hazard to fire-fighting personnel. Listed compliance acceptable.
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| electrically safe fixed fog nozzles shall be provided at locations where high-voltage shock hazards exist. All hose nozzles shall have shutoff capability and be able to control water flow from full open to full closed.
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| Provisions shall be made to supply water at least to The licensee stated that compliance has been standpipes and hose stations for manual fire suppression demonstrated through the use of an EEEE.
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| in all areas containing systems and components needed Complies with Based on the licensee's justification of continued 3.6.4 to perform the nuclear safety functions in the event of a Use of EEEE validity and evaluation quality, the NRC staff finds SSD earthquake (SSE).
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| the licensee's statement of compliance acceptable.
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| [Note: This entry modified per 10 CFR 50.48(c)(2)(vi)1 Where the seismic required hose stations are cross-connected to essential seismic non-fire protection water The NRC staff finds the licensee's statement of 3.6.5 Comply supply systems, the fire flow shall not degrade the compliance acceptable.
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| essential water system requirement.
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| Fire Extinguishers. Where provided, fire extinguishers The licensee stated that compliance has been of the appropriate number, size, and type shall be demonstrated through the use of an EEEE.
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| provided in accordance with NFPA 10, Standard for Complies with Based on the licensee's justification of continued 3.7 Portable Fire Extinguishers. Extinguishers shall be Use of EEEE validity and evaluation quality, the NRC staff finds permitted to be positioned outside of fire areas due to the licensee's statement of compliance radiological conditions. acceptable.
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| 3.8 Fire Alarm and Detection Systems. Individual elements reviewed below.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 145 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement Fire Alarm. Alarm initiating devices shall be installed in accordance with NFPA 72, National Fire Alarm Code.
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| Alarm annunciation shall allow the proprietary alarm The licensee stated that compliance has been system to transmit fire-related alarms, supervisory demonstrated through the use of an EEEE.
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| signals, and trouble signals to the control room or other Complies with Based on the licensee's justification of continued 3.8.1 constantly attended location from which required Use of EEEE validity and evaluation quality, the NRC staff finds notifications and response can be initiated. Personnel the licensee's statement of compliance assigned to the proprietary alarm station shall be acceptable.
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| permitted to have other duties. The following fire-related signals shall be transmitted:
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| The licensee stated that compliance has been demonstrated through the use of an EEEE.
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| Complies with Based on the licensee's justification of continued 3.8.1.(1 ) Actuation of any fire detection device Use of EEEE validity and evaluation quality, the NRC staff finds the licensee's statement of compliance acceptable.
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| The licensee stated that compliance has been demonstrated through the use of an EEEE.
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| Complies with Based on the licensee's justification of continued 3.8.1.(2) Actuation of any fixed fire suppression system Use of EEEE validity and evaluation quality, the NRC staff finds the licensee's statement of compliance acceptable.
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| The licensee stated that compliance has been demonstrated through the use of an EEEE.
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| Complies with Based on the licensee's justification of continued 3.8.1.(3) Actuation of any manual fire alarm station Use of EEEE validity and evaluation quality, the NRC staff finds the licensee's statement of compliance acceptable.
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| OFFICIAL USE O~JLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 146 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement The licensee stated that compliance has been demonstrated through the use of an EEEE.
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| Complies with Based on the licensee's justification of continued 3.8.1.(4) Starting of any fire pump Use of EEEE validity and evaluation quality, the NRC staff finds the licensee's statement of compliance acceptable.
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| The licensee stated that compliance has been demonstrated through the use of an EEEE.
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| Complies with Based on the licensee's justification of continued 3.8.1.(5) Actuation of any fire protection supervisory device Use of EEEE validity and evaluation quality, the NRC staff finds the licensee's statement of compliance acceptable.
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| The licensee stated that compliance has been demonstrated through the use of an EEEE.
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| Complies with Based on the licensee's justification of continued 3.8.1.(6) Indication of alarm system trouble condition Use of EEEE validity and evaluation quality, the NRC staff finds the licensee's statement of compliance acceptable.
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| Means shall be provided to allow a person observing a fire at any location in the plant to quickly and reliably The NRC staff finds the licensee's statement of 3.8.1.1 Comply communicate to the control room or other suitable compliance acceptable.
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| constantly attended location.
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| Means shall be provided to promptly notify the following 3.8.1.2 of any fire emergency in such a way as to allow them to Individual elements reviewed below.
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| determine an appropriate course of action:
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| The NRC staff finds the licensee's statement of 3.8.1.2.(1 ) General site population in all occupied areas. Comply compliance acceptable.
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| Members of the industrial fire brigade and other groups The NRC staff finds the licensee's statement of 3.8.1.2.(2) Comply supporting fire emergency response. compliance acceptable.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 147 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement Off-site fire emergency response agencies. Two The NRC staff finds the licensee's statement of 3.8.1.2.(3) independent means shall be available (e.g., telephone Comply compliance acceptable.
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| and radio) for notification of off-site emerQencv services.
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| The licensee stated that compliance has been demonstrated through the use of an EEEE.
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| Based on the licensee's justification of continued validity and evaluation quality, the NRC staff finds the licensee's statement of compliance acceptable.
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| Detection. If automatic fire detection is required to meet the performance or deterministic requirements of Complies with IMPLEMENTATION ITEM - The licensee 3.8.2 Chapter 4, then these devices shall be installed in Use of EEEE identified an action to complete the development accordance with NFPA 72, National Fire Alarm Code, of compliance calculations for fire detection (SE and its applicable appendixes.
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| Section 2.9, Table 2.9-1, Item 11).
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| MODIFICATION - The licensee evaluated the fire detection coverage and additional detection is required (SE Section 2.8, Table 2.8.1-1, Item 5).
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| Automatic and Manual Water-Based Fire 3.9 Individual elements reviewed below.
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| Suppression Systems.
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| If an automatic or manual water-based fire suppression system is required to meet the performance or 3.9.1 deterministic requirements of Chapter 4, then the system Individual elements reviewed below.
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| shall be installed in accordance with the appropriate NFPA standards includina the following:
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELJ\TED INFORMATION 148 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement The licensee stated that compliance has been demonstrated through the use of an EEEE.
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| Based on the licensee's justification of continued validity and evaluation quality, the NRC staff finds the licensee's statement of compliance NFPA 13, Standard for the Installation of Sprinkler Complies with acceptable.
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| 3.9.1.(1 )
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| Systems Use of EEEE IMPLEMENTATION ITEM - The licensee identified an action to complete the hydraulic calculations for all required automatic or manual water-based suppression systems (SE Section 2.9, Table 2.9-1, Item 12).
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| The licensee stated that compliance has been demonstrated through the use of an EEEE.
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| Based on the licensee's justification of continued validity and evaluation quality, the NRC staff finds the licensee's statement of compliance NFPA 15, Standard for Water Spray Fixed Systems for Complies with acceptable.
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| 3.9.1.(2)
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| Fire Protection Use of EEEE IMPLEMENTATION ITEM - The licensee identified an action to complete the hydraulic calculations for all required automatic or manual water-based suppression systems (SE Section 2.9, Table 2.9-1, Item 12).
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| NFPA 750, Standard on Water Mist Fire Protection The licensee has not credited any of these 3.9.1.(3)
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| Systems systems in LAR Table 4-4.
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| NFPA 16, Standard for the Installation of Foam-Water The licensee has not credited any of these 3.9.1.(4)
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| Sprinkler and Foam-Water Spray Systems systems in LAR Table 4-4.
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| The NRC staff finds the licensee's statement of 3.9.2 Each system shall be equipped with a water flow alarm. Comply compliance acceptable.
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| OFFICIAL USE ONLY SECURITY RELJ\TED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 149 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staff's Evaluation Statement All alarms from fire suppression systems shall The NRC staff finds the licensee's statement of 3.9.3 annunciate in the control room or other suitable Comply compliance acceptable.
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| constantly attended location.
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| Diesel-driven fire pumps shall be protected by automatic The licensee has stated that there are no diesel-3.9.4 sprinklers. driven fire pumps installed at ONS.
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| Each system shall be equipped with an OS&Y gate valve The NRC staff finds the licensee's statement of 3.9.5 Comply or other approved shutoff valve. compliance acceptable.
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| All valves controlling water-based fire suppression systems required to meet the performance or The NRC staff finds the licensee's statement of 3.9.6 Comply deterministic requirements of Chapter 4 shall be compliance acceptable.
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| supervised as described in 3.5.14.
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| 3.10 Gaseous Fire Suppression Systems Individual elements reviewed below If an automatic total flooding and local application gaseous fire suppression system is required to meet the The licensee has not credited any of these 3.10.1 performance or deterministic requirements of Chapter 4, systems in LAR Table 4-4.
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| then the system shall be designed and installed in accordance with the followinq applicable NFPA codes:
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| NFPA 12, Standard on Carbon Dioxide Extinguishing The licensee has not credited any of these 3.10.1.(1)
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| Systems systems in LAR Table 4-4.
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| NFPA 12A, Standard on Halon 1301 Fire Extinguishing The licensee has not credited any of these 3.10.1.(2)
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| Systems systems in LAR Table 4-4.
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| NFPA 2001, Standard on Clean Agent Fire Extinguishing The licensee has not credited any of these 3.10.1.(3)
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| Systems systems in LAR Table 4-4.
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| Operation of gaseous fire suppression systems shall The licensee has not credited any of these 3.10.2 annunciate an alarm in the control room or other systems in LAR Table 4-4.
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| constantly attended location identified.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 150 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement Ventilation system design shall take into account prevention from over-pressurization during agent The licensee has not credited any of these 3.10.3 injection, adequate sealing to prevent loss of agent, and systems in LAR Table 4-4.
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| confinement of radioactive contaminants.
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| In any area required to be protected by both primary and backup gaseous fire suppression systems, a single The licensee has not credited any of these 3.10.4 active failure or a crack in any pipe in the fire systems in LAR Table 4-4.
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| suppression system shall not impair both the primary and backup fire suppression capability.
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| Provisions for locally disarming automatic gaseous The licensee has not credited any of these 3.10.5 suppression systems shall be secured and under strict systems in LAR Table 4-4.
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| administrative control.
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| Total flooding carbon dioxide systems shall not be used The licensee has not credited any of these 3.10.6 in normally occupied areas. systems in LAR Table 4-4.
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| Automatic total flooding carbon dioxide systems shall be equipped with an audible pre-discharge alarm and The licensee has not credited any of these 3.10.7 discharge delay sufficient to permit egress of personnel.
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| systems in LAR Table 4-4.
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| The carbon dioxide system shall be provided with an odorizer.
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| Positive mechanical means shall be provided to lock out The licensee has not credited any of these 3.10.8 total flooding carbon dioxide systems during work in the systems in LAR Table 4-4.
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| protected space.
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| The possibility of secondary thermal shock (cooling) damage shall be considered during the design of any The licensee has not credited any of these 3.10.9 gaseous fire suppression system, but particularly with systems in LAR Table 4-4.
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| carbon dioxide.
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| Particular attention shall be given to corrosive The licensee has not credited any of these 3.10.10 characteristics of agent decomposition products on systems in LAR Table 4-4.
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| safety systems.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 151 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement Passive Fire Protection Features. This section shall be used to determine the design and installation requirements for passive protection features. Passive fire protection features include wall, ceiling, and floor assemblies, fire doors, fire dampers, and through fire 3.11 Individual elements reviewed below barrier penetration seals. Passive fire protection features also include electrical raceway fire barrier systems (ERFBS) that are provided to protect cables and electrical components and equipment from the effects of fire.
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| Building Separation. Each major building within the power block shall be separated from the others by barriers having a designated fire resistance rating of 3 The licensee stated that compliance has been hours or by open space of at least 50 ft (15.2 m) or demonstrated through the use of an EEEE.
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| space that meets the requirements of NFPA BOA, Complies with Based on the licensee's justification of continued 3.11.1 Recommended Practice for Protection of Buildings from Use of EEEE validity and evaluation quality, the NRC staff finds Exterior Fire Exposures. the licensee's statement of compliance Exception: Where a performance-based analysis acceptable.
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| determines the adequacy of building separation, the requirements of 3.11.1 shall not apply.
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| Fire Barriers. Fire barriers required by Chapter 4 shall include a specific fire-resistance rating. Fire barriers The licensee has stated that compliance has shall be designed and installed to meet the specific fire been demonstrated through the use of an EEEE.
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| resistance rating using assemblies qualified by fire tests.
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| Complies with Based on the licensee's justification of continued 3.11.2 The qualification fire tests shall be in accordance with Use of EEEE validity and evaluation quality, the NRC staff finds NFPA 251, Standard Methods of Tests of Fire the licensee's statement of compliance Endurance of Building Construction and Materials, or acceptable.
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| ASTM E 119, Standard Test Methods for Fire Tests of Building Construction and Materials.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 152 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staff's Evaluation Statement Fire Barrier Penetrations. Penetrations in fire barriers The NRC staff finds that the licensee's shall be provided with listed fire-rated door assemblies or explanation of their method of compliance with Complies with listed rated fire dampers haVing a fire resistance rating these requirements acceptable based on the Clarification consistent with the designated fire resistance rating of information provided in the associated LAR B-1 the barrier as determined by the performance Table element.
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| requirements established by Chapter 4. (See 3.11.3.4 for penetration seals for through penetration fire stops.)
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| Passive fire protection devices such as doors and The licensee stated that compliance has been dampers shall conform with the following NFPA demonstrated through the use of an EEEE.
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| standards, as applicable:
| |
| Based on the licensee's justification of continued (1) NFPA 80, Standard for Fire Doors and Fire Windows validity and evaluation quality, the NRC staff finds 3.11.3 (2) NFPA gOA, Standard for the Installation of Air-the licensee's statement of compliance Conditioning and Ventilating Systems acceptable.
| |
| (3) NFPA 101, Life Safety Code Complies with Exception: Where fire area boundaries are not wall-to Use of EEEE MODIFICATION - The licensee has evaluated wall, floor-to-ceiling boundaries with all penetrations the fire barriers separating the power block sealed to the fire rating required of the boundaries, a buildings using Code Conformance Reviews.
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| performance-based analysis shall be required to assess The licensee identified some modification of fire the adequacy of fire barrier forming the fire boundary to doors are required (SE Section 2.8, Table 2.8.1-1, determine if the barrier will withstand the fire effects of Item 3).
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| the hazards in the area. Openings in fire barriers shall be permitted to be protected by other means as acceptable to the AHJ.
| |
| Through Penetration Fire Stops. Through penetration The NRC staff finds the licensee's statement of 3.11.4 Comply fire stops for penetrations such as pipes, conduits, bus compliance acceptable.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 153 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement ducts, cables, wires, pneumatic tubes and ducts, and similar building service equipment that pass through fire The NRC staff has previously approved an barriers shall be protected as follows: alternative to this requirement that the licensee is (a) The annular space between the penetrating item and carrying forward into the RI/PB FPP. The NRC Complies by the through opening in the fire barrier shall be filled with staff has approved specific deviations regarding Previous a qualified fire-resistive penetration seal assembly the fire penetration stops in NRC SE dated NRC capable of maintaining the fire resistance of the fire August 21, 1989, Reference 27.. Based on the Approval barrier. The assembly shall be qualified by tests in licensee's justification of continued validity, the accordance with a fire test protocol acceptable to the NRC staff finds the licensee's statement of AHJ or be protected by a listed fire-rated device for the compliance acceptable.
| |
| specified fire-resistive period.
| |
| (b) Conduits shall be provided with an internal fire seal that has an equivalent fire-resistive rating to that of the fire barrier through opening fire stop and shall be The licensee stated that compliance has been permitted to be installed on either side of the barrier in a demonstrated through the use of an EEEE.
| |
| location that is as close to the barrier as possible. Based on the licensee's justification of continued Exception: Openings inside conduit 4 in. (10.2 cm) or validity and evaluation quality, the NRC staff finds less in diameter shall be sealed at the fire barrier with a Complies with the licensee's statement of compliance fire-rated internal seal unless the conduit extends greater Use of EEEE acceptable.
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| than 5 ft (1.5 m) on each side ofthe fire barrier. In this case the conduit opening shall be provided with MODIFICATION - Modification of penetrations noncombustible material to prevent the passage of seals are required (SE Section 2.8, Table 2.8.1-1, smoke and hot gases. The fill depth of the material Items 2 and 3).
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| packed to a depth of 2 in. (5.1 cm) shall constitute an acceptable smoke and hot aas seal in this aoolication.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 154 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement Electrical Raceway Fire Barrier Systems (ERFBS).
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| ERFBS required by Chapter 4 shall be capable of resisting the fire effects of the hazards in the area.
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| ERFBS shall be tested in accordance with and shall meet the acceptance criteria of NRC Generic Letter 86 10, Supplement 1, "Fire Endurance Test Acceptance Criteria for Fire Barrier Systems Used to Separate SSD The licensee has stated that there are no ERFBS 3.11.5 Trains Within the Same Fire Area." The ERFBS needs to credited at ONS.
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| adequately address the design requirements and limitations of supports and intervening items and their impact on the fire barrier system rating. The fire barrier system's ability to maintain the required nuclear safety circuits free of fire damage for a specific thermal exposure, barrier design, raceway size and type, cable size, fill, and type shall be demonstrated.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 155 Attachment A, NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix Table 3.1-1 NFPA 805, Chapter 3, Fundamental Elements Compliance Matrix ONS Element NFPA 805 Requirement Compliance NRC Staffs Evaluation Statement Exception NO.1: When the temperatures inside the fire barrier system exceed the maximum temperature allowed by the acceptance criteria of Generic Letter 86 10, "Fire Endurance Acceptance Test Criteria for Fire Barrier Systems Used to Separate Redundant SSD Training Within the Same Fire Area," Supplement 1, functionality of the cable at these elevated temperatures shall be demonstrated. Qualification demonstration of these cables shall be performed in accordance with the electrical testing requirements of Generic Letter 86-10, Supplement 1, Attachment 1, "Attachment Methods for Demonstrating Functionality of Cables Protected by Raceway Fire Barrier Systems During and After Fire Endurance Test Exposure."
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| Exception NO.2: ERFBS systems employed prior to the issuance of Generic Letter 86-10, Supplement 1, are acceptable providing that the system successfully met the limiting end point temperature requirements as specified by the AHJ at the time of acceptance.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 156 Attachment B, Nuclear Safety Capability Assessment Method Review Regulatory Guide 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light Water Nuclear Power Plants," Revision 1, endorses, with certain exceptions, NEI 04-02, "Guidance for Implementing a Risk-Informed, Performance-Based FPP Under 10 CFR 50.48(c)," Revision 2, and Chapter 3 of t\lEI 00-01, Revision 2, "Guidance for Post-fire SSD Circuit Analysis", and promulgates the method outlined in NEI 04-02 for conducting an NSCA.
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| This NRC-endorsed method documents in a table format (i.e., NEI 04-02 Table B-2, "NFPA 805 Chapter 2 - Nuclear Safety Transition - Methodology Review") the licensee's comparison of its post-fire SSD analyses to the guidance in NEI 00-01, Chapter 3, which has been determined to address the related requirements of NFPA 805, Section 2.4.2, "Nuclear Safety Capability Assessment."
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| This attachment contains Table 3.2-1, which identifies each applicable NEI 00-01 guidance section, documents whether the licensee stated that it met the guidance, or provided justification for meeting the intent of that guidance or not meeting the guidance, and presents the staff's evaluation of each NEI 00-01, Chapter 3 attribute for which the licensee stated its process/justification for meeting the intent of the guidance or not meeting the guidance.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 157 Attachment B, Nuclear Safety Capability Assessment Method Review Table 3.2-1: Nuclear Safety Capability Assessment Method Review NE100-01 Section Title ONS Alignment Basis NRC Staffs Evaluation Section 3.0 Deterministic Aligns with intent - ONS states that it By letter dated July 30, 2010 (Reference 43), the staff Methodology conforms to NEI 00-01, Revision 1, with requested the licensee to clarify differences in Alignment certain exceptions, as noted in the individual Bases statements of the April 2010 LAR (Reference 11) paragraph or section comparisons below. and the October 2008 LAR (Reference 2) and, where necessary, provide additional information to readily conclude that each sub-criterion of the NEI 00-01 section has been satisfied. In its response, dated September 13, 2010 (Reference 12), the licensee identifies 55 sections of NEI 00-01 where the Alignment Basis was found to differ between submittals. Of these, changes made to 13 sections, it was determined to have resulted in a lack of detail needed to confirm alignment with the NEI guidance. The licensee states that for each of these 13 sections, additional details will be added back to the alignment basis. The specific Sections of NEI 00-01 requiring revision and proposed changes are identified in Reference 12 and are identified as an implementation item (SE Section 2.9, Table 2.9-1, Item 38).
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| As documented in the detailed discussions for each of the table attributes below, the NRC staff finds that the licensee has, in most instances, achieved either alignment with the NEI 00-01 guidance document, or alignment with the intent of the guidance. For those attributes that do not align or align with intent, the licensee has described the process and actions being taken to bring the attribute into alignment. The NRC staff finds this approach acceptable.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 158 Attachment 8, Nuclear Safety Capability Assessment Method Review Table 3.2-1: Nuclear Safety Capability Assessment Method Review NE100-01 Section Title ONS Alignment Basis NRC Staff's Evaluation Section 3.1 [A] SSD Systems and Aligns The NRC staff finds the licensee's statement of Intro Path Development alignment to the endorsed guidance acceptable.
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| 3.1 [B) SSD Systems and Aligns The NRC staff finds the licensee's statement of Goals Path Development alignment to the endorsed guidance acceptable.
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| 3.1 [C) SSD Systems and Aligns The NRC staff finds the licensee's statement of Spurious Path Development alignment to the endorsed guidance acceptable.
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| Operation 3.1.1 Criteria/Assumption Detailed aliQnment discussed below. Individual elements reviewed below.
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| 3.1.1.1 GE BWR Paths N/A N/A 3.1.1.2 SRVs / LP Systems N/A N/A 3.1.1.3 PWR Pressurizer Aligns The NRC staff finds the licensee's statement of Heaters alignment to the endorsed guidance acceptable.
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| 3.1.1.4 Alternative Aligns with intent. The transfer of control to The NRC staff agrees that the intent of the guidance is to Shutdown the SSF isolates required systems and ensure that following transfer of control to the SSF, Capability equipment from the effects of a fire for the cables and equipment credited for shutdown (alternative fire areas of concern. The intent of the or dedicated) are independent of the fire area of guidance is that dedicated cables and concern.
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| equipment is independent of the fire area of concern. Following transfer of control to the The NRC staff finds the licensee's statement of SSF, the dedicated equipment credited for alignment to the endorsed guidance acceptable.
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| an SSF shutdown meets the intent of the guidance.
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| 3.1.1.5 Initial Conditions Aligns The NRC staff finds the licensee's statement of alignment to the endorsed guidance acceptable.
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| 3.1.1.6 Other Events in Aligns The NRC staff finds the licensee's statement of Conjunction with alignment to the endorsed guidance acceptable.
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| Fire OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 159 Attachment B, Nuclear Safety Capability Assessment Method Review Table 3.2-1: Nuclear Safety Capability Assessment Method Review NE100-01 Section Title ONS Alignment Basis NRC Staffs Evaluation Section 3.1.1.7 Offsite Power Aligns With Intent. Oconee relies upon By letter dated July 30, 2010 (Reference 43), the Keowee hydro station to provide emergency licensee clarified an apparent discrepancy between the onsite power. The cascading power supply information presented in Table B-2 of the October 2008 analysis determines fire impact to offsite LAR (Reference 2) and the April 2010 LAR (Reference power sources and is utilized in the analysis 11). Specifically, Section 3.1.1.7 of the October 2008 of fire areas for SSD functions to determine LAR B-2 Table states that offsite power has not been availability of all credited power sources. analyzed or demonstrated to be free of fire damage for The adverse consequences of offsite power redundant shutdown. However, the April 2010 LAR being available is considered in the NSCA. Alignment Basis indicates that the availability of offsite power has, in fact, been analyzed. In its response, dated September 13, 2010 (Reference 12), the licensee states that the credited power supplies are the Keowee Hydro Station (KHS) and the SSF DG and neither the KHS nor the SSF DG requires offsite power. The licensee also states that the adverse consequences of offsite power being available are considered in the NSCA. To ensure clarity, the licensee has created an action item in the ONS corrective action program (CAP) to revise calculation OSC- 9291, NFPA 805 Transition B 2 Table, Section 3.1.1.7 to reword the alignment basis to clearly state that offsite power is not credited for the deterministic analysis and therefore not analyzed for its availability in the deterministic analysis. The licensee also states that alignment statement will also be revised to ensure the proper relationship with the alignment basis. Completion of the CAP item is an implementation item (SE Section 2.9, Table 2.9-1, Item 47). Based on the response provided in the September 13, 2010 letter (Reference 12), the staff finds the licensee's statement of aliQnment to the endorsed guidance acceptable.
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| OFFICIAL USE ONLY SECURITY RELJ\TED INFORMATION 160 Attachment B, Nuclear Safety Capability Assessment Method Review Table 3.2-1: Nuclear Safety Capability Assessment Method Review NE100-01 Section Title ONS Alignment Basis NRC Staffs Evaluation Section 3.1.1.8 Safety Related Aligns The NRC staff finds the licensee's statement of Equipment alignment to the endorsed guidance acceptable.
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| 3.1.1.9 72-Hour Coping The licensee stated that the approach used See NRC staff evaluation for SE Section 3.2.1, Item 5.
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| aligns with the intent of the 72-hour coping Based on the information provided in the LAR, as criterion specified in the NEI 00-01 supplemented, the NRC staff finds that the licensee's guidance. NFPA 805 does not have any approach to demonstrate the capability to maintain the explicit requirements to achieve cold plant in a safe and stable condition following a fire is shutdown within 72 hours; therefore, the acceptable.
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| NFPA-805 criteria for nuclear safety performance goals have been applied to ensure the fuel is maintained safe and stable.
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| 3.1.1.10 Manual I Automatic The licensee stated that the approach used Manual initiation of components and systems may Initiation of aligns with the intent of the NEI 00-01 provide an acceptable compliance strategy where Systems criterion. Manual initiation of components permitted by current regulations or approved by NRC.
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| and systems from either the MCR or The NRC staff finds the licensee's statement of Emergency (local) control stations have alignment to the endorsed guidance acceptable.
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| been credited as acceptable compliance strategies where permitted by current regulations or approved by NRC. Automatic initiation of components is not credited.
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| 3.1.1.11 Multiple Affected The licensee stated that the approach used Although Oconee shares some equipment between Units aligns with the intent of the NEI 00-01 units, the evaluation considered the impact of fire criterion. Oconee shares some equipment damage on each unit. The NRC staff finds the licensee's between units. Fire impacts at the statement of alignment to the endorsed guidance component level have been evaluated for acceptable.
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| impact on each unit.
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| 3.1.2 Shutdown Detailed alignment discussed below. See NRC staff's evaluation for individual subsections Functions below.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 161 Attachment B, Nuclear Safety Capability Assessment Method Review Table 3.2-1: Nuclear Safety Capability Assessment Method Review NE100-01 Section Title ONS Alignment Basis NRC Staff's Evaluation Section 3.1.2.1 Reactivity Control Aligns The NRC staff's finds the licensee's statement of alignment to the endorsed guidance acceptable.
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| 3.1.2.2 Pressure Control The licensee stated that the approach used The systems discussed in this section of NEI 00-01 are Systems aligns with the intent of the NEI 00-01 examples of systems that can be used for pressure criterion. Pressure control is accomplished control. This does not restrict the use of other systems utilizing reactor makeup from the SSF or for this purpose. The licensee states that an assured injection from HPI in conjunction with success path is determined for each fire area. The NRC pressurizer heaters, safety relief valves, staff finds the licensee's statement of alignment to the PORV's, RCS loop high point vent valves, endorsed guidance acceptable.
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| or reactor head vent valves and controlling decay heat removal rates. An assured success path is determined during the Fire Area Analysis.
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| 3.1.2.3 Inventory Control Aligns The NRC staff finds the licensee's statement of alignment to the endorsed guidance acceptable.
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| 3.1.2.4 Decay Heat The licensee stated that the approach used The NRC staff agrees that NFPA 805 does not have any Removal aligns with the intent of the NEI 00-01 explicit requirements to achieve cold shutdown, therefore criterion. SG(s) are fed from emergency the NFPA 805 criteria for the Nuclear Safety feed water to remove decay heat under Performance Goals have been applied to ensure the fuel natural circulation conditions. Main steam is maintained in a safe and stable condition. In addition, safety valves are utilized for decay heat there is no restriction on the use of systems other than removal in hot standby. NFPA 805 does not those identified in this criterion. The NRC staff finds the have any explicit requirements to achieve licensee's statement of alignment to the endorsed cold shutdown, therefore the NFPA 805 guidance acceptable.
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| criteria for the Nuclear Safety Performance Goals have been applied to ensure the fuel is maintained in a safe and stable condition.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 162 Attachment B, Nuclear Safety Capability Assessment Method Review Table 3.2-1: Nuclear Safety Capability Assessment Method Review NE100-01 Section Title ONS Alignment Basis NRC Staffs Evaluation Section 3.1.2.5 Process Monitoring The licensee stated that the approach used NEI 00-01 refers to Information Notice 84-09 Attachment to address the process monitoring function 1, Section IX, which specifies the process variables, aligns with the intent of the NEI 00-01 deemed necessary to perform and control the reactor criterion because an exemption for boron shutdown functions. The specific instruments provided sampling in lieu of neutron source range may be based on operator preference, SSD procedural monitoring instrumentation has been guidance strategy (symptomatic vs. prescriptive), and granted for the SSF, therefore neutron systems and paths selected for SSD.
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| source range instrumentation has not been provided or analyzed for the SSF. SG In its letter dated April 14, 2010, (Reference 11), the pressure instruments are also not provided licensee identified the process monitoring and diagnostic in the SSF and was accepted by the NRC. instrumentation credited for SSD from both the MCR and Both Neutron Instrumentation and SG SSF. The licensee further states that the variances from pressure indication are provided in the Main the process variables identified in Information Notice (IN)
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| Control Room. 84-09 are consistent with its license basis to the extent that the use of boron sampling and a lack of SG pressure instruments at the SSF have been previously approved in an SE dated August 31, 1983 (Reference 40). This approved exemption is being carried forward into the RI/PB FPP and the licensee's statement that the original basis for the exemption remains valid was found acceptable by the NRC staff (see SE Attachment D).
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| Based on the above, the NRC staff finds the licensee's statement of alignment to the endorsed guidance acceptable.
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| 3.1.2.6 Support Systems Detailed alignment discussed below. See NRC staff evaluation for subsections below.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 163 Attachment B, Nuclear Safety Capability Assessment Method Review Table 3.2-1: Nuclear Safety Capability Assessment Method Review NE100-01 Section Title ONS Alignment Basis NRC Staffs Evaluation Section 3.1.2.6.1 Electrical Systems The licensee stated that the approach used In lieu of using emergency DGs, the licensee will rely aligns with the intent of the NEI 00-01 upon the KHS as an emergency onsite power source. In criterion. Emergency onsite power is addition, the licensee states that alternating current (AC) provided from the KHS on the site. and direct connect (DC) power supplies have been analyzed and both the SSF and credited plant battery chargers will be available. The NRC staff finds the licensee's statement of alignment to the endorsed guidance acceptable.
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| 3.1.2.6.2 Cooling Systems Aligns The NRC staff finds the licensee's statement of
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| [Main Section] alignment to the endorsed guidance acceptable.
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| 3.1.2.6.2 Cooling Systems Aligns The NRC staff finds the licensee's statement of
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| [HVAC] alignment to the endorsed guidance acceptable.
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| 3.1.3 Methodology for Detailed alignment discussed below. See NRC staff's evaluation for subsections below.
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| Shutdown System Selection 3.1.3.1 Identify SSD Aligns The NRC staff finds the licensee's statement of Functions alignment to the endorsed guidance acceptable.
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| 3.1.3.2 Identify Aligns The NRC staff finds the licensee's statement of Combinations of alignment to the endorsed guidance acceptable.
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| Systems that Satisfy Each SSD Function 3.1.3.3 Define Aligns The NRC staff finds the licensee's statement of Combinations of alignment to the endorsed guidance acceptable.
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| Systems for Each SSD Path 3.1.3.4 Assign Shutdown Aligns The NRC staff finds the licensee's statement of Paths to Each alignment to the endorsed guidance acceptable.
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| Combination of Systems OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 164 Attachment B, Nuclear Safety Capability Assessment Method Review Table 3.2-1: Nuclear Safety Capability Assessment Method Review NE100-01 Section Title ONS Alignment Basis NRC Staffs Evaluation Section 3.2 SSD Equipment Detailed alignment discussed below. See NRC staff's evaluation for subsections below.
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| Selection 3.2.1 Criteria/ Detailed alignment discussed below. See NRC staff's evaluation for subsections below.
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| Assumptions 3.2.1.1 Primary Secondary The licensee stated that the approach used The licensee stated that there was no segregation of Components aligns with the intent of the NEI 00-01 SSEL components. Some of the components defined as criterion. secondary were captured by the cable selection process and others are captured within the cascading interlocks analysis as pseudo-components. The NRC staff finds the licensee's statement of alignment to the endorsed guidance acceptable.
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| 3.2.1.2 Fire Damage to Aligns The NRC staff finds the licensee's statement of Mechanical alignment to the endorsed guidance acceptable.
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| Components (not electrically supervised 3.2.1.3 Manual Valve Aligns The NRC staff finds the licensee's statement of Positions alignment to the endorsed guidance acceptable.
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| 3.2.1.4 Check Valves Aligns The NRC staff finds the licensee's statement of alignment to the endorsed guidance acceptable.
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| 3.2.1.5 Instrument Failures Aligns The NRC staff finds the licensee's statement of alignment to the endorsed guidance acceptable.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 165 Attachment 8, Nuclear Safety Capability Assessment Method Review Table 3.2-1: Nuclear Safety Capability Assessment Method Review NE100-01 Section Title ONS Alignment Basis NRC Staffs Evaluation Section 3.2.1.6 Spurious The licensee stated that the approach used The licensee states that spurious operation was Components aligns with the intent of the NEI 00-01 considered in identification of SSEL components. RIS criterion that states that equipment that 2004-03 (Reference 45) was referenced, however no could spuriously operate should be initial effort was made to 'bin' the types of potential identified in the equipment selection phase spuriously operating components or their cables.
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| and that Bin 1 of RIS 2004-03 should be Spurious operation was considered later during considered during the equipment compliance assessment when circuit analysis was identification process. performed to determine if potential spurious operation was a concern requiring mitigating actions or other compliance strategies. The NRC staff finds the licensee's statement of alignment to the endorsed guidance acceptable.
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| 3.2.1.7 Instrument Tubing Aligns The NRC staff finds the licensee's statement of alignment to the endorsed guidance acceptable.
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| 3.2.2 Methodology for Detailed alignment discussed below. See NRC staff's evaluation for subsections below.
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| Equipment Selection 3.2.2.1 Identify the System The licensee stated that the approach used Flow and diversion paths were identified and translated Flow Path for Each to identify the system flow paths of SSD into SSD Success path logic diagrams. The NRC staff Shutdown Path components aligns with the intent of the finds the licensee's statement of alignment to the requirement since the piping and instrument endorsed guidance acceptable.
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| drawings (P&ID's) were marked up to determine flow and diversion paths which were then translated into SSD Success path logic diagrams. These logic diagrams were then used to identify potential SSEL components.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 166 Attachment 8, Nuclear Safety Capability Assessment Method Review Table 3.2-1: Nuclear Safety Capability Assessment Method Review NE100-01 Section Title ONS Alignment Basis NRC Staffs Evaluation Section 3.2.2.2 Identify the The licensee stated that the approach used The approach used is functionally equivalent to that Equipment in Each to identify the system flow paths of SSD specified in NEI 00-01. The NRC staff finds the SSD System Flow components aligns with the intent of the licensee's statement of alignment to the endorsed Path Including requirement since P&ID's and electrical one guidance acceptable.
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| Equipment That lines were marked up to determine flow and May Spuriously diversion paths for SSD functions and to Operate and Affect identify potential SSEL components System Operation including spurious operations. SSD success paths were then translated into SSD Loqic Diaqrams.
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| 3.2.2.3 Develop a SSD The licensee stated that the approach used The approach implemented by the licensee is Equipment List and to develop the SSD Equipment List (SSEL) functionally equivalent to that specified in NEI 00-01.
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| Assign the aligns with the intent of this requirement. The NRC staff finds the licensee's statement of Corresponding P&ID's were marked up to determine flow alignment to the endorsed guidance acceptable.
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| System and SSD and diversion paths for SSD functions and Path(s) to Each to identify potential SSEL components including spurious operations. An iterative process was utilized to arrive at the final SSEL based on additional support components identified during the cable selection process. NEI 00-01 Attachment 3 was not utilized, since the SSD database has its own data entry format, which provides the necessary equipment information.
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| 3.2.2.4 Identify Equipment Aligns The NRC staff finds the licensee's statement of Information alignment to the endorsed guidance acceptable.
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| Required for the SSD Analysis OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE O~JLY SECURITY RELATED INFORMATION 167 Attachment B, Nuclear Safety Capability Assessment Method Review Table 3.2-1: Nuclear Safety Capability Assessment Method Review NE100-01 Section Title ONS Alignment Basis NRC Staffs Evaluation Section 3.2.2.5 Identify Aligns The NRC staff finds the licensee's statement of Dependencies alignment to the endorsed guidance acceptable.
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| Between Equipment, Supporting Equipment, SSD Systems and SSD Paths 3.3 SSD Cable Detailed alignment discussed below. See NRC staff's evaluation for subsections below.
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| Selection and Location 3.3.1 Criteria! Detailed alignment discussed below. See NRC staff's evaluation for subsections below.
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| Assumptions 3.3.1.1 Cable Selection Aligns The NRC staff finds the licensee's statement of aliqnment to the endorsed quidance acceptable.
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| 3.3.1.2 Cables Affecting Aligns The NRC staff finds the licensee's statement of Multiple alignment to the endorsed guidance acceptable.
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| Components 3.3.1.3 Isolation Devices Aligns The NRC staff finds the licensee's statement of alignment to the endorsed guidance acceptable.
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| 3.3.1.4 Identify "Not Aligns The NRC staff finds the licensee's statement of Required" Cables alignment to the endorsed guidance acceptable.
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| 3.3.1.5 Identification of Aligns The NRC staff finds the licensee's statement of Power Supplies aliqnment to the endorsed quidance acceptable.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 168 Attachment B, Nuclear Safety Capability Assessment Method Review Table 3.2-1: Nuclear Safety Capability Assessment Method Review NE100-01 Section Title ONS Alignment Basis NRC Staff's Evaluation Section 3.3.1.6 Engineered safety The licensee stated that the approach used The approach used to evaluate the impact of the ESFAS features actuation aligns with the intent of this requirement. initiation is functionally equivalent to NEI 00-01. The system (ESFAS) Automatic initiation logic was not credited licensee states that VFDRs were identified and Initiation for performance of SSD functions. Manual evaluated in the FREs to assess the impact of the VFDR operation of components from the Main and any necessary recovery actions to mitigate the Control Room, SSF or locally were identified effects of the VFDR. The NRC staff finds the licensee's during the fire area compliance assessment statement of alignment to the endorsed guidance task as needed. To preclude adverse acceptable.
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| impact from automatic initiation logic circuits or control logic circuits where multiple components receive signals from common control logic, the control logic was analyzed as a primary component and a pseudo component was created for the logic with cables selected accordingly. This same methodology was used for similar circuit scenarios such as common power supplies.
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| In this way the effects of a fire-induced failure causing spurious component operation were fully evaluated.
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| 3.3.1.7 Circuit Coordination Does Not Align. Based on the LAR, as supplemented, the NRC staff finds Proper coordination of common power that the licensee's approach addressed the issue of supplies for all circuits was an assumption inadequate breaker coordination is acceptable and that of the analysis. ONS existing coordination the licensee states ONS will comply with the study does not include all SSEL related requirements of NFPA 805, Section 2.4.2.2.2.(a) upon power supplies. The coordination study completion of the committed plant modifications and needs to be updated with the additional implementation items. (see SE Sections 2.8, power supplies to ensure that the Table 2.8.1-1 and 2.9, Table 2.9-1).
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| assumptions of the EIR remain valid.
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| 3.3.2 Associated Circuit Detailed alignment discussed below. See NRC staff's evaluation for subsections below.
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| Cables OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 169 Attachment B, Nuclear Safety Capability Assessment Method Review Table 3.2-1: Nuclear Safety Capability Assessment Method Review NE100-01 Section Title ONS Alignment Basis NRC Staffs Evaluation Section 3.3.2 [A] Associated Circuit Not a review criterion - Generic paragraph. See NRC staff's evaluation for Section 3.3.3 below.
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| Cables - Cables This section only describes spurious Whose Failure May actuation concern. Section 3.3.3 below Cause Spurious addresses the methodology for selecting Actuations cables whose failure may cause spurious operations.
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| 3.3.2 [B] Associated Circuit Not a review criterion - Generic paragraph. See NRC staff's evaluation for Section 3.5.2.4 below.
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| Cables - Common Provides brief description of common power Power Source source concern. Section 3.5.2.4 below Cables addresses the methodology for analyzing circuit failures due to inadequate circuit coordination.
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| 3.3.2 [C] Associated Circuit Not a review criterion - Generic paragraph. See NRC staff's evaluation for Section 3.5.2.5 below.
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| Cables - Common Provides brief description of common Enclosure Cables enclosure concern. Section 3.5.2.5 below addresses the methodology for analyzing circuit failures due to common enclosure concerns.
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| 3.3.3 Methodology for Detailed alignment discussed below. See NRC staff's evaluation for subsections below.
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| Cable Selection and Location 3.3.3.1 Identify Circuits Aligns The NRC staff finds the licensee's statement of Required for the alignment to the endorsed guidance acceptable.
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| Operation of the SSD Equipment OFFICIAL USE O~JLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 170 Attachment 8, Nuclear Safety Capability Assessment Method Review Table 3.2-1: Nuclear Safety Capability Assessment Method Review NE100-01 Section Title ONS Alignment Basis NRC Staffs Evaluation Section 3.3.3.2 Identify Interlocked The licensee stated that the approach used The approach used by the licensee addresses Circuits and Cables aligns with the intent of this requirement; for interlocked components. The staff finds this acceptable Whose Spurious control logic circuits where multiple based on the description of the process used, where a Operation or Mal- components receive signals from common deviation from the endorsed guidance occurs, a operation Could control logic or interlocks, the control logic conservative assumption was used in the circuit analysis Affect Shutdown was analyzed as a primary component and (interlocked contact or relay will be assumed to be in the a pseudo component was created on the worst-case position). The NRC staff finds the licensee's SSEL for the logic with cables selected statement of alignment to the endorsed guidance accordingly. Pseudo-components whose acceptable.
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| associated cabling can affect another primary component based on common power were identified in the cable selection for the affected component as an interlocked primary component. The cascading power supply and cascading interlocks analyses evaluate these interlocked components.
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| 3.3.3.3 Assign Cables to Does not Align Based on the information provided in the LAR, as the SSD Equipment Coordination of power supplies was supplemented, the NRC staff finds that the licensee's assumed when assigning cables to the SSD approach has adequately addressed the issue of equipment; Oconee does not meet the inadequate breaker coordination and that, upon intent of the guidance since it did not completion of the modifications and implementation consider inadequate breaker coordination items, ONS will align with the endorsed guidance for this when selecting cables. attribute. (See SE Section 3.2.1).
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| 3.4 Fire Area Detailed alignment discussed below. See NRC staff's evaluation for subsections below.
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| assessment and Compliance Assessment 3.4.1 Criteria/Assumption Aligns The NRC staff finds the licensee's statement of alignment to the endorsed guidance acceptable.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 171 Attachment B, Nuclear Safety Capability Assessment Method Review Table 3.2-1: Nuclear Safety Capability Assessment Method Review NE100-01 Section Title ONS Alignment Basis NRC Staffs Evaluation Section 3.4.1.1 Number of Aligns The NRC staff finds the licensee's statement of Postulated Fires alignment to the endorsed guidance acceptable.
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| 3.4.1.2 Damage to Aligns The NRC staff finds the licensee's statement of Unprotected alignment to the endorsed guidance acceptable.
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| Equipment and Cables 3.4.1.3 Assess Impacts to Aligns The NRC staff finds the licensee's statement of Required alignment to the endorsed guidance acceptable.
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| Components 3.4.1.4 Manual Actions The licensee stated that the approach used The licensee states variances from the nuclear safety aligns with the intent of this requirement; the performance criteria deterministic approach were least impacted SSD success path was evaluated as a FRE per Section 4.2.4.2 of NFPA 805. If analyzed and Variances from the the FRE meets the acceptance criteria, this is Deterministic Requirements (VFDRs) were confirmation that a success path effectively remains free identified. Mitigating strategies to address of fire damage and that the PB approach is acceptable the VFDRs in a PB FRE were developed per Section 4.2.4.2 of NFPA 805. The NRC staff finds and documented. One of the potential the licensee's statement of alignment to the endorsed mitigating strategies is procedural action guidance acceptable.
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| (recovery action) to mitigate the operational effects from fire damage.
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| 3.4.1.5 Repairs The licensee stated that the approach used The licensee states variances from the nuclear safety aligns with the intent of this requirement; performance criteria deterministic approach were The least impacted SSD success path was evaluated as a FRE per Section 4.2.4.2 of NFPA 805. If analyzed and Variances from the the FRE meets the acceptance criteria, this is Deterministic Requirements (VFDRs) were confirmation that a success path effectively remains free identified. Mitigating strategies to address of fire damage and that the PB approach is acceptable the VFDRs in a PB FRE were developed per Section 4.2.4.2 of NFPA 805. NFPA 805 does not and documented. One of the potential have any explicit requirements to achieve cold mitigating strategies is repairs (recovery shutdown. The NRC staff finds the licensee's statement action) to mitigate the operational effects of alignment to the endorsed guidance acceptable.
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| from fire damaqe.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 172 Attachment B, Nuclear Safety Capability Assessment Method Review Table 3.2-1: Nuclear Safety Capability Assessment Method Review NE100-01 Section Title ONS Alignment Basis NRC Staffs Evaluation Section 3.4.1.6 Assess Compliance The licensee stated that the approach used The licensee states variances from the nuclear safety with Deterministic aligns with the intent of this requirement; performance criteria deterministic approach were Criteria The least impacted SSD success path was evaluated as a FRE per Section 4.2.4.2 of NFPA 805. If analyzed and Variances from the the FRE meets the acceptance criteria, this is Deterministic Requirements (VFDRs) were confirmation that a success path effectively remains free identified. Mitigating strategies to address of fire damage and that the PB approach is acceptable the VFDRs in a PB FRE were developed per Section 4.2.4.2 of NFPA 805. The NRC staff finds and documented. The methods described the licensee's statement of alignment to the endorsed above are options to satisfy the guidance acceptable.
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| deterministic criteria to preclude identification of VFDRs.
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| 3.4.1.7 Consider Additional The licensee stated that the approach used The licensee states variances from the nuclear safety Equipment aligns with the intent of this requirement; the performance criteria deterministic approach were least impacted SSD success path was evaluated as a FRE per Section 4.2.4.2 of NFPA 805. If analyzed and Variances from the the FRE meets the acceptance criteria, this is Deterministic Requirements (VFDRs) were confirmation that a success path effectively remains free identified. Mitigating strategies to address of fire damage and that the PB approach is acceptable the VFDRs in a PB FRE were developed per Section 4.2.4.2 of NFPA 805. The NRC staff finds and documented. The methods described the licensee's statement of alignment to the endorsed above are options to satisfy the guidance acceptable.
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| deterministic criteria to preclude identification of VFDRs.
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| 3.4.1.8 Consider Aligns The NRC staff finds the licensee's statement of Instrument Tubing alignment to the endorsed guidance acceptable.
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| Effects 3.4.2 Methodology for Detailed alignment discussed below. See NRC staff's evaluation for subsections below.
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| Fire Area Assessment 3.4.2.1 Identify the Affected Aligns The NRC staff finds the licensee's statement of Equipment By Fire alignment to the guidance acceptable.
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| Area OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 173 Attachment B, Nuclear Safety Capability Assessment Method Review Table 3.2-1: Nuclear Safety Capability Assessment Method Review NE100-01 Section Title ONS Alignment Basis NRC Staffs Evaluation Section 3.4.2.2 Determine the Aligns The NRC staff finds the licensee's statement of Shutdown Paths alignment to the endorsed guidance acceptable.
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| Least Impacted by a Fire in Each Fire Area 3.4.2.3 Determine SSD The licensee stated that the approach used The staff was concerned that the MSO analysis was Equipment Impacts aligns with the intent of this requirement; limited to only SSD cables and components. In its The SSEL and logics were developed October 14,2010 response (Reference 54), the licensee based on potential spurious operations and states that the methodology assumed multiple fire-other plant impacts by their selection from a induced failures and multiple spurious actuations based functional basis. The fire area analysis on the cables and components present in the fire area of methodology assumes multiple fire-induced concern, and was not limited to SSD cables and failures and multiple spurious actuations, components.
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| based on the SSD cables and components present in the fire area of concern. All Based on the information provided in the LAR, as postulated SSD cable and component supplemented, the NRC staff finds that the licensee's failures were identified and a resolution basis for alignment to Section 3.4.2.3 of NEI 00-01 provided at the cable or component level for acceptable.
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| the credited train.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 174 Attachment B, Nuclear Safety Capability Assessment Method Review Table 3.2-1: Nuclear Safety Capability Assessment Method Review NE100-01 Section Title ONS Alignment Basis NRC Staffs Evaluation Section 3.4.2.4 Develop a The licensee stated that the approach used The approach used to determine SSD equipment Compliance aligns with the intent of this requirement; the impacts is functionally equivalent to NEI 00-01. The Strategy or SSD success paths were analyzed and licensee states variances from the nuclear safety Disposition to potential impacts identified. These potential performance criteria deterministic approach were Mitigate the Effects impacts were resolved such that the least evaluated as a FRE per Section 4.2.4.2 of NFPA 805. If Due to Fire impacted SSD success path could be the FRE meets the acceptance criteria, this is Damage to Each identified. Variances from the Deterministic confirmation that a success path effectively remains free Required Requirements (VFDRs) were identified. of fire damage and that the PB approach is acceptable Component or Mitigating strategies to address the VFDRs per Section 4.2.4.2 of NFPA 805. The NRC staff finds Cable in a PB FRE were developed and the licensee's statement of alignment to the endorsed documented. Credit for existing features guidance acceptable.
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| and exemptions was taken wherever possible and procedural (recovery) action specified as a last resort.
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| 3.4.2.5 Document the Aligns The NRC staff finds the licensee's statement of Compliance alignment to the endorsed guidance acceptable.
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| Strategy or Disposition Determined to Mitigate the Effects Due to Fire Damage to Each Required Component or Cable 3.5 Circuit Analysis and Not Required. Generic paragraph. Detailed See NRC staff's evaluation for individual subsections Evaluation alignment discussed below. below.
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| 3.5.1 Criteria/Assumption Detailed alignment discussed below. See NRC staff's evaluation for subsections below.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 175 Attachment B, Nuclear Safety Capability Assessment Method Review Table 3.2-1: Nuclear Safety Capability Assessment Method Review NE100-01 NRC Staffs Evaluation Section Title ONS Alignment Basis Section 3.5.1.1 Circuit Failure The licensee stated that the approach used Based on the information provided in Reference 10, the Types and Impact aligns with the intent of this requirement; All NRC staff finds that the licensee's approach has combinations of circuit failures except inter- adequately addressed the issue of grounding of armored cable hot shorts are considered and cable to preclude inter-cable shorts.
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| evaluated to determine if spurious (See SE Section 3.2.1).
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| component actuation can occur. Intercable hot shorts were not considered due to the use of armored cable at Oconee.
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| 3.5.1.2 Circuit Contacts Aligns The NRC staff finds the licensee's statement of and Operational alignment to the endorsed guidance acceptable.
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| Modes 3.5.1.3 Duration of Circuit Does not Align, but has previous NRC In a letter dated November 19, 2010 (Reference 52) the Failures approval. Previous design considerations licensee agreed to eliminate the" 10 minute free of fire did not assume spurious actuations or hot damage" assumptions and to perform an evaluation shorts due to a fire for the first 10 minutes of using NFPA 805 risk-informed processes. Specifically, the event. This was stated in the the licensee states that it will utilize a risk-informed referenced SER for the SSF. Other approach to evaluate conflicts that previously relied upon spurious operations beyond this assumption the 1O-minute prior approval. This will involve a were postulated to occur until mitigating thorough review of existing analyses to identify new actions are taken. VFDRs. Changes to the FPP, as a result of these VFDRs, will be resolved using the change evaluation process. Upon completion of this activity, all applicable FRE(s) will be updated and compliance will be demonstrated consistent with NFPA 805, Section 4.2.4.2. (See SE Section 3.2.1, Item 2, for more detail)
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| Based on the information provided, the NRC staff finds that the licensee's process for eliminating the "10-minute free of fire damage assumption" provides reasonable assurance that the safety objectives of NFPA 805 will be satisfied.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 176 Attachment B, Nuclear Safety Capability Assessment Method Review Table 3.2-1: Nuclear Safety Capability Assessment Method Review NE100-01 Section Title ONS Alignment Basis NRC Staffs Evaluation Section 3.5.1.4 Cable Failure Aligns The NRC staff finds the licensee's statement of Configurations alignment to the endorsed guidance acceptable.
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| 3.5.1.5 [A] Circuit Failure Risk Detailed alignment discussed below. See NRC staff evaluation for subsections below.
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| Assessment Guidance 3.5.1.5 [B] Cable Failure The licensee stated that the approach used See NRC staff's evaluation for SE Section 3.2.1.
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| Modes aligns with the intent of this requirement; Oconee has armored sheathing - cable-to Based on the information contained in the licensee's cable hot shorts are not postulated for letter dated September 13, 2010 (Reference 12), the armor-jacketed cables. NRC staff finds that the licensee's approach has adequately addressed the issue.
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| 3.5.1.5 [C] Likelihood of Does not Align The NRC staff was concerned that the MSO analysis Undesired Treatment of multiple spurious actuations is was limited to only SSD cables and components. In its Consequences being resolved through transition to NFPA October 14, 2010 letter (Reference 54), the licensee 805. states that the methodology assumed multiple fire-induced failures and multiple spurious actuations based on the cables and components present in the fire area of concern, and was not limited to SSD cables and components.
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| Based on the information provided in the LAR, as supplemented, the NRC staff finds that the licensee's basis for alignment to Section 3.5.1.5[C] of NEI 00-01 acceptable.
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| 3.5.2 Types of Circuit Detailed alignment discussed below. See NRC staff's evaluation for subsections below.
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| Failures OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 177 Attachment B, Nuclear Safety Capability Assessment Method Review Table 3.2-1: Nuclear Safety Capability Assessment Method Review NEI 00-01 Section Title ONS Alignment Basis NRC Staffs Evaluation Section 3.5.2.1 Circuit Failures Due Does not Align By letter dated September 13, 2010 (Reference 12), the to Open Circuits Open circuits are analyzed as shown on the licensee states, in part, that the assumption associated referenced figures from NEI 00-01 except with the secondary CT circuits is being removed to CT circuits. ensure that ONS has properly evaluated the effects of an open secondary CT as prescribed in NFPA 805, Section 2.4.2 and guided in NEI 00-01, Section 3.5.2.1.
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| Completion of this action is an implementation item (SE Section 2.9, Table 2.9-1, Item 40). (See SE Section 3.2.1)
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| The NRC staff finds the licensee's statement of aliQnment to the endorsed Quidance acceptable.
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| 3.5.2.2 Circuit Failures Due Detailed alignment discussed below. See NRC staff's evaluation for subsections below.
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| to Shorts to Ground fGenerall 3.5.2.2 Circuit Failures Due Aligns The NRC staff finds the licensee's statement of to Shorts to Ground alignment to the endorsed guidance acceptable.
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| [A, Grounded Circuits]
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| 3.5.2.2 Circuit Failures Due Aligns The NRC staff finds the licensee's statement of to Shorts to Ground alignment to the endorsed guidance acceptable.
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| [B, Ungrounded Circuits]
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| 3.5.2.3 Circuit Failures Due Detailed alignment discussed in subsequent See NRC staff's evaluation for subsections below.
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| to a Hot Short paragraphs below.
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| fGenerall OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 178 Attachment B, Nuclear Safety Capability Assessment Method Review Table 3.2-1: Nuclear Safety Capability Assessment Method Review NE100-01 Section Title ONS Alignment Basis NRC Staffs Evaluation Section 3.5.2.3 Circuit Failures Due The licensee stated that the approach used Based on the information contained in the licensee's to a Hot Short [A, aligns with the intent of this requirement letter dated September 13, 2010 (Reference 12), the Grounded Circuits] intra-cable conductor-to-conductor hot NRC staff finds that the licensee's approach has shorts are analyzed. The external hot short adequately addressed the issue. (See SE Section 3.2.1, is not considered credible at ONS due to the Item 1).
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| armored cable configuration.
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| 3.5.2.3 Circuit Failures Due Aligns The NRC staff finds the licensee's statement of to a Hot Short [B, alignment to the endorsed guidance acceptable.
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| Ungrounded Circuits]
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| 3.5.2.4 Circuit Failures Due DOES NOT ALIGN Based on the information provided in the LAR, as to Inadequate Proper coordination of common power supplemented, the NRC staff finds that the licensee's Circuit Coordination supplies for all circuits was an assumption approach has adequately addressed the issue of of the analysis. The licensee's existing inadequate breaker coordination at ONS and that, upon coordination study does not include all completion of the modifications and implementation SSEL related power supplies. items, ONS will align with the endorsed guidance for this attribute. (see SE Section 3.2.1, Item 1) 3.5.2.5 Circuit Failures Due DOES NOT ALIGN Based on the information provided in the LAR, as to Common The electrical circuit design for ONS is supplemented, the NRC staff finds that the licensee's Enclosure assumed to provide proper circuit protection approach has adequately addressed the Common Concerns in the form of circuit breakers, fuses and Enclosure Associated Circuit concern at ONS and that other devices that are designed to isolate ONS complies with the requirements of NFPA 805, cable faults before ignition temperature is Section 2.4.2.2.2.(b). (See SE Section 3.2.1, Item 4).
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| reached. Adequate electrical circuit protection and cable sizing were included as part of ONS plant electrical design.
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| However, as discussed in Section 3.3.3.3, the breaker coordination study for ONS does not include all SSD equipment and the analysis is required to be updated to ensure coordination exists. Due to the uncertainty OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 179 Attachment B, Nuclear Safety Capability Assessment Method Review Table 3.2-1: Nuclear Safety Capability Assessment Method Review NE100-01 Section Title ONS Alignment Basis NRC Staffs Evaluation Section of breaker coordination, ONS does not meet the intent of the Quidance.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 180 Attachment C, Fire Risk Evaluation Tables The licensee evaluated the technical adequacy of the portions of its internal events probabilistic risk assessment (PRA) model used to support development of the Fire PRA model by first performing a peer review of the ONS internal events PRA model. Subsequently, a contractor review and a licensee self-assessment were performed. Table 3.4-1, "Internal Events PRA Findings and Observations Resolution," summarizes the NRC staff's review of the licensee's resolution of the internal events PRA findings and observations (F&Os), which demonstrates the technical adequacy for this application.
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| Since ONS is an industry pilot for NFPA 805, consistent with RG 1.205, Revision 0, the NRC staff performed the review of the licensee's Fire PRA model to determine its technical adequacy because an industry peer review of the ONS Fire PRA model had not yet been performed. In addition, because a full-scope industry peer review of the ONS Fire PRA was not performed, the NRC staff reviewed a number of aspects of the Fire PRA model in detail. Table 3.4-2, "Fire PRA Findings and Observations Resolution," summarizes the NRC staff's review of the licensee's resolution of findings from the NRC staff's review (including both F&Os as well as supporting requirements (SRs) evaluated as less than Capability Category II without any specific F&O). Table 3.4-2 includes both the licensee's reported Topic, and the NRC staff's text from the NRC staff's F&Os. This evaluation establishes the technical adequacy of the ONS Fire PRA for this application.
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| The licensee provided detailed information regarding the correlations and fire models used to support implementation of NFPA 805 at ONS, as well as a cross reference between major sections of American Society for Testing and Materials (ASTM) guidance document ASTM E 1355-05a, "Standard Guide for Evaluating Predictive Capability of Deterministic Fire Models,"
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| and the associated correlations in terms of their applicability and validation. Table 3.4-3, "V&V Basis for Fire Modeling Correlations Used at ONS," identifies the empirical correlations and models used in the screening tool, the basis for acceptability with respect to verification and validation (V&V), and the NRC staffs evaluation of that basis.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 181 Attachment C, Fire Risk Evaluation Tables Table 3.4-1, Internal Events PRA F&O Resolution Facts and Observations (F&Os) Licensee's Disposition NRC Staff's Findings AS-B3 The AS [accident sequence] notebooks Accident sequence notebooks and The NRC staff does not accept document phenomenological conditions created by system model notebooks should identify that the F&O is resolved solely by the accident progression. For example, sump those environmental effects of the documentation. However, in temperature is examined when considering the initiating event and the impact on response to RAI 5-9i (Reference potential for LPI [low pressure injection] pump mitigation systems. 8) and 5-53 (Reference 9), the cavitation, and influences from ambient conditions licensee described the possible are examined for high energy line breaks. However, impact of the environment in the as noted for supporting requirements SY-B8 and SY three example accident scenarios B15, SSCs [structures, systems, and components] in the comment. Based on the that may be required to operate in conditions beyond disposition of the examples, the their environmental qualifications are not completely NRC staff finds that this deficiency identified and/or documented. Examples include: a) is not likely to cause the estimated LOCA [loss of coolant accident] inside containment transition risk decrease to become with failure of the Reactor Building Cooling System a risk increase.
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| would expose SG [steam generator] instrumentation to a harsh environment; b) Steam line breaks in the TB [turbine building] could expose equipment other than just the 4 kV switchgear and EFW control panel to an adverse environment; c) Clogging of the RBES
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| [reactor building emergency sump] is not discussed.
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| DA-B1 As documented in calculation OSC-8796, Revise the data calc. to segregate Based on the type of changes to mechanical components are grouped according to standby and operating component data. the PRA identified by the licensee, component type and system (HPI, EFW, SW, etc.). Segregate components by service the NRC staff finds that changes Electrical components are grouped based on condition to the extent supported by the are expected to be very small and component type and voltage level. However, since a data. This is a refinement to the thus this deficiency is not likely to generic database (NUREG/CR-6928) that separates equipment failure rates. However, cause the estimated transition risk standby and operating component failure rates is now since most components are grouped decrease to become a risk available, consider separating these groups. appropriately, the overall impact should increase.
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| be small.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 182 Attachment C, Fire Risk Evaluation Tables Table 3.4-1, Internal Events PRA F&O Resolution Facts and Observations (F&Os) Licensee's Disposition NRC Staff's Findings DA-D4 There is no evidence in the documentation Enhance the documentation to include The NRC staff does not accept that the specific checks required by this SR have a discussion of the specific checks that the F&O is resolved solely by been performed on the Bayesian-updated data to performed on the Bayesian-updated documentation. However, the ensure that the data is appropriate. However, a data, as required by this SR. NRC staff finds that a Bayesian verification of the proper operation of the software update is not likely to cause the within the expected data range (item d of the SR) estimated transition risk decrease was performed. A quick review of the current data to become a risk increase.
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| did not reveal any unusual or unexpected results, however.
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| DA-D6 Plant-specific CCF [common cause failure] Provide documentation in SAAG 637 of The NRC staff does not accept failure documentation (OSC-8797) was reviewed to the comparison of the component that the F&O is resolved solely by ensure that the generic CCF estimates were boundaries assumed for the generic documentation. However, the consistent with plant operating experience. However CCF estimates to those assumed in the licensee provided a review no evidence is provided to show that the component ONS PRA to ensure that these comparing the ONS CCF events boundaries used in the CCF generic estimates are boundaries are consistent. with generic electric system CCFs consistent with the component boundaries assumed in its response to RAI 5-ge for the PRA. (Reference 8). The review identified differences in the boundaries but concluded that plant operating experience is properly monitored for potential CCF events based on ONSs modeling. Therefore, the NRC staff finds that this deficiency is not likely to cause the estimated transition risk decrease to become a risk increase.
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| HR-A2 No documentation was found of a review of Enhance the HRA [human reliability Based on the assessment that procedures and practices to identify calibration analysis] to consider the potential for calibration (human error activities that if performed incorrectly can have an calibration errors. Based on preliminary probabilities (HEPs) are not OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 183 Attachment C, Fire Risk Evaluation Tables Table 3.4-1, Internal Events PRA F&O Resolution Facts and Observations (F&Os) Licensee's Disposition NRC Staff's Findings adverse impact on the automatic initiation of standby evaluations using the EPRI HRA expected to contribute safety equipment. The Oconee PRA assumes a calculator, calibration errors that result significantly to overall equipment Type A screening value for each standby PRA in failure of a single channel are unavailability, the NRC staff finds system train, as well as a common cause Type A expected to fall in the low 10-3 range. that this deficiency is not likely to screening value where identified to be appropriate. A Calibration errors that result in failure of cause the estimated transition risk review of procedures and practices would provide a multiple channels are expected to fall in decrease to become a risk worthwhile cross-check of this approach, to ensure the low 10-5 range. Relative to post- increase.
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| that no Type A events have been overlooked. For initiator HEPs], equipment random example, the review may identify a tank level failure rates and maintenance instrument calibration that would prevent the unavailability, calibration HEPs are not operation of redundant pump trains due to interlocks, expected to contribute significantly to which may not have been previously captured as part overall equipment unavailability.
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| of assigning common cause Type A screening values.
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| HR-A3 The Oconee PRA identified common cause Identify maintenance and calibration Based on the assessment that Type A HFEs [human failure events] that effect activities that could simultaneously calibration HEPs are not expected different trains of redundant systems where affect equipment in either different to contribute significantly to considered to be appropriate. No documentation was trains of a redundant system or diverse overall equipment unavailability, found of a review of procedures and practices to systems. Relative to post initiator the NRC staff finds that this identify maintenance and calibration activities that HEPs, equipment random failure rates deficiency is not likely to cause could render equipment unavailable if performed and maintenance unavailability, the estimated transition risk incorrectly. Such a review of procedures and calibration HEPs are not expected to decrease to become a risk practices would provide a worthwhile cross-check of contribute significantly to overall increase.
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| this approach, to ensure that no Type A events have equipment unavailability. See the been overlooked that simultaneously affect Expected Impact on Applications for equipment in either different trains of a redundant requirement HR-A2 above.
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| system or diverse systems.
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| HR-D6 The Type A HEPs are not identified to be Develop mean values for pre-initiator Based on the sensitivity study mean values and error factors are not provided in the HEPs. Pre-initiator HEPs are generally provided in the September 29, summary table of the HR rhuman reliability] notebook set to relatively high screening values. 2009 RAI response, the NRC staff OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE O~JLY SECURITY RELATED INFORMATION 184 Attachment C, Fire Risk Evaluation Tables Table 3.4-1, Internal Events PRA F&O Resolution Facts and Observations (F&Os) Licensee's Disposition NRC Staff's Findings (Table 2) Thus the suggested data refinement is finds that this deficiency is not not expected to have a significant likely to cause the estimated impact on this application. transition risk decrease to become a risk increase.
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| HR-G3 The human cognitive reliability [HCR] model Document in more detail the influence The NRC staff does not accept or the caused-based approach was used to quantify of performance shaping factors on that this F&O is resolved solely by cognition errors, and an abbreviated version of execution of human error probabilities. documentation. The NRC staff THERP to quantify execution errors. These finds that enhancing the human evaluation of cognitive errors explicitly included reliability analysis (HRA) models considerations of plant-specific and scenario-specific to address this deficiency is not performance shaping factors noted by this supporting likely to cause the estimated requirement. Execution errors were not computed if transition risk decrease to become cognition error probabilities dominated the human a risk increase.
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| error probabilities, but more detailed documentation would be desirable to support such conclusions. In many instances execution errors were assigned bounding screening values. These screening value assignments took into account a simplified set of scenario specific performance shaping factors. It is not clear that full consideration of performance shaping factors has been adequately taken into account.
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| HR-G4 The total time available to complete actions Enhance HRA documentation See NRC staff finding in HR-G3.
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| was generally obtained from thermal-hydraulic accordingly.
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| calculations for the accidents of interest (e.g., from MAAP analyses, hand calculations, or other sources).
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| In many cases, human interactions were assessed to not be time critical, and no estimate of the time available was needed. In other instances, adequate estimates on event timing were not available to OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 185 Attachment C, Fire Risk Evaluation Tables Table 3.4-1, Internal Events PRA F&O Resolution Facts and Observations (F&Os) Licensee's Disposition NRC Staff's Findings support HCR assessment and the caused-based decision tree approach was used. More detailed documentation to support these decisions and conclusions is desirable.
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| HR-G6 The PRA notebooks do not document a Document a review of the HFEs and See NRC staff finding in HR-G3.
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| review of the HFEs and their final HEPs relative to their final HEPs relative to each other to each other to check reasonableness given the confirm their reasonableness given the scenario context, plant history, procedures, scenario context, plant history, operational practices, and experience. procedures, operational practices, and experience.
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| HR-G9 The Oconee post-initiator HRA does not Develop mean values for post-initiator See NRC staff finding in HR-D6.
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| provide mean HEP values for use in the HEPs. The use of mean values for quantification of the PRA results. HEPs instead of lower probability median values can affect the PRA results. The fire analysis will include a sensitivity study to evaluate the use of different HEPs if the calculated risk is close to the threshold.
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| HR-H2 The HR notebook documents that some Develop more detailed documentation See NRC staff finding in HR-G3.
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| operator recovery actions are credited in the Oconee of operator cues, relevant performance PRA for which no procedures are available. More shaping factors, and availability of detailed documentation would be desirable of the sufficient manpower to perform the cues that alert the operator to perform the recovery action.
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| actions, relevant performance shaping factors, and availability of sufficient manpower to perform the action.
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| IF-B3 The current analysis identifies the capacities of Enhance the Internal Flood analysis to The licensee stated in its LAR that various tanks in the plant and identifies the flow rates address the potential for spray, jet accident scenarios initiated by for the defined initiating event definitions. The impingement, and pipe whip failures. internal flooding are not expected current analysis does not address spray events, does Additionally, document how these to impact the results of the fire OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE O~JLY SECURITY RELATED INFORMATION 186 Attachment C, Fire Risk Evaluation Tables Table 3.4-1, Internal Events PRA F&O Resolution Facts and Observations (F&Os) Licensee's Disposition NRC Staff's Findings not identify the potential volume of water that can be failures are included in the analysis. The NRC staff finds that lost through a pipe rupture (limited source versus an quantification. Internal flooding only flooding as a result of fire-unlimited source), and does not use system modeling issues do not impact fire risk. fighting could impact the fire pressures to calculate potential flow rate from a pipe results, and such flooding is rupture. Such information is necessary to satisfy the addressed independently of the requirements of the ASME PRA Standard. internal events PRA as described in Section 4.2.4 of the LAR.
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| Therefore, the NRC concludes that this deficiency will not cause the estimated transition risk decrease to become a risk increase.
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| IF-C2c For those flood areas addressed in the Given the expected increase in number See NRC staff finding in IF-B3.
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| current flooding analysis, equipment important to of flood areas needed to satisfy accident mitigation and the associated critical flood requirement IF-A1, additional heights are identified. However, given the expected equipment will need to be identified and increase in number of flood areas to be explicitly discussed in order to meet the addressed, additional equipment will need to be requirements of the ASME PRA identified and discussed in order to meet the Standard. The current flooding analysis requirements of the ASME PRA Standard. The does not discuss flood mitigative current flooding analysis does not discuss flood features and this will have to be mitigative features and this will have to be corrected corrected to satisfy the requirements of to satisfy the requirements of the ASME PRA the ASME PRA Standard. Internal Standard. flooding modeling issues do not impact fire risk.
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| IF-C3 The current flooding analysis identifies the The current flooding analysis identifies See NRC staff finding in IF-B3.
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| submergence failure height of the equipment the submergence failure height of the important to accident mitigation, but never addresses equipment important to accident the impact of spray. Spray as a failure mechanism mitigation, but never addresses the needs to be addressed in the analysis or a note impact of spray. Spray as a failure OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE O~JLY SECURITY RELATED INFORMATION 187 Attachment C, Fire Risk Evaluation Tables Table 3.4-1, Internal Events PRA F&O Resolution Facts and Observations (F&Os) Licensee's Disposition NRC Staff's Findings made explaining why it was omitted. mechanism needs to be addressed in the analysis or a note made explaining why it was omitted. Internal flooding modeling issues do not impact fire risk.
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| IF-C3b Discussion of propagation in the current Provide more analysis of flood See NRC staff finding in IF-B3.
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| flooding analysis is simply that water from any pipe propagation flowpaths. Address break in the AB will eventually drain down to the potential structural failure of doors or basement where it will begin to accumulate. The walls due to flooding loads and the mechanisms by which water will propagate are not potential for barrier unavailability.
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| discussed. Given the expected increase in number Internal flooding modeling issues do not of flood areas, additional propagation paths will likely impact fire risk.
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| be identified and the mechanisms by which water will propagate will have to be added to the discussion.
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| IF-E5 Some discussion of human errors is provided Develop and document any HEPs in a See NRC staff finding in IF-B3.
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| in the current flooding analysis, however the details manner comparable to that used in that impact the development of performance shaping developing HEPs for the internal events factors and the timelines that are essential to any PRA. Use of EPRI HRA method will human reliability analysis are not included. address this SR. Internal flooding modeling issues do not impact fire risk.
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| IF-E5a Some discussion of human errors is provided Develop and document any HEPs in a See NRC staff finding in IF-B3.
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| in the current flooding analysis; however the details manner comparable to that used in that impact the development of performance shaping developing HEPs for the internal events factors and the timelines that are essential to any PRA. Use of EPRI HRA method will human reliability analysis are not included. address this SR. Internal flooding modelinQ issues do not impact fire risk.
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| IF-E6b The current flooding analysis addresses the Address potential indirect effects. See NRC staff finding in IF-B3.
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| submergence failure heights of various equipment, Internal flooding modeling issues do not however no discussion of spray, jet impingement, or impact fire risk.
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| pipe whip failures is included. Additionally, due to the lack of quantification information it is unknown how OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 188 Attachment C, Fire Risk Evaluation Tables Table 3.4-1, Internal Events PRA F&O Resolution Facts and Observations (F&Os) Licensee's Disposition NRC Staff's Findings any of these failures were included in the quantification. A complete discussion of the quantification in needed to satisfy the requirements of the ASME Standard IF-F2 The flood analysis documentation does not Need to document how the analysis See NRC staff finding in IF-B3.
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| address all of the items identified in this requirement. addressed all of the items identified in this requirement. Internal flooding modeling issues do not impact fire risk.
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| LE-C6 The only mention of operator actions credited The only operator action expected to be See NRC staff finding in HR-D6.
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| in the Level 2 analysis was in Section 7.3 of SAAG important is RCS [reactor coolant
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| #818. However, the contribution to LERF [large early system] depressurization for small release frequency] from failure to perform this action LOCAs. However, the current analysis is assumed to be negligible compared to other LERF lacks a formal dependency analysis for sequences. this action. The result is expected to be insensitive to this impact given that the SGTR [steam generator tube rupture]
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| so dominates the result.
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| LE-F2 The assumptions in the model development Perform and document sensitivity In response to RAls 5-55 and 5 are presented in Section 5 of SAAG #818. However, studies to determine the impact of the 56 (Reference 9), the licensee their impact on the resulting LERF has not been assumptions and sources of model reported re-evaluating various evaluated. A parametric uncertainty analysis of the uncertainty on the LERF results. assumptions used in the LERF LERF results is presented in the Integration estimates. Based on this re Notebook. Sensitivity analyses and their insights evaluation of the more significant have not been documented. LERF pathways, the NRC staff concludes that a sensitivity study, including developing the appropriate documentation, would not cause the estimated transition risk decrease to become a risk increase. Therefore, the NRC OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED I~JFORMATION 189 Attachment C, Fire Risk Evaluation Tables Table 3.4-1, Internal Events PRA F&O Resolution Facts and Observations (F&Os) Licensee's Disposition NRC Staff's Findings staff finds that this deficiency is not likely to cause the estimated transition risk decrease to become a risk increase.
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| LE-F3 Tables 4.5.8-2 d and e of the ASME PRA Compare LERF results and Comparing LERF results with Standard include requirements such as documenting uncertainties to similar plants and similar plants is not likely to a review of the dominant contributors to LERF, include in the LERF documentation. significantly change the results.
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| comparing the overall LERF and LERF dominant The NRC staff finds that this contributors to similar plants, and evaluating the deficiency is not likely to cause overall LERF uncertainty intervals. The Integration the estimated transition risk Notebook presents the uncertainty band around the decrease to become a risk mean LERF, but these other requirements have not increase.
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| been performed.
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| LE-G3 SAAG #818 describes the development of the Evaluate the relative contribution of the Documenting the relative Oconee LERF models, and the Integration Notebook various contributors to the total LERF contribution of the PDSs is less presents the point estimate and uncertainty results. important than properly linking the The documentation does not assess the relative PDS frequencies with the contribution of the PDSs [plant damage states], etc. conditional LERF. The review to the LERF. included no findings on inappropriate linking. The NRC staff finds that developing the appropriate documentation would not cause the estimated transition risk decrease to become a risk increase.
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| LE-G4 The assumptions in the model development Perform and document sensitivity This observation is made on an are presented in Section 5 of SAAG #818. However, studies to determine the impact of the SR that identifies documentation their impact on the resulting LERF has not been assumptions and sources of model requirements instead of technical evaluated. The mean, 5th and 95 th percentile LERF uncertainty on the LERF results. requirements. The NRC staff uncertainty values are presented in the Integration finds that developing the OFFICIP.L USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 190 Attachment C, Fire Risk Evaluation Tables Table 3.4-1, Internal Events PRA F&O Resolution Facts and Observations (F&Os) Licensee's Disposition NRC Staff's Findings Notebook. Sensitivity analyses and their insights appropriate documentation would have not been documented. not cause the estimated transition risk decrease to become a risk increase.
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| LE-G5 No evaluation of the limitations in the LERF Include in the LERF documentation an The NRC staff does not accept analysis that could impact applications was assessment that identifies the that the F&O is resolved solely by presented in the documentation. limitations in the LERF analysis that documentation. However, in could impact applications. response to RAI 5-55 (Reference 9), the licensee reported that previously screened containment penetrations and Interfacing System Loss of Coolant Accident pathways (which are the most likely non-SGTR LERF pathways) were re-evaluated including the potential for fire induced multiple spurious operations. Therefore, the NRC staff finds that this deficiency is not likely to cause the estimated transition risk decrease to become a risk increase.
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| LE-G6 The ASME PRA Standard defines "large, Provide a discussion of the significant The NRC staff finds that this early" as: "the rapid, unmitigated release of airborne cut sets and sequences. observation is an observation that fission products from the containment to the a documentation SR is not met environment occurring before the effective (as opposed to a technical implementation of offsite emergency response and element). Significant fire LERF protective actions such that there is a potential for sequences are reported in the early health effects". SAAG #818 presents similar LAR. Based on the assessment wording in Section 7.1. that this is a documentation issue OFFICIAL USE O~JLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 191 Attachment C, Fire Risk Evaluation Tables Table 3.4-1, Internal Events PRA F&O Resolution Facts and Observations (F&Os) Licensee's Disposition NRC Staff's Findings only, the NRC staff finds that developing the appropriate documentation would not cause the estimated transition risk decrease to become a risk increase.
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| QU-D3 The Integration Notebook does not include a Perform and document a comparison of The NRC staff finds that a comparison of results between the ONS PRA and results between the ONS PRA and comparison of the internal events other similar plants. other similar plants. results with results at similar plants is not likely to cause the estimated transition risk decrease to become a risk increase.
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| QU-E4 Although general modeling assumptions are Perform and document a set of The NRC staff finds that the Fire provided in the PRA Modeling Guidelines (XSAA sensitivity cases to determine the PRA results are dominated by 115) and specific assumptions related to system impact of the assumptions and sources assumptions regarding the fire design, operation, and modeling are documented in of model uncertainty on the results. caused failures, and that the various PRA notebooks, the sensitivity of the Perform and document sensitivity sensitivity to assumptions not results to model uncertainties and assumptions has analyses to determine the impact of the directly related to fires is unlikely not been thoroughly evaluated. assumptions and sources of model to cause the large estimated risk uncertainty on the Fire PRA results. decrease to become a risk increase.
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| QU-F2 The model integration process and basic Expand the documentation of ONS This observation is made on an quantification results are documented in the PRA model results to address all SR that identifies documentation Integration Notebook. However, there is no required items. requirements instead of technical discussion of sensitivity analyses or of some of the requirements. Based on the other expected contents. Note that although the assessment that this is a Modeling Guidelines (XSAA-115) specify that the documentation issue only, the Integration Notebook is to include an overall NRC staff finds that developing description of Train A vs. Train B model asymmetries the appropriate documentation and a discussion of the impact on results, this would not cause the estimated OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 192 Attachment C, Fire Risk Evaluation Tables Table 3.4-1, Internal Events PRA F&O Resolution Facts and Observations (F&Os) Licensee's Disposition NRC Staff's Findings discussion is not included in SAAG 517. transition risk decrease to become a risk increase.
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| QU-F6 Section 12.0 of the Integration Notebook Document the required definitions. This observation is made on an notes that the CDF result is dominated by external SR that identifies documentation events tornado, fire and flood. There is no discussion requirements instead of technical of a specific quantitative definition for significant basic requirements. Based on the events, cutsets, accident sequences or functional assessment that this is a failures. documentation issue only, the NRC staff finds that developing the appropriate documentation would not cause the estimated transition risk decrease to become a risk increase.
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| SC-B5 Review of the PRA Modeling Guidelines Provide evidence that an acceptability In its response to RAI 5-9a and 5 (XSAA-115), the success criteria (SC) documentation review of the TH analyses is 9b (Reference 8) about AS-A9 provided (LPI with no LPSW SAAG 569, and HPI for performed. (since closed by the licensee), the Small and Medium LOCAs SAAG 213), and samples licensee described the update of of the AS (ATWS [anticipated transient without its thermal-hydraulic calculations scram], Transients - SAAG 671, LOCAs - SAAG 241) and provided a new table clearly and SY [systems] (SSF, EFW SAAG 259, and identifying the SC based on the HPI/CC) notebooks did not indicate that the TH update. The NRC finds that the
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| [thermal-hydraulic] results and results of other SC update is sufficient to close AS-A9 evaluations were consistently check for and that any slight modification reasonableness and acceptability. Only SAAG 569 that might arise by comparing with described a cross-disciplinary check of results where other studies will not cause the the Oconee LPI system engineer reviewed the RB estimated transition risk decrease sump temperatures and RB pressures calculated by to become a risk increase.
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| MAAP to confirm no NPSH [net positive suction head] problems would occur with operating the LPI pumps.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 193 Attachment C, Fire Risk Evaluation Tables Table 3.4-1, Internal Events PRA F&O Resolution Facts and Observations (F&Os) Licensee's Disposition NRC Staff's Findings SC-C1 Only a portion of the SC documentation was Improve the documentation on the TH See NRC staff finding in SC-B5.
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| provided for review (LPI with no LPSW SAAG 569, bases for all safety function SC for all and HPI for Small and Medium LOCAs SAAG 213). initiators.
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| SC information was also found in some of the AS notebooks reviewed (e.g., LOCA SAAG 241), which included additional SC runs in the appendices; other AS and SY notebooks included general statements of SC and associated references (not all of which were reviewed). In general, the SC documentation is somewhat scattered and dated; a consolidated SC document would facilitate uPQrades and review.
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| SC-C2 The PRA Modeling Guidelines (XSAA-115), Improve the documentation on the TH See NRC staff finding in SC-B5.
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| the SC documentation provided (LPI with no LPSW bases for all safety function SC for all SAAG 569, and HPI for Small and Medium LOCAs initiators.
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| SAAG 213), and samples of the AS (ATWS, Transients - SAAG 671, LOCAs - SAAG 241) and SY (SSF, EFW SAAG 259, and HPI/CC) notebooks all contain documentation of some aspect of the process to develop overall PRA SC. According to the HRA notebook, the total time available for performing operator actions was generally obtained from thermal-hydraulic calculations for the accidents of interest (e.g., from MAAP analyses, hand calculations, or other sources). Review of the HRA toolbox spreadsheets indicates that sometimes no reference for time available is provided, or sometimes only a reference number, which were not provided for review. The documentation reviewed is not consistently thorough in presenting input, methods, and results, and not all of the aspects listed above OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 194 Attachment C, Fire Risk Evaluation Tables Table 3.4-1, Internal Events PRA F&O Resolution Facts and Observations (F&Os) Licensee's Disposition NRC Staff's Findings are clearly documented.
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| SY-A14 Some failure modes are excluded in a Provide quantitative evaluations for In response to RAI 5-9g qualitative fashion rather than by using quantitative screening. (Reference 8), the licensee criteria. Examples: the containment isolation system reviewed the qualitative screening write-up notes that "electrical penetrations are not process and concluded that modeled due to their low probability of failure;" the changing to a quantitative High Pressure Service Water (HPSW) system write- approach is not expected to up uses a redundancy argument for excluding change the results of the inadvertent isolation of the main headers; the RCS qualitative screening. This was write-up states that "transfer failure events for motor- verified specifically for the HPSW operated valves (MOVs), manual valves and check and MOV examples identified in valves with 24 hr exposure times are not modeled the F&O. Based on the results of unless probabilistically significant with respect to the review of the qualitative
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| 'neighboring' basic events; the RPS write-up uses a screening process, the NRC staff diversity argument for excluding common mode finds that this deficiency is not failure of sensors or instrument strings that generate likely to cause the estimated risk a reactor scram signal. decrease to become a risk increase.
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| SY-A4 Some of the notebooks include a walk down Enhance the system documentation to The NRC staff does not accept checklist and system engineer review, others do not. include an up-to-date system walkdown that the F&O is resolved solely by checklist and system engineer review documentation. However, the for each system. Consider revising results from the Fire PRA used to workplace procedure XSAA-106 to support the transition are require that such documentation be dominated by scenarios where the revisited with each major PRA revision. fire-caused SSC failures are followed by a few random SSC failure events and/or human error events. The NRC staff finds that location-specific attributes, other than affects of fire damage, which OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 195 Attachment C, Fire Risk Evaluation Tables Table 3.4-1, Internal Events PRA F&O Resolution Facts and Observations (F&Os) Licensee's Disposition NRC Staff's Findings might not be fully reflected in the PRA will not cause the estimated transition risk decrease to become a risk increase.
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| SY-A8 Component boundaries are consistent with Enhance systems analysis The NRC staff does not accept those identified in the data analysis (OSC-8796). For documentation to discuss component that the F&O is resolved solely by example, breakers that supply a single load are boundaries. In response to RAI 5-9h documentation. In response to generally included within the boundary of the load. (Reference 8), the licensee clarified that RAI 5-9h (Reference 8), the (An exception is the 4 kV circuit breakers.) However, the component boundaries are licensee clarified that the boundaries are not discussed. Interlocks are consistent with source documents, such component boundaries are explicitly modeled in the system models. as NUREG/CR-6928, and that the consistent with source reviewer did not identify any technical documents, such as NUREG/CR issues with the assessment. 6928, and that the reviewer did not identify any technical issues with the assessment. Based on the results of the review and subsequent completion of the breaker co-ordination evaluation reported elsewhere in the SE and Attachment C, Table 3.4-2, F&O CS-B1-1, the NRC staff finds that any changes to the PRA that might be caused by adjusting component boundaries will not cause the estimated transition risk decrease to become a risk increase.
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| SY-B8 Duke's PRA modeling guidelines (XSAA-115) Per Duke's PRA modeling guidelines, The NRC staff does not accept include a walkdown checklist for documenting spatial ensure that a walkdown/system that the F&O is resolved solely by dependencies for modeled equipment such as engineer interview checklist is included documentation. However, the OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 196 Attachment C, Fire Risk Evaluation Tables Table 3.4-1, Internal Events PRA F&O Resolution Facts and Observations (F&Os) Licensee's Disposition NRC Staff's Findings inadvertent sprinkler operation, missiles, high in each system notebook. Based on spatial dependencies related to temperatures, flooding, fire, close proximity of the results of the system walkdown, fires clearly dominates the results equipment, and dependencies on HVAC that could summarize in the system write-up any of the Fire PRA and the potential significantly degrade the equipment. However, some possible spatial dependencies or impact of suppression activation of the system notebooks do not include this checklist. environmental hazards that may impact was evaluated separately as system operation. described in Section 4.2.4 in the LAR. The NRC staff finds that spatial dependencies not associated with the fire or fire suppression is not likely, if identified, to cause the estimated transition risk decrease to become a risk increase.
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| SY-B15 SSCs that may be required to operate in Cut set review during applications The NRC staff does not accept conditions beyond their environmental qualifications should address this. Suggest adding that the F&O is resolved solely by are not identified. Examples include: LOCA inside this guidance to workplace procedure documentation. Further, the NRC containment with failure of the RB cooling system XSAA-103. staff does not accept that would expose SG instrumentation to a harsh application cutset reviews resolve environment steam line breaks in the TB could this issue. However, in response expose equipment other than just the 4 kV to RAI 5-9 (Reference 8) and 5-53 switchgear and EFW control panel to an adverse (Reference 9), the licensee environment; clogging of the RBES is not discussed described the possible impact of and is not included in the system models. the environment in three example accident scenarios in the F&O for AS-B3. In the response, the licensee determined that the failure modes and/or loss of the identified equipment would have an insignificant impact on the PRA results. Based on the disposition OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 197 Attachment C, Fire Risk Evaluation Tables Table 3.4-1, Internal Events PRA F&O Resolution Facts and Observations (F&Os) Licensee's Disposition NRC Staff's Findings of the examples cited in the F&O for AS-B3, the NRC staff finds that this deficiency is not likely to cause the estimated transition risk decrease to become a risk increase.
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| SY-C2 The system notebooks contain much of the Enhance system model documentation The NRC staff finds that this F&O information listed in this SR. However, system model to comply with all ASME PRA Standard is made on an SR that identifies documentation should be enhanced to comply with all requirements. In response to RAI 5-9j documentation requirements ASME PRA Standard requirements. (Reference 8), the licensee noted that, instead of technical requirements.
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| prior to the advent of the ASME PRA Based on the assessment that Standard, the quality of the ONS PRA this is a documentation issue only, had been demonstrated to be adequate the NRC staff finds that for supporting applications and clarified developing the appropriate that the next version of the ONS PRA documentation would not cause model (Revision 4) will have the estimated transition risk significantly enhanced documentation decrease to become a risk to bring it into compliance with the increase.
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| ASME PRA Standard supporting requirements. The licensee also noted that the reviewer did not identify any technical issues with the assessment.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 198 Attachment C, Fire Risk Evaluation Tables Table 3.4-2, Fire Events PRA F&O Resolution Facts and Observations Licensee's Disposition NRC Staff's Findings F&O CS-B1-1 Breaker coordination study Breaker coordination issue not yet resolved [at In response to RAI 5-57, the licensee incomplete (self identified). the time of licensee's LAR submittal]; PIP 0-08 clarified that the draft breaker coordination 2444 has been generated to track completion. study confirmed that coordination exists at The top 50% risk contributing scenarios involve higher voltage levels and that therefore lack NRC staff F&O text: Plant personnel self- of coordination at lower voltage levels are loss of 4KV power and reliance on SSF identified some issues with over current mitigation which are not significantly impacted not expected to be significant.
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| coordination during the inspection. The plant is by additional failures due to improper breaker Subsequently, the licensee completed its working to resolve issues with molded-case breaker coordination study and therefore coordination.
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| circuit breaker instantaneous over current the licensee has appropriately resolved this tripping coordination.
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| F&O (Reference 12).
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| F&O CS-C4-1 Breaker coordination This item is the documentation component of The licensee completed its breaker documentation. the breaker coordination issue discussed in CS coordination study and therefore the B1-1. licensee has appropriately resolved this F&O (Reference 12).
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| NRC staff F&O text: No specific documentation of overcurrent coordination was provided during the inspection.
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| ES-B4 The NRC audit found that this supporting The licensee evaluated this supporting Based on the licensee statement that there requirement was met at capability category I requirement as meeting capability category III was no limit to the number of spurious based on identifying the 4 Main Coolant Pump since no limit was placed on the number of fire- operations considered in the ISLOCA seal return isolation valves as additions to the induced spurious operations in consideration of analysis, the NRC staff finds that this F&O equipment list based on fire-induced spurious potentiallSLOCA or containment bypass has been appropriately resolved by the operations leading to an ISLOCA [interfacing scenarios. licensee for this application.
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| systems loss-of-coolant accident]..
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 199 Attachment C, Fire Risk Evaluation Tables Table 3.4-2, Fire Events PRA F&O Resolution Facts and Observations Licensee's Disposition NRC Staff's Findings CS-A10 The NRC audit found that this The licensee evaluated this supporting In response to RAI 5-29, the licensee further supporting requirement was met at capability requirement as meeting capability category III clarified that (1) an evaluation of "Y1" and category I based on cable location information since cable location information is available in "Y2" components determined that these not being available in the ARTRAK database for ARTRAK at the fire area, fire zone, and raceway components more closely met capability "Y2" components and cable routing information level. While compliance is on an area basis, the category III than I and (2) the impact of the not being available for "Y3" components. Fire PRA scenarios included identification of use of the "Y3" components on the Fire targets (including raceways), where applicable. PRA quantification results was evaluated in It is noted that credit by exclusion was only a sensitivity analysis in the ONS NFPA 805 applied to low risk components (designated as Fire PRA Application Calculation. Based on
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| 'Y3') for which cable routing information was not identifying cable locations at the raceway assembled. level, with the exception of low risk components, and the evaluation of low risk components in a sensitivity analysis, the NRC staff finds that this F&O has been resolved by the licensee for this application.
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| F&O FQ-A2-1 Initiating events not defined for all Loss of Condenser Vacuum was the default Based on the identification of corrective fire scenarios. initiator for the Fire PRA but some scenarios actions being completed, the NRC staff assumed a different initiator as discussed in the finds that this F&O has been appropriately Fire PRA Model Development Report. The resolved by the licensee for this application NRC staff F&O text: Fire scenario frequencies applied initiator has been added to information because any additional changes are not are documented and reported based on plant provided in the quantification results summary likely to cause the estimated transition risk analysis unit by unit and/or ignition source. The table (Fire PRA Summary Report, Appendix A). decrease to become a risk increase.
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| corresponding failed components are identified, but there is no identification of which initiating event is initiated by the fire.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 200 Attachment C, Fire Risk Evaluation Tables Table 3.4-2, Fire Events PRA F&O Resolution Facts and Observations Licensee's Disposition NRC Staff's Findings F&O FQ-B1-1 Demonstrate convergence for the Reference IEPRA-2 below for closure of FPIE Based on the licensee's assessment that selected truncation limit (see IEPRA-2 for truncation issue. The Fire PRA solves for there is no truncation issue, the NRC staff related FPIE [full power internal event] issue). conditional core damage probability (CCDP) finds that this F&O has been appropriately (prior to application of ignition frequency) at one resolved by the licensee for this application order of magnitude greater than the FPIE. because any additional changes are not NRC staff F&O text: Internal events QU-B3 is Since a typical scenario frequency is typically likely to cause the estimated transition risk incorporated by reference and is not met: An much less than 0.1, there is not a truncation decrease to become a risk increase.
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| iterative demonstration of convergence versus (convergence) issue for Fire PRA.
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| truncation level has not yet been performed.
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| F&O FQ-C1-1 Use of nominal HEP values may To address the retention of additional cutsets, In response to RAI 5-58 (Reference 9), the result in loss of cutsets before application of the HEP values were set to 0.1 for the initial applicant indicated that they use the larger recoveries/multipliers. solve prior to the application of recoveries (as of 0.1 or the nominal HEP for initial cutset described in the Fire PRA Model Development generation, retaining hundreds of joint Report). human errors in the cutsets. While using NRC staff F&O text: Using nominal HEPs 1.0 as the screening HEP would increase during quantification can result in cutsets being the number of retained cutsets, the licensee truncated. Rule Based Recovery will not correct did not expect this would introduce new this, since the cutsets are not present in the important combinations, but merely burden results.
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| the quantification effort unduly. Based on the licensee's conclusion that no new important human error combinations are expected, the NRC finds that reducing the screening HEP will not cause the estimated transition risk decrease to become a risk increase.
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| OFFICIAL USE ONLY SECURITY RELJ\TED INFORMATION 201 Attachment C, Fire Risk Evaluation Tables Table 3.4-2, Fire Events PRA F&O Resolution Facts and Observations Licensee's Disposition NRC Staff's Findings F&O FQ-E1-1 Identification of significant Risk insights from the risk significant fire Based on the identification of corrective contributors. initiating events with identification of significant actions being completed, the NRC staff contributors have been included in the Fire PRA finds that this F&O has been appropriately Application Calculation. resolved by the licensee for this application.
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| NRC staff F&O text: The chosen method (see PRM-FQ-A) does not produce different significant contributor categories to support results review. As of audit review, limited review of available results had been performed.
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| F&O FQ-F1-1 Improve LERF documentation. Documentation concerns were largely confined Based on the identification of corrective to LERF. Accordingly, the insights section in actions being completed, the NRC staff the Fire PRA Summary Report was expanded to finds that this F&O has been appropriately NRC staff F&O text: It is not possible to see the address LERF. Inconsistencies between LERF resolved by the licensee for this application initiating event assigned to each scenario unless and CDF have been reconciled. Additionally, because any additional changes are not one looks in the cutset output files and deduces the risk insights section of the Fire PRA likely to cause the estimated transition risk the initiating event based on the failed Application Calculation and the insights section decrease to become a risk increase.
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| equipment. Method used by the licensee can in the Fire PRA Summary Report have been produce the correct numerical results without expanded to address LERF. Also, the applied meeting the standard.
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| initiator has been added to information provided in the quantification results summary as discussed in response to FQ-A2-1.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 202 Attachment C, Fire Risk Evaluation Tables Table 3.4-2, Fire Events PRA F&O Resolution Facts and Observations Licensee's Disposition NRC Staff's Findings F&O FSS-A5-1 Horizontal propagation for PVC jacketing impact on horizontal fire spread The NRC staff finds that the licensee's cables with PVC jackets (self identified). is discussed in the Fire Scenario Report. While evaluation that targets associated with armored cables are generally considered to be horizontal fire propagation along armored noncombustible (refer to NUREG/CR-6850 cables having PVC coating has not been NRC staff F&O text: Horizontal propagation Section R.4.1.4), the armored cables at ONS completely addressed. However, it is outside zone of influence (ZOI) not done yet, as have a PVC jacketing. The justification unlikely that this deficiency would cause the the effects of PVC jacket (typically TP concluded that the PVC coating will not sustain estimated transition risk decrease to
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| [thermoplastic] on horizontal spread is an open propagation of fire along the armored cable for a become a risk increase for this application.
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| item self identified by the licensee.
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| significant distance; any horizontal fire propagation along cable trays is adequately captured within the target set of each scenario involving overhead tray failures.
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| FSS-C2 The NRC audit found that this The licensee dispositions capability category I Based on the use of conservative supporting requirement was met at capability acceptable for the application given that the assumptions for HRR for the initial ZOI and category I based on the peak heat release rate results are conservative and that no changes to the evaluation of possible failure of manual being assumed at t=O when establishing the conclusions are anticipated if time dependent suppression, the NRC staff finds that this HGL threshold and for development of fire growth profiles are assumed given the observation has been investigated and scenarios for individual ignition sources. conservatism inherent in the NUREG/CR-6850 addressed for this application appropriately maximum heat release rates and available fire because any additional changes are not growth profiles relative to the ignition likely to cause the estimated transition risk frequencies. decrease to become a risk increase.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 203 Attachment C, Fire Risk Evaluation Tables Table 3.4-2, Fire Events PRA F&O Resolution Facts and Observations Licensee's Disposition NRC Staff's Findings F&O FSS-C5-1 Potential for PVC pooling may PVC pooling is addressed in the Fire Scenario The NRC staff finds that the licensee's impact assumed damage threshold. Report. The justification centers on the evaluation that PVC pooling within cable expectation that cables would tend to have the trays is not expected to occur has not been PVC jacket melt and flow away creating voids completely addressed, but any additional NRC staff F&O text: For those cases with for the flow of melting materials from other changes are not likely to cause the potentially pooling thermoplastic (TP), TP cables. In addition, the ridges that are estimated transition risk decrease to characteristics for failure should be attributed to characteristic of the armor jacketing provide become a risk increase.
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| ONS cable.
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| additional free space for the flow of material. As such, it is not expected that pooling of PVC within cable trays is likely to occur even if multiple layers of armored cables exist.
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| F&O FSS-D5-1 Justify use of 75% HRR for Justification in the Fire Scenario Report is Based on the information provided by the transient fires. expanded. Use of the 75% HRR for transients licensee, the NRC staff finds that this F&O is considered realistic treatment (more has been appropriately resolved by the characteristic of actual transient fire scenarios licensee for this application because any NRC staff F&O text: The licensee has provided identified in the fire events database) and additional changes are not likely to cause a weak basis for applying only the 75% fire HRR appropriate for PRA application. Use of a trash the estimated transition risk decrease to for transient combustibles, excluding the larger bag as the transient fuel package provides a become a risk increase.
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| HRR.
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| bounding characterization of the behavior of observed transient fire ignition sources, but ignores the ignition element. Consequently, while administrative controls are factored into the development of transient ignition frequencies for each compartment, the actual transient combustible loading that may be allowed or present does not directly impact the numerical results of the Fire PRA.
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| F&O FSS-D6-1 Justify fire brigade response The licensee developed support for fire brigade The topic identified by the licensee does not time with respect to formation of HGL. response time of 20 minutes from the review of completely address the NRC staff's finding actual fire brigade drill performance, which is that a manual suppression failure probability included in the Fire Scenario Report. should be developed for the fire brigade NRC staff F&O text: A probability should be Specifically, the applicant reviewed 100 plant intervening to prevent damage from the OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE O~JLY SECURITY RELATED INFORMATION 204 Attachment C, Fire Risk Evaluation Tables Table 3.4-2, Fire Events PRA F&O Resolution Facts and Observations Licensee's Disposition NRC Staff's Findings developed for manual suppression for the fire drills, in which the time exceeded 20 minutes HGL, or upward propagation through brigade intervening to prevent damage from the only once, and only slightly in that one instance. successive cable trays. In response to RAI HGL. In lieu of a sensitivity analysis, the applicant 5-27 (Reference 9), the license summarized bounded the effect of delayed brigade response the site experience used to support the 20 by assuming the more conservative minute response time. Based on the phenomenology for damage due to lOI effects licensee's documentation supporting its vs. HGL formation, showing that the assumed brigade response time the NRC compartments screened for the former and, staff finds that a high likelihood of response therefore, bounded the latter. The applicant within 20 minutes has been established.
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| further cited an inherent conservatism that, However, the assumptions used by the while the brigade might not initiate suppression licensee in the lOI determination, while activities within 20 minutes, credit has not been conservative in the HRR attributed to the taken for the brigade taking the simple action to ignition source, are non-conservative in that open a door to prevent HGL formation in a the fire never propagates by igniting shorter time. additional combustible material (and thereby increasing the HRR) beyond the original ignition source.. The applications of conservative initial lOI with the non-conservative assumption that the fire never propagates to combustibles beyond the original ignition source yields an indeterminate result. However, it is unlikely that deficiency would cause the estimated transition risk decrease for this application.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 205 Attachment C, Fire Risk Evaluation Tables Table 3.4-2, Fire Events PRA F&O Resolution Facts and Observations Licensee's Disposition NRC Staff's Findings FSS-D9 The NRC audit found that this The licensee evaluated this supporting Based on the licensee evaluating the supporting requirement was met at capability requirement to meet capability category 11/111 potential for smoke damage to Fire PRA category I based on smoke damage to Fire PRA based on the potential for smoke damage to equipment when identifying targets, the equipment not being considered. equipment not already failed by fire affects NRC staff finds that this F&O has been being added to the Fire Scenario Report. appropriately resolved by the licensee for this application because any additional changes are not likely to cause the estimated transition risk decrease to become a risk increase.
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| FSS-E3 The NRC audit found that this The licensee dispositions capability category I Based on the use of fire modeling supporting requirement was met at capability acceptable for the application given that parameters from NUREG/CR-6850, the category I based on the uncertainty analyses parameters for modeling fire scenarios (such as NRC staff finds that capability category I is presented in the "Generic Fire Modeling heat release rates and severity factors) are acceptable for this application because any Treatments" document. taken from NUREG/CR-6850, which is the additional changes are not likely to cause consensus method for Fire PRA development. the estimated transition risk decrease to become a risk increase.
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| FSS-F2 The NRC audit found that this The licensee dispositions this supporting The NRC staff finds that the justification that supporting requirement was met at capability requirement as not applicable since no FSS-F2 is not applicable to ONS is category I based on no criterion being scenarios were selected for quantification acceptable.
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| established or justified for structural collapse (structural steel damage required no further due to exposure of structural steel to a fire. quantitative treatment).
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| FSS-F3 The NRC audit found that this The licensee dispositions this supporting The NRC staff finds that the justification that supporting requirement was met at capability requirement as not applicable since no FSS-F3 is not applicable to ONS is category I based on the qualitative assessment scenarios were selected for quantification acceptable.
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| that the MFW oil fire scenario bounds the CCDP (structural steel damage required no further of a structural collapse of the TB. quantitative treatment).
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 206 Attachment C, Fire Risk Evaluation Tables Table 3.4-2, Fire Events PRA F&O Resolution Facts and Observations Licensee's Disposition NRC Staff's Findings F&O FSS-G1-1 MUlti-compartment analysis Addressed via multi-compartment screening Based on the licensee's multi-compartment incomplete. analysis added as Attachment D to the Fire evaluation to identify targets in adjacent Scenario Report. compartments that are within the zone of influence for a given fire scenario and NRC staff F&O text: MUlti-compartment including these targets in the set of analysis does not include a range of potential equipment that are damaged by the fire, the multi-compartment fire scenarios.
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| NRC staff finds that this F&O has been appropriately resolved by the licensee for this application because any additional changes are not likely to cause the estimated transition risk decrease to become a risk increase.
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| F&O FSS-G2-1 MUlti-compartment analysis Screening criteria added to multi-compartment See NRC staff finding on F&O FSS-G1-1.
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| screening criteria not defined. discussion in the Fire Scenario Report.
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| NRC staff F&O text: No screening criteria for multi-compartment fires have been defined.
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| F&O FSS-G3-1 Multi-compartment analysis Addressed via multi-compartment screening See NRC staff finding on F&O FSS-G1-1.
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| screening incomplete; no MCA scenarios analysis added as Attachment D to the Fire defined for quantification. Scenario Report. The screening criteria were applied to all analyzed compartments with the end result being that all compartments (physical NRC staff F&O text: The analysis has not analysis units) screened.
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| screened potential multi-compartment combinations of interest nor defined any multi-compartment fire scenarios beyond those that are inherently captured in the treatment of fire scenarios for the TB fire zones (see PP).
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 207 Attachment C, Fire Risk Evaluation Tables Table 3.4*2, Fire Events PRA F&O Resolution Facts and Observations Licensee's Disposition NRC Staff's Findings F&O FSS-G4-1 Multi-compartment analysis did Potential for fire barrier failure addressed via See NRC staff finding on F&O FSS-G1-1.
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| not consider potential for barrier failure. multi-compartment screening analysis added as Attachment 0 to the Fire Scenario Report.
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| NRC staff F&O text: Multi-compartment analysis is incomplete and has not included an assessment of credits given to passive fire barrier features.
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| F&O FSS-G5-1 MUlti-compartment analysis did No active fire barriers have been credited in the Based on the licensee's statement that no not assess active fire barriers. Fire PRA for limiting the zone of influence. The active fire barriers are being credited in the active fire barrier between BH12 and CT4 is Fire PRA, the NRC staff finds that this F&O only credited for the deterministic fire area has been appropriately resolved by the NRC staff F&O text: Active fire barriers have boundary definition. licensee for this application.
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| been credited in partitioning but not assessed per this SR.
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| F&O FSS-G6-1 Assessment of multi- Addressed via mUlti-compartment screening See NRC staff finding on F&O FSS-G1-1.
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| compartment analysis scenarios relative to fire analysis added as Attachment 0 to the Fire risk not performed. Scenario Report. No additional scenarios were identified for quantification.
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| NRC staff F&O text: No assessment of a range of potential multi-compartment scenarios has been provided.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 208 Attachment C, Fire Risk Evaluation Tables Table 3.4-2, Fire Events PRA F&O Resolution Facts and Observations Licensee's Disposition NRC Staff's Findings F&O FSS-H2-1 document resolution of PVC Addressed in Fire Scenario report: while the Based on the identification of corrective pooling issue. PVC jacket is thermoplastic, the cable insulation actions being completed, the NRC staff within the armor is consistent with thermoset concludes that this deficiency has not been (flame retardant cross-linked polyethylene). completely addressed, but any additional NRC staff F&O text: Treatment of PVC jacket changes are not likely to cause the on cable failure is not addressed and is an open estimated transition risk decrease to item. The cable jacket affects Oconee See FSS-CS-1.
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| become a risk increase.
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| conclusion that TS failure criteria should be used. Oconee is collecting information regarding the PVC jacket to establish the nature of the PVC; however, PVC is typically thermoplastic material.
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| FSS-H6 The NRC audit found that this The licensee evaluated this supporting Based on the Bayesian updating of fire supporting requirement was met at capability requirement to meet capability category 11/111 frequencies and the consideration of ONS-category I based on the lack of documentation based on the use of conservative scoping fire specific fire brigade response times, the supporting the contention that using the area modeling criteria from NUREG/CR-68S0 and NRC staff concludes that this deficiency has ratio is adequate and justifying the use of the that, other than the Bayesian update of fire been appropriately resolved by the licensee 7S th percentile of the NUREG/CR-68S0 heat frequencies and consideration of actual fire for this application because any changes release rate for transient combustibles. brigade response times, no plant specific are not likely to cause the estimated updates were applied. transition risk decrease to become a risk increase.
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| F&O FSS-H8-1 Multi-compartment analysis Addressed via supplemental discussion and Based on the identification of corrective documentation incomplete. multi compartment screening analysis added as actions being completed, the NRC staff Attachment D to Fire Scenario Report. concludes that this deficiency has been appropriately resolved by the licensee for NRC staff F&O text: The multi-compartment this application.
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| analysis remains incomplete (see FSS-F and its SRs); hence, documentation is also incomplete.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 209 Attachment C, Fire Risk Evaluation Tables Table 3.4-2, Fire Events PRA F&O Resolution Facts and Observations Licensee's Disposition NRC Staff's Findings F&O FSS-H9-1 Document justification for fire Support for fire brigade response time of 20 See NRC staff finding on F&O FSS-D6-1.
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| brigade response time. minutes based on review of actual fire brigade drill performance has been added to Fire Scenario Report. Specifically, the applicant There NRC staff F&O text: are no uncertainties reviewed 100 plant drills, in which the time listed for manual fire brigade suppression which exceeded 20 minutes only once, and only limits the development of the hot gas layer as slightly in that one instance. In lieu of a given in the document Oconee FPRA 031408 sensitivity analysis, the applicant bounded the Tasks 8 and 11 Scenario Development effect of delayed brigade response by assuming Attachment A, Scenario Summary Report.
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| the more conservative phenomenology for damage due to ZOI effects vs. HGL formation, showing that the compartments screened for the former and, therefore, bounded the latter. The applicant further cited an inherent conservatism that, while the brigade might not initiate suppression activities within 20 minutes, credit has not been taken for the brigade taking the simple action to open a door to prevent HGL formation in a shorter time.
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| F&O HR-G7-1 (Internal Events PRA) Corrective Action 2 of PIP 0-08-2915 Based on the identification of corrective Dependencies should be reviewed with respect dispositions issue as minimally conservative actions being completed, the NRC staff to timing. (compared to other sources of conservatism in concludes that this deficiency has been Fire PRA) based on the degree of dependency appropriately resolved by the licensee for between actions with different available this application because any additional response times. changes are not likely to cause the estimated transition risk decrease to become a risk increase. The internal events HRA methodology is addressed in the NRC staff finding HR-G3.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 210 Attachment C, Fire Risk Evaluation Tables Table 3.4-2, Fire Events PRA F&O Resolution Facts and Observations Licensee's Disposition NRC Staff's Findings F&O HRA-A1-1 SR considered met but Discussion relative to reliance on Based on the identification of corrective documentation that SSD actions carried over EOP[emergency operating procedure], actions being completed, the NRC staff from FPIE model remain valid for Fire PRA was abnormal operating procedures (AOPs), and concludes that this deficiency has been not provided. alarm response procedures given a fire has appropriately resolved by the licensee for been added to Fire PRA Model Development this application because any additional report (see HRA-E1-1). changes are not likely to cause the NRC staff F&O text: No documentation that for estimated transition risk decrease to each fire scenario, for each SSD action carried become a risk increase.
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| over from the Internal Events PRA, each action remains valid in the context of the Fire PRA.
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| F&O HRA-B2-1 HRA documentation for No impact on quantification; Corrective Action 3 See NRC staff's finding on F&O HRA-A1-1.
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| CASWHPIDHE and CEFOASWDHE insufficient. of PIP 0-08-2915 indicates that documentation deficiency will be addressed with issuance of future Revision 4 of the ONS PRA Model.
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| NRC staff F&O text: For events CASWHPIDHE and CEFOASWDHE the definition of the HFEs is not as detailed as that for the other HFEs.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 211 Attachment C, Fire Risk Evaluation Tables Table 3.4-2, Fire Events PRA F&O Resolution Facts and Observations Licensee's Disposition NRC Staff's Findings F&O HRA-C1-1 Need to consider relative timing Risk significant HFE's revisited with respect to See NRC staff's finding on F&O HRA-A1-1.
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| of HFE in fire scenario; time from cue versus timing of cues; documentation provided in Fire time from fire. PRA Model Development Report.
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| NRC staff F&O text: The approach to the quantification of the HEPs is to revise the internal events HEPs using a set of rules revising the HEPs based on, among other things, allowable action time. The basis, or the set of assumptions upon which this set of rules is based is not provided. This does not seem to have been applied correctly.
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| F&O HRA-G7-1.
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| NRC staff F&O text: In reviewing the documentation for ZHFC-2-058 there is evidence that there is a lack of appreciation of the relative timing of events. The comment in the documentation on relative timing focuses on cognitive response time (2 minutes and 15 minutes for the two events (NSFORCMDHE and CASWHPIDHE) respectively. However, these two events are separated in time by a significant time, the first event being required at 30 minutes, the second at four hours respectively.
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| The dependency evaluations should be reviewed carefully for the internal events model and the fire model.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 212 Attachment C, Fire Risk Evaluation Tables Table 3.4-2, Fire Events PRA F&O Resolution Facts and Observations Licensee's Disposition NRC Staff's Findings F&O HRA-C1-2 Post-initiator HEP Top 6 risk significant operator actions in the Fire See NRC staff's finding on F&O HRA-A1-1.
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| quantifications need to be checked for PRA were checked for consistent application of consistency. criteria (Corrective Action 4 of PIP 0-08-2915).
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| The following operator actions were reviewed:
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| TTRHPITDHE was compared to NRC staff F&O text: (internal events SR CHPHPMUDHE, HHPHPRODHE was compared referenced by HRA-C1) There is no evidence to LLPLPRODHE. And finally, FEFEFW2DHE that the fire related post-initiator HEP was compared to FEFEFW1 DHE. The basic quantifications have been checked for events in each pair were similar in consistency.
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| characteristics and they were accordingly mapped to the same HEP adjustment case (i.e.
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| Case 3 or 4 etc.).
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| F&O HRA-E1-1 Address how alarm response Discussion relative to reliance on EOPs, AOPs, See NRC staff's finding on F&O HRA-A1-1.
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| and EOP/AOPs are followed given a fire. and alarm response procedures given a fire has been added to Fire PRA Model Development report to support the conclusion that credited NRC staff F&O text: There is no documentation Fire PRA actions are consistent with the to describe the procedures and their use during expected plant response to a fire event a fire scenario. There is no documentation of including the decision to man the SSF.
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| the assumptions underlying the screening approach. There is no justification that the timing associated with the analyzed HFEs is appropriate for the accident scenarios.
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| F&O IEPRA-1 Resolve issues from gap Appendix D to PRA Quality Self Assessment Based on the NRC staff's conclusions about assessment of the ONS PRA Revision 3a. addresses ONS PRA technical adequacy for the technical adequacy of the ONS PRA NFPA 805. Open SRs with potential impact on Revision 3a discussed in this SE, the NRC quantification of delta risk for change concludes that resolving the remaining NRC staff F&O text: Document resolution of the evaluations have been addressed in sensitivity issues from the gap assessment on the issues identified in the Maracor review; justify analysis within NFPA 805 Fire PRA Application internal event PRA will not cause the any exceptions.
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| Calculation. estimated transition risk decrease to become a risk increase.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 213 Attachment C, Fire Risk Evaluation Tables Table 3.4-2, Fire Events PRA F&O Resolution Facts and Observations Licensee's Disposition NRC Staff's Findings F&O IEPRA-2 Demonstrate convergence for OSC-8863 demonstrates that the ONS PRA The NRC staff F&O addressed many selected truncation limit (FPIE issue). CDF results converge at 1E-09. updates credited in the "Oconee Responses," not simply a convergence issue as identified in the ONS Topic NRC staff F&O text: Need to perform updates description. For disposition of the part of credited in "Oconee Responses" to the 2006 this F&O related to general incorporation of Maracor review. See also F&O PRM-B1-1.
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| F&Os related to the Maracor review of the internal events PRA see the NRC staffs conclusions in Attachment C, Table 3.4-1 in this SE. The NRC staff finds that the issue of convergence has been appropriately resolved for this application.
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| F&O IGN-A5-1 Use of reactor year/critical year. To be addressed when NUREG/CR-6850 is Based on the use of acceptable ignition updated with the correct numbers based on frequency data from NUREG/CR-6850, the reactor-year basis; impact is expected to be NRC staff concludes that this deficiency has NRC staff F&O text: Need to update ignition insignificant. Note that Interim EPRI Report been appropriately resolved by the licensee frequency data once NUREG/CR-6850 is 1019189, which was not used at ONS, would for this application because any additional updated with the correct numbers based on lower ignition frequencies for most bins. Also changes are not likely to cause the reactor-year basis.
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| Bayesian update to 'lower' ignition frequencies estimated transition risk decrease to was not applied at ONS (only 1 bin was become a risk increase.
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| increased). Both measures, if applied, would offset expected increase.
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| F&O MUD-B4-1 Procedure lacks reference to Per Corrective Action 5 of PIP 0-08-2915, PRA Based on the identification of corrective PRA combined standard (draft). Workplace Procedure XSAA-1 06 was revised to actions being completed, the NRC staff reference the Combined PRA standard concludes that this deficiency has been appropriately resolved by the licensee for NRC staff F&O text: Fire standard needs to be this application.
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| referenced in XSAA-106 and the Fire PRA model should be explicitly in the scope of the procedure.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 214 Attachment C, Fire Risk Evaluation Tables Table 3.4-2, Fire Events PRA F&O Resolution Facts and Observations Licensee's Disposition NRC Staff's Findings F&O MUD-E1-1 Qualify the FRANC computer Per Corrective Action 6 of PI P 0-08-2915, the Based on the identification of corrective code for use on Fire PRA. FRANC computer code was qualified as actions being completed, the NRC staff documented in SDQA-30271-NGO. concludes that this deficiency has been appropriately resolved by the licensee for NRC staff F&O text: The FRANC computer this application.
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| code and corresponding Microsoft Access databases have not been evaluated and documented at any software and data quality assurance (SDQA) classification per NSD-800.
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| F&O PP-B2-1 Justification for credit of nonrated Partially addressed via multi-compartment See NRC staff's finding on F&O FSS-G1-1.
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| partition boundaries insufficient. analysis; failure to meet SR poses no adverse impact on the analysis quality or completeness.
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| Deviation from "enclosed boundary" definition NRC staff F&O text: The FPRA credited applied to compartment frequency calculation partitioning elements that lacked fire resistance which has no impact on overall CDF/LERF rating.
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| results. Use of zone of influence for defining extent of fire scenario regardless of location of zone boundary ensures that scenario impacts are accurate.
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| F&O PP-B3-1 Use of open fire zone boundaries Partially addressed via multi-compartment See NRC staff's finding on F&O FSS-G1-1.
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| implies credit for spatial separation. analysis; failure to meet SR poses no adverse impact on the analysis quality or completeness (see disposition for PP-B2-1).
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| NRC staff F&O text: Some TB Fire Compartments have boundaries that do not correspond to a physical wall, and thus have no fire rating. The use of these boundaries implies crediting spatial separation; therefore the standard is not met at Category 1.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 215 Attachment C, Fire Risk Evaluation Tables Table 3.4-2, Fire Events PRA F&O Resolution Facts and Observations Licensee's Disposition NRC Staff's Findings PP-B5 The NRC audit found that this supporting The licensee dispositions capability category I See NRC staff's finding on F&O FSS-G1-1.
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| requirement was met at capability category I acceptable for the application given that the based on not crediting active fire barrier results are scenario driven (the zone of elements as a partitioning feature for the roll-up influence was not arbitrarily limited to the zone fire door between Block House 1&2 and CT-4 or boundary). Other than the rollup door between fire door closure devices in the SSF. BH12 and CT4 which was credited for deterministic fire area partitioning, active fire barrier elements were not credited.
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| F&O PP-C3-1 Improve general description and Partially addressed via multi-compartment See NRC staff's finding on F&O FSS-G1-1.
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| identification of unique fire protection features. analysis (see PP-B2-1); failure to meet SR poses no adverse impact on the analysis quality or completeness.
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| NRC staff F&O text: The lack of rated barriers was documented in OSC-8979:
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| F&O PRM-B1-1 Impact of the internal event See IEPRA-1. See NRC staff's finding on F&O IEPRA-1.
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| PRA peer review open items on Fire PRA not addressed.
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| NRC staff F&O text: Oconee used a version of the internal events model with a substantial number of outstanding issues (see F&O IEPRA
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| : 1) as the base model. The finding is based on the fact that the issues identified in the peer review have not been resolved.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 216 Attachment C, Fire Risk Evaluation Tables Table 3.4-2, Fire Events PRA F&O Resolution Facts and Observations Licensee's Disposition NRC Staff's Findings F&O PRM-D1-1 Circa 2005 fire structure in the Eliminated reliance on initiators % TBOFIRE and The licensee's response did not completely PRAE model not peer reviewed (SR PRM-D1 %CSFIRE from pre-existing fire structure in Fire address the NRC staff's observation that the deleted and is now PRM-C1). PRA. final documentation needs to clearly describe all fire related changes made to the internal events PRA. However, based NRC staff F&O text: Final documentation on the NRC staff's review of the Fire PRA should identify and clearly describe all fire-and the licensee's response to the NRC related changes made to the internal event PRA staff's review, the NRC staff finds the in one document to meet the standard. The licensee has appropriately resolved this current documentation provides an incomplete issue for this application, because any and sometimes contradictory description of additional changes are not likely to cause proposed changes versus real changes.
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| the estimated transition risk decrease to become a risk increase.
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| F&O QLS-A3-1 Discussion in Partitioning & No impact on quantification; updated calculation Based on the identification of corrective Ignition Frequency calc implied that actions to provide justification for exclusion for the 4 actions being completed, the NRC staff pertaining to 4 structures that were excluded structures to reflect that no further action was concludes that this deficiency has been from ignition source counting had not been necessary for these structures based on the appropriately resolved by the licensee for completed. mUlti-compartment analysis. this application.
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| NRC staff F&O text: 22 buildings structures screened, 4 left unresolved.
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| QNS-C1 The NRC audit found that this The licensee dispositions this supporting The NRC staff finds that the justification that supporting requirement was met at capability requirement as not applicable since screening QNS-C1 is not applicable to ONS is category I based on no evaluation having been criteria was not applied; if a building or structure acceptable.
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| performed to demonstrate that capability (or an area in the case of the switchyard) category II or III was met. contained PRA credited equipment/cables and/or could result in loss of offsite power, it was not screened.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 217 Attachment C, Fire Risk Evaluation Tables Table 3.4-2, Fire Events PRA F&O Resolution Facts and Observations Licensee's Disposition NRC Staff's Findings F&O SF-A2-1 Conduct assessment of the No impact on quantification of Fire PRA or In response to RAI 5-28, the licensee further potential for diversion of suppression flow. Change Evaluations (seismic-fire interaction is clarified that ONS complies with the purely qualitative per NUREG/CR-6850). See requirements of NFPA 805 Section 3.6.4, qualitative discussion of seismic fires in the Unit "Standpipe and Hose Stations Earthquake NRC staff F&O text: A seismic induced 3 Fire PRA Summary report. PIP G-09-00698 Provisions," thus satisfying this supporting assessment of the potential for diversion of will track the resolution of this open item. requirement. Since this SR requires only a suppressants from areas where it is needed for qualitative assessment and will not impact fire suppression systems associated with a quantification of the Fire PRA, and based on common suppressant supply was not the licensee's evaluation that this supporting conducted.
| |
| requirement is already met, the NRC staff concludes that this deficiency has been appropriately resolved by the licensee for this application.
| |
| F&O SF-A4-1 Plant seismic response No impact on quantification of Fire PRA or In response to RAI 5-28, the licensee further procedures do not cover seismically induced Change Evaluations (seismic-fire interaction is clarified that the fire response procedure is fire. purely qualitative per NUREG/CR-6850). PIP entered either via a fire alarm annunciator G-09-00698 will track the resolution of this open or the report of a fire, either of which applies item. at all times and under any plant operating NRC staff F&O text: The plant seismic conditions and that, therefore, a reference response procedures cover seismically induced to the fire responses procedure in the flooding, but not seismically induced fires.
| |
| seismic response procedure is unnecessary. Since this SR requires only a qualitative assessment and will not impact quantification of the Fire PRA, and based on the licensee's evaluation that existing ONS procedures adequately address response to seismically induced fires, the NRC staff concludes that this deficiency has been appropriately resolved by the licensee for this application.
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| OFFICIAL USE ONLY SECURITY REL'\TED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 218 Attachment C, Fire Risk Evaluation Tables Table 3.4-2, Fire Events PRA F&O Resolution Facts and Observations Licensee's Disposition NRC Staff's Findings F&O SF-A5-1 Assessment of earthquake impact No impact on quantification of Fire PRA or In response to RAI 5-28, the licensee further on fire brigade not documented Change Evaluations (seismic-fire interaction is noted that fire brigade response during a purely qualitative per NUREG/CR-6850). PIP seismic event has been considered in ONS G-09-00698 will track the resolution of this open Standard Operating Guide (SOG) #1 in that NRC staff F&O text: No assessment has been item. ONS has staged fire brigade equipment so conducted on the potential that an earthquake that one single event will not render the fire might compromise one or more of the fire brigade ineffective. Since this SR requires brigade members.
| |
| only a qualitative assessment and will not impact quantification of the Fire PRA, and based on the licensee's evaluation that existing ONS procedures adequately address fire brigade response to seismically induced fires, the NRC staff concludes that this deficiency has been appropriately resolved by the licensee for this application.
| |
| F&O UNC-A1-1 Uncertainty and sensitivity Uncertainty & Sensitivity Matrix added as Based on the NRC staff's review of the Fire analysis incomplete (not reviewed). Appendix 0 to Fire PRA Summary Report; PRA and the licensee's response to the sensitivity quantitatively addressed in NFPA 805 NRC staff's review, the NRC staff finds that Fire PRA Application Calculation. it is unlikely that a completed uncertainty NRC staff F&O text: Not reviewed. When the and sensitivity analysis will cause the analysis is complete and stable, the sources of estimated risk decrease to become a risk model uncertainty should be identified.
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| increase.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 219 Attachment C, Fire Risk Evaluation Tables Table 3.4-3, V&V Basis for Fire Modeling Correlations Application at Correlation ONS V&V Basis NRC Staff Evaluation of Acceptability Flame Height Provide a limit on NUREG-1805
| |
| * The licensee stated that the flame height correlation is used in the use of the ZOL NUREG-1824 both the Consolidated Fire and Smoke Transport Model (CFAST) and NUREG-1805 fire models, for which V&V was documented in NUREG-1824.
| |
| * The licensee stated that use of the correlation was limited to its range of applicability.
| |
| Since the (1) V&V basis is NUREG-1824 and (2) the licensee stated that the correlation was applied within the limits of its applicability, the NRC staff finds use of this correlation in the ONS application acceptable.
| |
| Radiant Heat Flux Calculates target NUREG-1805
| |
| * The licensee stated that this correlation produces the most Method of Shokri heat flux to NUREG-1824 conservative results of the four correlations considered in the and Seyler determine the SFPE calculation, while maintaining a credible radiated energy (detailed) lateral extent of the Engineering fraction from the flame shapes postulated.
| |
| ZOL Guide,1999
| |
| * The licensee stated that the correlation is used in the NUREG 1805 fire model, for which V&V was documented in NUREG 1824, and the V&V basis for the correlation is documented in an authoritative publication of the SFPE Engineering Guide.
| |
| * The licensee stated that the correlation was used within the limits of its range of applicability.
| |
| Since (1) the correlation produces conservative results, (2) the V&V basis is NUREG-1824 and an authoritative publication of the SFPE Engineering Guide, and (3) the licensee stated that the correlation was applied within the limits of its applicability, the NRC staff finds use of this correlation in the ONS application acceptable.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 220 Attachment C, Fire Risk Evaluation Tables Table 3.4-3, V&V Basis for Fire Modeling Correlations Application at Correlation ONS V&V Basis NRC Staff Evaluation of Acceptability Plume Heat Fluxes Calculates the Wakamatsu et
| |
| * The licensee stated that the lOI vertical separation distance vertical separation aI., 2003 SFPE used in the ONS application is based on the more severe result distance to the Handbook 4 th from this calculation and the calculation for plume centerline target in order to Edition, Chap. 2 temperature (see below).
| |
| determine the 14, Lattimer, B.,
| |
| vertical extent of 2008
| |
| * The licensee stated that the correlation is documented in an the lOI. authoritative publication of the SFPE Handbook of Fire Protection Engineering and in a published conference report.
| |
| * The licensee stated that the correlation was used within the limits of its range of applicability.
| |
| Since (1) the application uses the more conservative of the two correlations for determining the vertical extent of the lOI, (2) the V&V basis is an authoritative publication of the SFPE Handbook of Fire Protection Engineering and a published article in a conference report, and (3) the licensee stated that the correlation was applied within the limits of its applicability, the NRC staff finds use of this correlation in the ONS application acceptable.
| |
| Plume Centerline Calculates the NUREG-1805
| |
| * The licensee stated that the lOI vertical separation distance Temperature vertical separation used in the ONS application is based on the more severe result distance to the NUREG-1824 from this calculation and the calculation for Plume Heat Fluxes target to determine (see above).
| |
| the vertical extent SFPE Handbook of the lOI. 4 th Edition,
| |
| * The licensee stated that the correlation is used in the NUREG Chap. 2-1, 1805 fire model, for which V&V was documented in NUREG Heskestad, G., 1824 and the V&V basis for the correlation is documented in an 2008 authoritative publication of the SFPE Handbook of Fire Protection Engineering.
| |
| * The licensee stated that the correlation was used within the OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 221 Attachment C, Fire Risk Evaluation Tables Table 3.4-3, V&V Basis for Fire Modeling Correlations Application at Correlation ONS V&V Basis NRC Staff Evaluation of Acceptability limits of its range of applicability.
| |
| Since (1) the application uses the more conservative of the two correlations for determining the vertical extent of the ZOI, (2) the V&V basis is NUREG-1824 and an authoritative publication of the SFPE Handbook of Fire Protection Engineering, and (3) the licensee stated that the correlation was applied within the limits of its applicability, the NRC staff finds use of this correlation in the ONS application acceptable.
| |
| Hydrocarbon Spill Determines the SFPE Handbook
| |
| * The licensee stated that the correlation is documented in an Fire Size heat release rate 3rd Edition, authoritative publication of the SFPE Handbook of Fire for unconfined Chap. 3-11, Protection Engineering.
| |
| hydrocarbon spill Seyler, C., 2002 fires.
| |
| * The licensee stated that there were no limits in treatment of range for this correlation because the spill transition from unconfined to deep pool burning would be abrupt.
| |
| Since the V&V basis is an authoritative publication of the SFPE Handbook of Fire Protection Engineering, the NRC staff finds use of this correlation in the ONS application acceptable.
| |
| Flame Extension Determines the fire SFPE Handbook
| |
| * The licensee stated that the correlation is documented in an offset for open 3rd Edition, authoritative publication of the SFPE Handbook of Fire panel fires; only Chap. 3-11, Protection Engineering.
| |
| used when the Beyler, C., 2002 possibility of flame
| |
| * The licensee stated that the correlation applies to fires ranging extensions are in size from about 10 kW to 1.0 MW, which bounds the fire size present. bins for open electronic equipment cabinets defined in NUREG/CR-6850.
| |
| Since the V&V basis is an authoritative publication of the SFPE Handbook of Fire Protection Engineering, and since the applicable fire OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 222 Attachment C, Fire Risk Evaluation Tables Table 3.4-3, V&V Basis for Fire Modeling Correlations
| |
| ...
| |
| Application at Correlation ONS V&V Basis NRC Staff Evaluation of Acceptability size range bounds the fire size bins for open electronic equipment cabinets defined in NUREG/CR-6850 (Reference 37), the NRC staff finds use of this correlation in the ONS application acceptable.
| |
| Corner Flame Determines the SFPE Handbook
| |
| * The licensee stated that this correlation is used to ensure that a Height heat release rate 3rd Edition, conservative separation distance is calculated when the fire for fires that are Chap. 2-14, dynamics have limited entrainment conditions.
| |
| proximate to a Lattimer, S.,
| |
| corner to 2002
| |
| * The licensee stated that the correlation is documented in an determine the authoritative publication of the SFPE Handbook of Fire vertical extent of Protection Engineering.
| |
| the lOI.
| |
| * The licensee stated that the correlation applies to corner fires ranging in size from about 10 kW to 1.0 MW, which bounds the fire size bins for open electronic equipment cabinets defined in NUREG/CR-6850.
| |
| Since (1) the correlation produces conservative results, (2) the V&V basis is an authoritative publication of the SFPE Handbook of Fire Protection Engineering, and (3) since the applicable fire size range bounds the fire size bins for open electronic equipment cabinets defined in NUREG/CR-6850 (Reference 37), the NRC staff finds use of this correlation in the ONS application acceptable ..
| |
| Line Fire Plume Calculates Yuan et aI., 1996
| |
| * The licensee stated that the correlation is documented in a Centerline simplified (Peer-reviewed peer reviewed journal article.
| |
| Temperature separation journal distances to the experimental
| |
| * The licensee stated that the approach bounds the methods for target to provide a data) predicting the plume centerline temperature above the line type limit on the use of source fire.
| |
| the lOI and the extent of the lOI OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 223 Attachment C, Fire Risk Evaluation Tables Table 3.4-3, V&V Basis for Fire Modeling Correlations Application at Correlation ONS V&V Basis NRC Staff Evaluation of Acceptability for cable tray fires. Since the V&V basis is a peer-reviewed journal article, and the approach bounds the methods for predicting the plume centerline temperature above the line type source fire, the NRC staff finds use of this correlation in the ONS application acceptable.
| |
| Ventilation Limited Assesses the SFPE
| |
| * The licensee stated that this correlation was used to develop Fire Size significance of vent Engineering generic rules for various ventilation opening sizes while position on the hot Guide, 2004 ensuring bounding cases are applied.
| |
| gas layer temperature to
| |
| * The licensee stated that the correlation is documented in an determine the fire authoritative publication of the SFPE Engineering Guide.
| |
| size due to ventilation
| |
| * The licensee stated that there were no limits in treatment of limitations. range for this correlation because the most severe environment predicted was the most conservative for the given room volumes and ventilation opening areas.
| |
| Since (1) the methodology ensures use of bounding results, (2) the V&V basis is an authoritative publication of the SFPE Engineering Guide, and (3) the correlation was only used for fire scenarios where it was applicable. The NRC staff finds use of this correlation in the ONS application acceptable.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 224 References for Table 3.4-3
| |
| : 1. NUREG-1805, "Fire Dynamics Tools (FDTS ) Quantitative Fire Hazard Analysis Methods for the U.S. Nuclear Regulatory Commission Fire Protection Inspection Program," U.S. NRC, Washington, DC, December 2004.
| |
| : 2. NUREG-1824, "Verification & Validation of Selected Fire Models for Nuclear Power Plant Applications", U.S. NRC, Washington, DC, May, 2007.
| |
| : 3. The SFPE Engineering Guide, "Assessing Flame Radiation to External Targets from Pool Fires," Society of Fire Protection Engineers, Bethesda, Maryland, June, 1999.
| |
| : 4. Wakamatsu, T., Hasemi, Y., Kagiya, K., and Kamikawa, D., "Heating Mechanism of Unprotected Steel Beam Installed Beneath Ceiling and Exposed to a Localized Fire:
| |
| Verification Using the Real-scale Experiment and Effects of the Smoke Layer," Proceedings of the Seventh International Symposium on Fire Safety Science, International Association for Fire Safety Science, London, UK, 2003.
| |
| : 5. Lattimer, B. Y., "Heat Fluxes from Fires to Surfaces," Chapter 2-14, The SFPE Handbook of Fire Protection Engineering, 4 th Edition, P. J. DiNenno, Editor-in-Chief, National Fire Protection Association, Quincy, MA, 2008.
| |
| : 6. Beyler, C.L., "Fire Hazard Calculations for Large, Open Hydrocarbon Fire," Chapter 3-11, The SFPE Handbook of Fire Protection Engineering, 3rd Edition, P. J. DiNenno, Editor-in Chief, National Fire Protection Association, Quincy, MA, 2002.
| |
| : 7. Heskestad, G., "Fire Plumes, Flame Height, and Air Entrainment," Chapter 2-1, The SFPE Handbook of Fire Protection Engineering, 4 th Edition, P. J. DiNenno, Editor-in-Chief, National Fire Protection Association, Quincy, MA, 2008.
| |
| : 8. Lattimer, B. Y., "Heat Fluxes from Fires to Surfaces," Chapter 2-14, The SFPE Handbook of Fire Protection Engineering, 3rd Edition, P. J. DiNenno, Editor-in-Chief, National Fire Protection Association, Quincy, MA, 2002.
| |
| : 9. Yuan, L. and Cox, F., "An Experimental Study of Some Line Fires," Fire Safety Journal,"
| |
| Volume 27, 1996.
| |
| : 10. The SFPE Engineering Guide, "Fire Exposures to Structural Elements," Society of Fire Protection Engineers, Bethesda, Maryland, May 2004.
| |
| : 11. Kleinsorg Group Risk Services, LLC, "Generic Fire Modeling Treatments," Revision 0, January 23, 2008, prepared for ERIN Engineering and Research, Inc.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 225 Attachment D, Nuclear Safety Capability Assessment Results by Fire Area is broken down into those ONS fire areas that were analyzed using the deterministic approach in accordance with NFPA 805, Section 4.2.3, and those using the PB approach in accordance with NFPA 805, Section 4.2.4.
| |
| Each fire area includes a discussion of how the licensee met the NFPA 805 requirement to evaluate the potential fire suppression effects on the ability to meet the nuclear safety performance criteria.
| |
| Each fire area also contains a section that addresses those NRC approved exemptions and other licensing actions that exempt the licensee from the existing deterministic fire protection licensing basis that the licensee desires to incorporate into the RI/PB FPP, as allowed by NFPA 805, Section 2.2.7. This discussion for each applicable fire area includes a description of the previously approved exemption or other licensing action exempting the licensee from the deterministic requirements, the basis for and continuing validity of the exemption or other licensing action, and the NRC staff's evaluation of that exemption or other licensing action.
| |
| Where required, each section of the attachment includes an evaluation of the DID recovery actions necessary for the applicable fire area. As discussed in SE Section 3.2.4, the licensee credited recovery actions to satisfy the DID requirements of NFPA 805, Section 1.2, but are not needed to maintain the availability of a success path and do not adversely impact risk. Because the licensee has identified these recovery actions as being necessary to provide adequate DID, the NRC staff has evaluated them as a part of the RI/PB FPP. As such, future removal of these DID recovery actions would require a plant change evaluation in accordance with NFPA 805, Section 2.4.4.
| |
| For all fire areas where the licensee utilized the PB approach to meet the nuclear safety performance criteria, each VFDR and the associated disposition has been listed.
| |
| As a part of the NSCA, the licensee evaluated fire detection and suppression systems on a fire area basis. Accordingly, the evaluation of each fire area includes a table that documents the licensee's review of these fire detection and suppression systems, as well as the NRC staff's evaluation of the review and its results.
| |
| Finally, each fire area includes a summary assessment documenting the NRC staff's conclusion regarding the ability to meet the NFPA 805 requirements and the associated nuclear safety performance criteria.
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| OFFICIAL USE ONLY SECURITY RELi\TED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 226 Attachment D1, Deterministic Compliance with NFPA 805 Section 4.2.3 For each fire area where the licensee selected the deterministic approach to demonstrate compliance with the NFPA 805 requirements, the NRC staff verified that the deterministic requirements of NFPA 805, Section 4.2.3, are met without the use of recovery actions. Fire areas that meet the deterministic requirements of NFPA 805 are deemed to adequately satisfy the nuclear safety performance criteria, as stated in NFPA 805, Section 1.5.1.
| |
| The licensee evaluated suppression and detection systems using a process that looked at several key aspects of the FPP to determine if a given system is required (i.e., deterministically in support of compliance with NFPA 805 Chapter 4, in support of a previous NRC approved alternative or in support of a licensee-developed EEEE)..
| |
| Accordingly, each of the fire areas listed below include a section discussing those fire suppression and fire detection systems the licensee has determined to be required to meet the nuclear safety performance criteria.
| |
| Fire Area CT-4, CT -4 Block House The licensee stated that deterministic compliance has been met in accordance with NFPA 805 Section 4.2.3.2, without the use of recovery actions, which requires that one success path of required cables and equipment be located in a separate area having boundaries containing fire barriers with a minimum fire resistance rating of three hours. The licensee identified the SSCs necessary to meet the nuclear safety performance criteria in this fire area.
| |
| Evaluation of Fire Suppression Effects on the Nuclear Safety Performance Criteria The licensee stated in LAR Attachment C, "NEI 04-02 Table B-3 Fire Area Transition," those safe and stable conditions can be achieved and maintained using equipment and cables outside the area of fire suppression activity. Flooding of the suppression areas and discharge of suppression water to adjacent compartments is controlled and will not jeopardize achievement of safe and stable conditions. Based on the information provided by the licensee in the NFPA 805 LAR, the NRC staff finds the licensee's evaluation of fire suppression effects on nuclear safety performance criteria acceptable because the results of the licensee's analysis indicate that fire suppression activities will not adversely affect achievement of the nuclear safety performance criteria.
| |
| Variation from Deterministic Requirements (VFDRs)
| |
| Based on the information provided in the LAR, the licensee did not identify any VFDRs or any previous exemptions or other licensing actions credited in transition.
| |
| Fire Detection & Suppression Systems Required to Meet the Nuclear Safety Performance Criteria The licensee performed an evaluation of the fire detection and suppression systems in this area. The results of the evaluation were documented in LAR Table 4-4, "Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features."
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 227 The applicable portion of Table 4-4 has been included below identifying the ionization smoke detection and CT-4 transformer deluge as required detection and suppression systems to support the engineering evaluation.
| |
| Suppression Required Detection Required System?
| |
| System?
| |
| Fire Fire Area Zone Description Zone E R D S E R D S CT4 46 CT-4 Block House Yes No No No Yes No No No Legend:
| |
| E - EEEE/LA: Systems required for acceptability of EEEE / NRC-approved Licensing Action (Section 2.2.7)
| |
| R - Risk: Systems required to meet the Risk Criteria for the PB Approach (Section 4.2.4)
| |
| D - Defense-in-Depth: Systems required to maintain adequate balance of DID for a PB Approach (Section 4.2.4.2)
| |
| S - Separation Criteria: Systems required for NFPA 805, Chapter 4 Separation Criteria in (Section 4.2.3)
| |
| MR - Modification Required Systems are committed to be modified as indicated in Table 4-4 and Attachment S of LAR Fire Area CT-4 Conclusion Based on the information provided in the LAR, as supplemented, the NRC staff finds Fire Area CT-4 meets the deterministic requirements of NFPA 805 Section 4.2.3. This conclusion is based on the following:
| |
| : 1. The licensee's documented compliance to NFPA 805 Section 4.2.3.2 and the licensee's assertion that the success path will be free of fire damage without reliance on recovery actions.
| |
| : 2. The licensee's assessment of the impact of suppression systems on the ability to meet the nuclear safety performance criteria.
| |
| : 3. The licensee's determination of the suppression and detection systems required to meet the nuclear safety performance criteria.
| |
| Fire Area KED, Keowee Hydro Station The licensee stated that deterministic compliance has been met in accordance with NFPA 805 Section 4.2.3.2, without the use of recovery actions, which requires that one success path of required cables and equipment be located in a separate area having boundaries containing fire barriers with a minimum fire resistance rating of three hours. The licensee identified the SSCs necessary to meet the nuclear safety performance criteria in this fire area.
| |
| Evaluation of Fire Suppression Effects on Nuclear Safety Performance Criteria The licensee stated in LAR Attachment C, "NEI 04-02 Table B-3 Fire Area Transition," those safe and stable conditions can be achieved and maintained using equipment and cables outside of the area of fire suppression activity. Flooding of the suppression areas and discharge of suppression water to adjacent compartments is controlled and will not jeopardize achievement of safe and stable conditions either. Based on the information provided by the licensee in the NFPA 805 LAR, the NRC staff finds the licensee's evaluation of fire suppression effects on nuclear safety performance criteria acceptable because the results of the licensee's analysis indicate that fire suppression activities will not adversely affect achievement of the nuclear safety performance criteria.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 228 Variation from Deterministic Requirements (VFDRs)
| |
| Based on the information provided in the LAR, the licensee did not identify any VFDRs or any previous exemptions or other licensing actions credited in transition.
| |
| Fire Detection & Suppression Systems Required to Meet the Nuclear Safety Performance Criteria The licensee performed an evaluation of the fire detection and suppression systems in this area.
| |
| The results of the evaluation were documented in LAR Table 4-4, "Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features." The applicable portions of Table 4-4 have been included below.
| |
| Suppression Required Detection Required System?
| |
| System?
| |
| Fire Fire Zone Description Area Zone E R D S E R D S KED KED Keowee Hydro Station No No No No No No No No Legend:
| |
| E - EEEE/LA: Systems required for acceptability of EEEE / NRC-approved Licensing Action (Section 2.2.7)
| |
| R - Risk: Systems required to meet the Risk Criteria for the PB Approach (Section 4.2.4)
| |
| D - Defense-in-Depth: Systems required to maintain adequate balance of DID for a PB Approach (Section 4.2.4.2)
| |
| S - Separation Criteria: Systems required for NFPA 805, Chapter 4 Separation Criteria in (Section 4.2.3)
| |
| MR - Modification Required Systems are committed to be modified as indicated in Table 4-4 and Attachment S of LAR Fire Area KEO Conclusion Based on the information provided in the LAR, as supplemented, the NRC staff finds Fire Area KEO meets the deterministic requirements of NFPA 805, Section 4.2.3. This conclusion is based on the following:
| |
| : 1. The licensee's documented compliance to NFPA 805 Section 4.2.3.2, and assertion that the success path will be free of fire damage without reliance on recovery actions.
| |
| : 2. The licensee's assessment of the impact of suppression systems on the ability to meet the nuclear safety performance criteria.
| |
| : 3. The licensee's determination that suppression and detection systems were not required to meet the nuclear safety performance criteria.
| |
| Fire Area PSW, Protected Service Water WSW) Building (Planned)
| |
| The PSW building is a plant modification that had not been constructed as of the issuance of this SE, but is credited by the licensee in its RI/PB FPP. The licensee stated that deterministic compliance has been met in accordance with NFPA 805, Section 4.2.3.2, without the use of recovery actions, which requires that one success path of required cables and equipment be located in a separate area having boundaries containing fire barriers with a minimum fire resistance rating of three hours. The licensee identified the SSCs necessary to meet the nuclear safety performance criteria in this fire area.
| |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 229 Evaluation of Fire Suppression Effects on Nuclear Safety Performance Criteria No evaluation performed since the PSW had not been constructed at the time of the LAR submittal.
| |
| Variation from Deterministic Requirements (VFDRs)
| |
| VFDR# VFDR Description Component PSW Modification (SE PSW-01 Ensure PSW modification is incorporated into ale the required documents.
| |
| Section 2.8.1, Item 1).
| |
| Recovery Actions (RAs)
| |
| The licensee did not identify any recovery actions required for this fire area.
| |
| Fire Detection & Suppression Systems Required to Meet the Nuclear Safety Performance Criteria The licensee performed an evaluation of the fire detection and suppression systems needed for this area when constructed. The results of the evaluation were documented in LAR Table 4-4, "Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features." The applicable portions of Table 4-4 have been included below. Detection will be installed during Implementation of the PSW modification (SE Section 2.8.12, Item 1). The identified fire detection system is relied upon to meet the DID criteria.
| |
| Suppression Required Detection Required System?
| |
| System?
| |
| Fire Fire Zone Description Area Zone E R D S E R D S Protected Service Water PSW PSW No No No No No No No No Buildino Legend:
| |
| E - EEEElLA: Systems required for acceptability of EEEE I NRC-approved Licensing Action (Section 2.2.7)
| |
| R - Risk: Systems required to meet the Risk Criteria for the PB Approach (Section 4.2.4)
| |
| D - Defense-in-Depth: Systems required to maintain adequate balance of DID for a PB Approach (Section 4.2.4.2)
| |
| S - Separation Criteria: Systems required for NFPA 805, Chapter 4 Separation Criteria in (Section 4.2.3)
| |
| MR - Modification Reauired Systems are committed to be modified as indicated in Table 4-4 and Attachment S of LAR Fire Area PSW Conclusion Based on the information provided in the LAR, as supplemented, the NRC staff finds Fire Area PSW meets the deterministic requirements of NFPA 805, Section 4.2.3. This conclusion is based on the following:
| |
| : 1. The licensee's documented compliance to NFPA 805, Section 4.2.3.2, and assertion that the success path will be free of fire damage without reliance on recovery actions.
| |
| : 2. The licensee's assessment of the impact of suppression systems on the ability to meet the nuclear safety performance criteria.
| |
| : 3. The licensee's determination that suppression and detection systems were not required to meet the nuclear safety performance criteria.
| |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 230 Attachment 02, Performance-Based Compliance with NFPA 805, Section 4.2.4 For each fire area where the licensee selected FRE as the PB approach, the NRC staff verified that the change in risk is appropriately defined, the magnitude is acceptable and DID and sufficient SMs are maintained. The NRC staff has also verified that the additional risk of recovery actions is acceptable.
| |
| The licensee included an assessment of DID and SMs in the FRE for each of the areas addressed using the PB approach. Each FRE assessed the aspects of DID, including passive fire protection features (fire barriers, through penetration fire stops, penetration seals, radiant energy shields, etc.), active fire protection features (doors and dampers), and programmatic controls (combustible controls, hot work, design-flame spread of surfaces, electrical design, etc.), as well as manual suppression using fire extinguishers and hoses.
| |
| The licensee evaluated suppression and detection systems using a process that looked at several key aspects of the FPP to determine if a given system is required (i.e., deterministically in support of compliance with NFPA 805 Chapter 4, in support of a previous NRC approved exemption or other licensing action, in support of a licensee-developed EEEE, or as a result of the PB evaluations).
| |
| Accordingly, in addition to a discussion regarding risk, recovery actions (as applicable), DID, and SMs, each of the fire areas listed below also include a discussion of those fire suppression and fire detection systems the licensee has determined to be required to meet the nuclear safety performance criteria.
| |
| The licensee included in the VFOR motor-operated valves that were susceptible to failure as described in NRC Information Notice 92-18 "Potential for Loss of Remote Shutdown Capability During a Control Room Fire" (Reference 55). This issue addressed hot shorts, combined with the absence of thermal overload protection, which could cause valve damage before the operator shifted control of the valves to the remote/ alternate shutdown panel. The Licensee's description of the variance includes "This valve may suffer IN 92-18 damage. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue." Individual risk evaluations for these failures were dispositioned as "E" for both the change in COF and change in LERF and fire detection coverage was identified as required for DID in each case. An evaluation for compliance using the PB approach of NFPA 805, Section 4.2.4 was performed for each potential failure.
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| Fire Area AB. Auxiliary Building The licensee analyzed this fire area using the FRE approach in accordance with NFPA 805 Section 4.2.4.2, but also applied deterministic simplifying assumptions in order to credit those portions of the facility design that meet the deterministic requirements of NFPA 805 Section 4.2.3. The licensee identified the SSCs necessary to meet the nuclear safety performance criteria in this fire area.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 231 Evaluation of Fire Suppression Effects on Nuclear Safety Performance Criteria for Fire Area AB The licensee stated in LAR Attachment C, "NEI 04-02 Table B-3 Fire Area Transition," that a safe and stable condition can be achieved and maintained using equipment and cables outside of the area of fire suppression activity. Flooding of the suppression areas and discharge of suppression water to adjacent compartments is controlled and will not jeopardize achievement of safe and stable conditions. Based on the information provided by the licensee in the NFPA 805 LAR, the NRC staff finds the licensee's evaluation of fire suppression effects on nuclear safety performance criteria acceptable because the results of the licensee's analysis indicate that fire suppression activities will not adversely affect achievement of the nuclear safety performance criteria.
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| Exemptions and Other Licensing Actions Based on the information provided in the LAR, the licensee credited four previously approved licensing actions and exemptions from the existing fire protection requirements. The licensee IJsed the process described in LAR Section 4.2.3, "Licensing Action Transition," and Attachment K, "Licensing Action Review," to carry forward these exemptions and other licensing actions, which requires a determination of the basis of acceptability and a determination that the basis of the acceptability is still valid. The NRC staff's evaluation of each exemption and other licensing actions is provided in the table below.
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| Exemption/Licensing Licensee's Statement on Basis and Continuing Validity NRC Staff Evaluation Action Appendix R The lack of source range monitoring is acceptable because: Based on the previous Exemption, SSF, Lack
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| * Unit held at hot standby. NRC staff approval of this of instrumentation per
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| * Control rods are inserted. exemption and the 1I1.L.2
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| * RCS makeup and boration is with SFP water as this statement by the licensee is the only source available with the existing piping that the basis remains design. valid, the NRC staff finds The lack of SG pressure instrumentation is acceptable this acceptable.
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| because:
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| * Steam pressure is not a control parameter for operators.
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| * SG level will be used to control auxiliary feedwater flow.
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| The bases for previous acceptance remain valid.
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| Approval of SSD The main design features: Based on the previous System (SSS) Design
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| * Capable of maintaining a hot standby condition in all NRC staff approval of this units without any damage control measures and the licensing action and the ability to withstand SSD earthquake seismic loadings. statement by the licensee
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| * Utilizes natural circulation to remove decay heat from that the basis remains the primary coolant, use of secondary side steam valid. the NRC staff finds valves to the atmosphere as a heat sink, and this acceptable.
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| providing an independent power system.
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| * The SSF is designed to provide an alternate and independent means to achieve and maintain hot standby conditions for one or more of the three ONS units.
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| The bases for previous acceptance remain valid.
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| Appendix R Provides the following justification for the lack of three hour fire Based on the previous Exemption, RB rated pipe penetrations: NRC staff approval of this Unrated Containment
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| * RB walls serve as a substantial heat sink. exemption and the Mechanical
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| * Combustible loading near penetrations is low. statement by the licensee Penetrations that the basis remains OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 232 Exemption/Licensing Licensee's Statement on Basis and Continuing Validity NRC Staff Evaluation Action
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| * Mechanical pipe penetrations are designed to meet valid, the NRC staff finds multiple containment integrity criteria and are this acceptable.
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| substantial.
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| * Large room volumes on both sides dissipate heat from a fire away from penetration area.
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| The bases for previous acceptance remain valid.
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| Appendix R Presented justification for the lack of three hour fire barriers Based on the previous Exemption, AB Lack because: NRC staff approval of this of three hour fire rated
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| * Low combustible loading in pipe tunnel access area. exemption and the barrier
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| * Fire propagation path is circuitous, consisting of statement by the licensee several unrated barriers and open areas. that the basis remains
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| * If a fire were to occur, it would develop slowly. valid, the NRC staff finds this acceptable.
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| * Fire brigade may use portable extinguishers, manual hose stations, or a fire hose supplied from a nearby fi re hydrant.
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| Although the exact number and configuration of combustibles may have changed over time, the bases for previous acceptance remain valid as substantiated bv field walkdown Variation from Deterministic Requirements (VFDRs)
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| Fire Area AB has a total of 52 VFDRs, which are provided in the table below. 36 of these VFDRs that are NFPA 805 Section 4.2.3 (separation issues) that were dispositioned with a FRE (SE Section 3.4.3). The licensee's FRE determined that these variances are acceptable based on 1) the change in CDF and LERF for the fire area and the total CDF and LERF for each unit meet the acceptance criteria of RG 1.174 (SE Section 3.4.6) and 2) adequate DID and SMs are maintained for each fire area (SE Section 3.4.2). This determination relies on the following fire protection systems and features to meet the acceptance criteria:
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| * General area and/or hazard detection for the fire area AB is required to meet the risk acceptance criteria. The Fire PRA makes assumptions regarding the time of fire discovery, fire brigade notification, and brigade manual suppression. These assumptions determine the impact of the fire, including the likelihood of a HGL being formed in the compartment. Specifically, the Fire PRA assumes a fire brigade response time of 20 minutes or less. The existing fire zone detection system coverage of the general area and/or hazard necessary for this assumption to be valid was not considered sufficient to conservatively meet the risk criteria. Therefore, modifications to the fire detection system for fire area AB are required to support the fire risk analysis assumption of 20 minute brigade response time.
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| Based on the reliance on fire detectors in fire area AB to meet the risk criteria, the licensee has committed to make modifications to the fire detection system, which may include fire detector upgrades and/or new installation. Improvements to the following fire zones in AS for general area and/or fire hazard detection are required: 61,68,72,77,94,99,103,108,110,112,115, 118, and 121 (SE Section 2.8.1).
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 233 Five of the 51 VFDRs, AB-34, AB-35, AB-40, AB-41 , and AB-42 are a VFDR of NFPA 805 Section 4.2.3 (separation issue) that will be corrected with a plant modification. According to the LAR, the walls separating the following areas will be modified:
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| * TB / AB
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| * AB/ West Penetration Rooms
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| * Unit 1 Purge Inlet Room / SFP Area
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| * Unit 2 Purge Inlet Room / SFP Area
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| * Unit 3 Purge Inlet Room / SFP Area These barriers are not currently three hour fire rated walls as required by NFPA 805 Section 3.11.1, and all of the penetrations in the wall do not have a fire resistance three hour rating as required by NFPA 805 Section 3.11.3. These barriers are credited for fire area separation using the deterministic approach of NFPA 805 Section 4.2.3. The licensee has committed to make modifications to the walls to bring them into compliance with the requirements of NFPA 805 (SE Section 2.8.1).
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| One VFDR, AB-39, regarding the SSF DG requires that the monitoring and/or adjustment of the following parameters is required during operation of the SSF DG: generator current, voltage, power and frequency. The controls and indications required to monitor and adjust these parameters are currently not included in the SSD analysis for those fire areas where the SSF is credited for accomplishing SSD (AB Fire Area only). Incorporation of these activities into the SSD procedure is an implementation item and resolve this VFDR (SE Section 2.9, Table 2.9 1,ltem 13).
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| By letter dated November 19, 2010 (Reference 52), the licensee identified nine additional VFDRs (AB-43 through AB-51) as a result of eliminating the "10-minute free ottire damage" assumption. This issue is discussed in greater detailing SE Section 3.2.1. Upon completion of these activities, all applicable FREs will be updated and compliance will be demonstrated.
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| These activities are an implementation item (SE Section 2.9, Table 2.9-1, Item 46).
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| In addition, by letter dated November 19, 2010 (Reference 52), the licensee identified an additional VFDR (AB-52) as a result of reclassifying the deployment and operation of the SSF submersible pump as a (recovery action) RA. This issue is discussed in greater detail in SE Section 3.2.4. The licensee conducted a bounding assessment from the additional risk of this RA. The evaluation determined the risk was sufficiently small and met the risk acceptance guidelines associated with pre-approved RAs. For a discussion of the NRC's staff review of this issue, see SE Section 3.4.4.
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| VFDR# VFDR Description Component The Protected Service Water (pSW) Pump is required to be off to isolate PSW flow to the SGs. Fire damage to cables may result in a spurious start AB-01 of the PSW Pump. Spurious operation could result in overfill of the SGs, OPSWPU0002 - PSW Pump overcooling of the RCS and a challenge to the Decay Heat Removal Nuclear Safety Performance Criterion.
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| The Main Feedwater (MFW) Pumps are required to be off to isolate MFW to the SGs. Fire damage to cables may result in spurious pump start. 1FDWPUOOO1, AB-02 Spurious operation could result in overfill of the SGs, overcooling of the 1FDWPU0002 - Main RCS and a challenge to the Decay Heat Removal Nuclear Safety Feedwater Pumps Performance Criterion.
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| The Emergency Feedwater (EFW) Pumps are required to be off to isolate 1FDWPUOOO3, AB-03 EFW to the SGs. Fire damaqe to cables may result in spurious pump 1FDWPUOOO4, OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 234 VFDR# VFDR Description Component start. Spurious operation could result in overfill of the SGs, overcooling of 1FDWPU0005 - EFW Pumps the RCS and a challenge to the Decay Heat Removal Nuclear Safety Performance Criterion.
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| These normally closed and required closed valves isolate the flow path from the LOST [letdown storage tank] to the containment sump. Fire induced cable damage may result in spurious opening of valve 1HP 1HP VA0939, 1HP VA0940 VA0939 and/or 1HP VA0940 resulting in a diversion of BWST inventory to AB-04 LOST to Emergency Sump the containment sump, flooding of the credited SSF reactor coolant MOVs makeup (RCMU) Pump and challenge to the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria. These valves may suffer IN 92-18 damaqe.
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| The HPI Pumps are required to be off to prevent an uncontrolled increase in RC inventory. Fire damage to cables may result in spurious pump start.
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| Spurious operation could increase RC pressure to the pressurizer safety 1HPIPU0001, 1HPIPUOOO2, AB-05 relief valve set point. Subsequent failure of the HPI pump(s) and failure of 1HPIPU0003 - HPI Pumps the relief valve to reseat could result in loss of RC inventory in excess of the makeup capability of the SSF RCMU Pump. This could challenge the Inventory and Pressure Control Nuclear Safety Performance Criteria.
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| These normally open, required closed valves isolate the flow path from the BWST to the Low Pressure Injection (LPI) Pumps, RB Spray (RBS)
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| Pumps, and containment sump. Fire damage to cables may prevent these valves from being closed or may result in spurious opening of the valves.
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| 1LP VA0021, 1LP VAOO22 AB-06 The failure to close these valves or the spurious opening of the valves may BWST Suction MOVs result in a diversion of BWST inventory to the containment sump, flooding of the credited SSF RCMU Pump and challenge to the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria.
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| These valves may suffer IN 92-18 damage.
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| These normally open, required closed valves isolate flow paths from the 1MS VA0017, 1MS VAOO24, Main Steam Headers (MSHs). Fire damage to cables may result in 1MS VA0026, 1MS VAOO33, spurious opening of the valves which could result in overcooling and 1MS VA0035, 1MS VAOO36, AB-07 shrinkage of RC inventory in excess of the makeup capability of the RCMU 1MS VA0076, 1MS VAOO79, Pump. This could challenge the Decay Heat Removal Nuclear Safety 1MS VA0082, 1MS VAOO84 Performance Criterion. These valves may suffer IN 92-18 damage. SG Isolation Valves The pressurizer heaters are required to be off to prevent an uncontrolled increase in RC pressure. Fire damage to cables may result in the spurious 1RC HE0001, 1RC HEOOO2 operation of these heaters resulting in an increase in RC pressure to the (Groups 0 & K), 1RC AB-08 setpoint of the RCMU pump discharge relief valve. Diversion of RC HE0003, 1RC HE0004 makeup flow from the relief valve could degrade performance of the SSF Pressurizer Heaters RCMU Pump and challenge the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria.
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| The Reactor Coolant Pumps (RCPs) are required off when SSD is being accomplished by the SSF. Fire damage to cables may result in spurious operation of the RCPs and place the SSF in an unanalyzed condition, e.g., 1RC PU0001, 1RC PUOOO2, AB-09 add RCP heat to RCS, disrupt natural circulation flow, cause seal leakage 1RC PU0003, 1RC PU0004 -
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| in excess of the makeup capability of the SSF RCMU Pump. These RCPs conditions challenge the Inventory and Pressure Control, and Decay Heat Removal Nuclear Safety Performance Criteria.
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| These normally closed, required closed valves isolate flow paths from the RCS to containment. Fire damage to cables behind the main control boards in the control room may result in spurious opening of the above 1RC VA0155, 1RC VA0157, AB-10 valves. The spurious opening of any of the above valves may result in a 1RC VA0159 - RC Hot Leg loss of RC inventory in excess of that provided by the SSF RCMU Pump and Head Vent Valves and challenge the Inventory and Pressure Control Nuclear Safety Performance Criteria.
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| This normally closed, required closed valve isolates the flow path from the RCS to the post accident liquid sampling system. Fire damage to cables 1RC VA0179 - Post Accident may result in spurious opening of the above valve. Spurious opening of AB-11 Sample Air-Operated Valve this valve may result in a loss of RC inventory in excess of that provided by (AOY) the SSF RCMU Pump and challenge the Inventory and Pressure Control Nuclear Safety Performance Criteria.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 235 VFDR# VFDR Description Component These normally closed, required closed valves isolate flow paths from the Main Steam Headers. Fire damage to cables may result in spurious 1SD VA0027, 1SD VA0290, opening of the valves. The spurious opening of the valves could result in 1SD VA0418, 1SD VA0419, AB-12 overcooling and shrinkage of RC inventory in excess of the makeup 1SD VA0420, 1SD VA0421 -
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| capability of the RCMU Pump and challenge the Decay Heat Removal SG Isolation MOVs Nuclear Safety Performance Criterion. These valves may suffer IN 92-18 damaQe.
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| The Main Feedwater (MFW) Pumps are required to be off to isolate MFW to the SGs. Fire damage to cables may result in spurious pump start. 2FDWPUOOO1, AB-13 Spurious operation of the MFW Pumps could result in overfill of the SGs, 2FDWPU0002 - Main overcooling of the RCS and a challenge to the Decay Heat Removal Feedwater Pumps Nuclear Safety Performance Criterion.
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| The EFW Pumps are required to be off to isolate EFW to the SGs. Fire damage to cables may result in spurious pump start. Spurious operation 2FDWPUOOO3, AB-14 of the EFW Pumps could result in overfill of the SGs, overcooling of the 2FDWPUOOO4, RCS and a challenge to the Decay Heat Removal Nuclear Safety 2FDWPU0005 - EFW Pumps Performance Criterion.
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| These normally closed and required closed valves isolate the flow path from the LDST to the containment sump. Fire induced cable damage may result in spurious opening of valve 2HP VA0939 and/or 2HP VA0940 2HP VA0939, 2HP VA0940 AB-15 resulting in a diversion of BWST inventory to the containment sump, LDST to Emergency Sump flooding of the credited SSF RCMU Pump and challenge to the Reactivity, MOVs Inventory and Pressure Control Nuclear Safety Performance Criteria.
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| These valves may suffer IN 92-18 damaQe.
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| The HPI Pumps are required to be off to prevent an uncontrolled increase in RC inventory. Fire damage to cables may result in spurious pump start.
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| Spurious operation of HPI pump(s) could increase RC pressure to the pressurizer safety relief valve set point. Subsequent failure of the HPI 2HPIPU0001, 2HPIPUOOO2, AB-16 pump(s) and failure of the relief valve to reseat could result in loss of RC 2HPIPU0003 - HPI Pumps inventory in excess of the makeup capability of the SSF RCMU Pump.
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| This could challenge the Inventory and Pressure Control Nuclear Safety Performance Criteria.
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| These normally open, required closed valves isolate the flow path from the BWST to the LPI Pumps, RBS Pumps, and containment sump. Fire damage to cables may prevent these valves from being closed or may result in spurious opening of the valves. The failure to close these valves 2LP VA0021, 2LP VAOO22 AB-17 or the spurious opening of the valves may result in a diversion of BWST BWST Suction MOVs inventory to the containment sump, flooding of the credited SSF RCMU Pump and challenge to the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria. These valves may suffer IN 92-18 damage.
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| These normally open, required closed valves isolate flow paths from the 2MS VA0017, 2MS VAOO24, Main Steam Headers. Fire damage to cables may result in spurious 2MS VA0026, 2MS VA0033, opening of the valves which could result in overcooling and shrinkage of 2MS VA0035, 2MS VAOO36, AB-18 RC inventory in excess of the makeup capability of the RCMU Pump. This 2MS VA0076, 2MS VAOO79, could challenge the Decay Heat Removal Nuclear Safety Performance 2MS VA0082, 2MS VAOO84 Criterion. These valves may suffer IN 92-18 damaQe. SG Isolation Valves The pressurizer heaters are required to be off to prevent an uncontrolled increase in RC pressure. Fire damage to cables may result in the spurious 2RC HE0001, 2RC HEOOO2 operation of these heaters resulting in an increase in RC pressure to the (Groups D & K), 2RC AB-19 setpoint of the RCMU pump discharge relief valve. Diversion of RC HE0003, 2RC HEOOO4 makeup flow from the relief valve could degrade performance of the SSF Pressurizer Heaters RCMU Pump and challenge the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria.
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| The Reactor Coolant Pumps (RCPs) are required off when SSD is being accomplished by the SSF. Fire damage to cables may result in spurious 2RC PU0001, 2RC PUOOO2, operation of the RCPs and place the SSF in an unanalyzed condition, e.g.,
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| AB-20 2RC PU0003, 2RC PU0004 -
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| add RCP heat to RCS, disrupt natural circulation flow, cause seal leakage RCPs in excess of the makeup capability of the SSF RCMU Pump. These conditions challenge the Inventory and Pressure Control, and Decav Heat OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 236 VFDR# VFDR Description Component Removal Nuclear Safety Performance Criteria.
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| These normally closed, required closed valves isolate flow paths from the RCS to containment. Fire damage to cables may result in spurious 2RC VA0155, 2RC VA0157, opening of the above valves. The spurious opening of any of the above AB-21 2RC VA0159 - RC Hot Leg valves may result in a loss of RC inventory in excess of that provided by and Head Vent Valves the SSF RCMU Pump and challenge the Inventory and Pressure Control Nuclear Safety Performance Criteria.
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| This normally closed, required closed valve isolates the flow path from the RCS to the post accident liquid sampling system. Fire damage to cables may result in spurious opening of the above valve. Spurious opening of 2RC VA0179 - Post Accident AB-22 this valve would result in a loss of RC inventory in excess of that provided SampleAOV by the SSF RCMU Pump and challenge the Inventory and Pressure Control Nuclear Safety Performance Criteria These normally closed, required closed valves isolate flow paths from the Main Steam Headers. Fire damage to cables may result in spurious 2SD VA0027, 2SD VA0290, opening of the valves. The spurious opening of the valves could result in 2SD VA0418, 2SD VA0419, AB-23 overcooling and shrinkage of RC inventory in excess of the makeup 2SD VA0420, 2SD VA0421 -
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| capability of the RCMU Pump and challenge the Decay Heat Removal SG Isolation MOVs Nuclear Safety Performance Criterion. These valves may suffer IN 92-18 damage.
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| The Main Feedwater (MFW) Pumps are required to be off to isolate MFW to the SGs. Fire damage to cables may result in spurious pump start. 3FDWPUOOO1, AB-24 Spurious operation of the MFW Pumps could result in overfill of the SGs, 3FDWPU0002 - Main overcooling of the RCS and a challenge to the Decay Heat Removal Feedwater Pumps Nuclear Safety Performance Criterion.
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| The EFW Pumps are required to be off to isolate EFW to the SGs. Fire damage to cables may result in spurious pump start. Spurious operation 3FDWPUOOO3, AB-25 of the EFW Pumps could result in overfill of the SGs, overcooling of the 3FDWPUOOO4, RCS and a challenge to the Decay Heat Removal Nuclear Safety 3FDWPU0005 - EFW Pumps Performance Criterion.
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| These normally closed and required closed valves isolate the flow path from the LOST to the containment sump. Fire induced cable damage may result in spurious opening of valve 3HPVA0939 and/or 3HP VA0940 3HP VA0939, 3HP VA0940 AB-26 resulting in a diversion of BWST inventory to the containment sump, LOST to Emergency Sump flooding of the credited SSF RCMU Pump and challenge to the Reactivity, MOVs Inventory and Pressure Control Nuclear Safety Performance Criteria.
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| These valves may suffer IN 92-18 damaQe.
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| The HPI Pumps are required to be off to prevent an uncontrolled increase in RC inventory. Fire damage to cables may result in spurious pump start.
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| Spurious operation of HPI pump(s) could increase RC pressure to the pressurizer safety relief valve set point. Subsequent failure of the HPI 3HPIPU0001, 3HPIPUOOO2, AB-27 pump(s) and failure of the relief valve to reseat could result in loss of RC 3HPIPU0003 - HPI Pumps inventory in excess of the makeup capability of the SSF RCMU Pump.
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| This could challenge the Inventory and Pressure Control Nuclear Safety Performance Criteria.
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| These normally open, required closed valves isolate the flow path from the BWST to the LPI Pumps, RBS Pumps, and containment sump. Fire damage to cables may prevent these valves from being closed or may result in spurious opening of the valves. The failure to close these valves 3LP VA0021, 3LP VAOO22 AB-28 or the spurious opening of the valves may result in a diversion of BWST BWST Suction MOVs inventory to the containment sump, flooding of the credited SSF RCMU Pump and challenge to the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria. These valves may suffer IN 92-18 damaqe.
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| These normally open, required closed valves isolate flow paths from the 3MS VA0017, 3MS VAOO24, Main Steam Headers. Fire damage to cables may result in spurious 3MS VA0026, 3MS VAOO33, opening of the valves which could result in overcooling and shrinkage of 3MS VA0035, 3MS VAOO36, AB-29 RC inventory in excess of the makeup capability of the RCMU Pump. This 3MS VA0076, 3MS VAOO79, could challenge the Decay Heat Removal Nuclear Safety Performance 3MS VA0082, 3MS VAOO84 Criterion. These valves may suffer IN 92-18 damaqe. SG Isolation Valves OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| Of:=f:=ICIAL USE ONLY SECURITY RELATED INf:=ORMATION 237 VFDR# VFDR Description Component The pressurizer heaters are required to be off to prevent an uncontrolled increase in RC pressure. Fire damage to cables may result in the spurious operation of these heaters resulting in an increase in RC pressure to the 3RC HE0001, 3RC HEOO02 setpoint of the RCMU pump discharge relief valve. Diversion of RC (Groups D & K), 3RC AB-30 makeup flow from the relief valve could degrade performance of the SSF HE0003, 3RC HE0004 RCMU Pump and challenge the Reactivity, Inventory and Pressure Control Pressurizer Heaters Nuclear Safety Performance Criteria.
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| The Reactor Coolant Pumps (RCPs) are required off when SSD is being accomplished by the SSF. Fire damage to cables may result in spurious operation of the RCPs and place the SSF in an unanalyzed condition, e.g., 3RC PU0001, 3RC PUOO02, AB-31 add RCP heat to RCS, disrupt natural circulation flow, cause seal leakage 3RC PU0003, 3RC PU0004 -
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| in excess of the makeup capability of the SSF RCMU Pump. These RCPs conditions challenge the Inventory and Pressure Control, and Decay Heat Removal Nuclear Safety Performance Criteria These normally closed, required closed valves isolate flow paths from the RCS to containment. Fire damage to cables may result in spurious 3RC VA0155, 3RC VA0157, opening of the above valves. The spurious opening of any of the above AB-32 3RC VA0159 - RC Hot Leg valves may result in a loss of RC inventory in excess of that provided by and Head Vent Valves the SSF RCMU Pump and challenge the Inventory and Pressure Control Nuclear Safety Performance Criteria.
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| These normally closed, required closed valves isolate flow paths from the Main Steam Headers. Fire damage to cables may result in spurious 3SD VA0027, 3SD VA0418, opening of the valves. The spurious opening of the valves could result in 3SD VA0419, 3SD VA0420, AB-33 overcooling and shrinkage of RC inventory in excess of the makeup 3SD VA0421 - SG Isolation capability of the RCMU Pump and challenge the Decay Heat Removal MOVs Nuclear Safety Performance Criterion. These valves may suffer IN 92-18 damaQe.
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| The wall separating the TB and AB is not three hour rated as required by NFPA 805, Section 3.11.1 and all the penetrations in the wall do not have AB-34 a fire resistance rating as required by NFPA 805, Section 3.11.3. This wall TB / ABWali is credited for area separation in the deterministic approach of NFPA 805, Section 4.2.3.
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| The wall separating the AB and the West penetration room does not have a fire-resistance rating required by NFPA 805, Section 3.11.2 and all the AB / West Penetration Room AB-35 penetrations in the wall do not have a fire resistance rating as required by Separation NFPA 805, Section 3.11.3. This wall is credited for area separation in the deterministic approach of NFPA 805, Section 4.2.3.
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| This normally closed, required closed valve isolates a flow path from the RCS to containment. Fire damage to cables in the penetration box in the East Penetration Room may result in spurious opening of the above valve. 1RC VA0155 - RC Hot Leg AB-36 The spurious opening of the valve may result in a loss of RC inventory in Vent Valve excess of that provided by the SSF RCMU Pump and challenge the Inventorv and Pressure Control Nuclear Safety Performance Criteria.
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| These normally closed, required closed valves isolate flow paths from the RCS to containment. Fire damage to cables in the penetration box in the East Penetration Room may result in spurious opening of these valves. 2RC VA0155, 2RC VA0157 AB-37 The spurious opening of the valves may result in a loss of RC inventory in RC Hot Leg Vent Valves excess of that provided by the SSF RCMU Pump and challenge the
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| . Inventory and Pressure Control Nuclear Safety Performance Criteria.
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| This normally closed, required closed valve isolates a flow path from the RCS to containment. Fire damage to cables in the penetration box in the East Penetration Room may result in spurious opening of the above valve. 3RC VA0155 - RC Hot Leg AB-38 The spurious opening of the valve may result in a loss of RC inventory in Vent Valve excess of that provided by the SSF RCMU Pump and challenge the Inventory and Pressure Control Nuclear Safety Performance Criteria.
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| The monitoring and/or adjustment of the following parameters is required during operation of the SSF DG; generator current, voltage, power and AB-39 SSF DG frequency. The controls and indications required to monitor and adjust these parameters are currently not included in the SSD analysis for those Of:=f:=ICIAL USE O~JLY SECURITY RELATED INf:=ORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 238 VFDR# VFDR Description Component fire areas where the SSF is credited for accomplishin~ SSD.
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| The areas separating the Unit 1 Purge Inlet Room and SFP area is not three hour rated as required by NFPA 805, Section 3.11.1 and the AB-40 penetrations (seals and doors) do not have a fire resistance rating as Purge Inlet Room / SFP Area required by NFPA 805, Section 3.11.3. These barriers are credited for area separation in the deterministic approach of NFPA 805, Section 4.2.3.
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| The areas separating the Unit 2 Purge Inlet Room and SFP area is not three hour rated as required by NFPA 805, Section 3.11.1 and the AB-41 penetrations (seals and doors) do not have a fire resistance rating as Purge Inlet Room / SFP Area required by NFPA 805, Section 3.11.3. These barriers are credited for area separation in the deterministic approach of NFPA 805, Section 4.2.3.
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| The areas separating the Unit 3 Purge Inlet Room and SFP area is not three hour rated as required by NFPA 805, Section 3.11.1 and the AB-42 penetrations (seals and doors) do not have a fire resistance rating as Purge Inlet Room / SFP Area required by NFPA 805, Section 3.11.3. These barriers are credited for area separation in the deterministic approach of NFPA 805, Section 4.2.3.
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| In supplementary information (Reference 52), the licensee identified that this normally open, required closed valve isolates flow path from the Pressurizer upon spurious opening of the PORV. Fire damage to cables may result in spurious opening of the valve. The spurious opening of the 1RC VA0004, RC PORV AB-43 valve may result in a loss of RC inventory in excess of that provided by the Motor Operated Block Valve SSF RCMU Pump and challenge the Inventory and Pressure Control Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 dama~e.
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| In supplementary information (Reference 52), the licensee identified that these normally open, required closed valves isolate the flow path from the RCS. Fire damage to cables may prevent these valves from being closed 1HP VA0003 and 1HP or may result in spurious opening of the valves. The failure to close these AB-44 VA0004 1A and 1B Letdown valves or the spurious opening of the valves may result in a diversion of Cooler Outlet Valves RCS inventory and challenge to the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria. These valves may suffer IN 92-18 damaqe.
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| In supplementary information (Reference 52), the licensee identified that this normally open, required closed valve isolates the flow path from the RCS. Fire damage to cables may prevent this valve from being closed or may result in spurious opening of the valve. The failure to close this valve -IHP VA0020 RCP Seal AB-45 or the spurious opening of the valve may result in a diversion of RCS Return Valve inventory and challenge to the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 damaqe.
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| In supplementary information (Reference 52), the licensee identified that this normally open, required closed valve isolates flow path from the Pressurizer upon spurious opening of the PORV. Fire damage to cables may result in spurious opening of the valve. The spurious opening of the 2RC VA0004, RC PORV AB-46 valve may result in a loss of RC inventory in excess of that provided by the Motor Operated Block Valve SSF RCMU Pump and challenge the Inventory and Pressure Control Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 dama~e.
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| In supplementary information (Reference 52), the licensee identified that these normally open, required closed valves isolate the flow path from the RCS. Fire damage to cables may prevent these valves from being closed 2HP VA0003 and 2HP or may result in spurious opening of the valves. The failure to close these AB-47 VA0004 2A and 2B Letdown valves or the spurious opening of the valves may result in a diversion of Cooler Outlet Valves RCS inventory and challenge to the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria. These valves may suffer IN 92-18 damaqe.
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| In supplementary information (Reference 52), the licensee identified that this normally open, required closed valve isolates the flow path from the 2HP VA0020 RCP Seal AB-48 RCS. Fire damage to cables may prevent this valve from being closed or Return Valve may result in spurious opening of the valve. The failure to close this valve OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 239 VFDR# VFDR Description Component or the spurious opening of the valve may result in a diversion of RCS inventory and challenge to the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 damage.
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| In supplementary information (Reference 52), the licensee identified that this normally open, required closed valve isolates flow path from the Pressurizer upon spurious opening of the PORV. Fire damage to cables may result in spurious opening of the valve. The spurious opening of the 3RC VA0004, RC PORV AB-49 valve may result in a loss of RC inventory in excess of that provided by the Motor Operated Block Valve SSF RCMU Pump and challenge the Inventory and Pressure Control Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 damaae.
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| In supplementary information (Reference 52), the licensee identified that these normally open, required closed valves isolate the flow path from the RCS. Fire damage to cables may prevent these valves from being closed 3HP VA0003 and 3HP or may result in spurious opening of the valves. The failure to close these AB-50 VA0004 1A and 1B Letdown valves or the spurious opening of the valves may result in a diversion of Cooler Outlet Valves RCS inventory and challenge to the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria. These valves may suffer IN 92-18 damaQe.
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| In supplementary information (Reference 52), the licensee identified that this normally open, required closed valve isolates the flow path from the RCS. Fire damage to cables may prevent this valve from being closed or may result in spurious opening of the valve. The failure to close this valve 3HP VA0020 RCP Seal AB-51 or the spurious opening of the valve may result in a diversion of RCS Return Valve inventory and challenge to the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 damaqe.
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| In supplementary information (Reference 52), the licensee identified that the retrieval, assembly, and water body deployment of the portable submersible pump including necessary hose(s) and electrical power are not predominantly conducted in the SSF or deployed during the initial transfer of control from the control room. The deployment and operation of the SSF submersible pump is only credited for floods and is not currently AB-52 credited for any fire scenarios. In a fire scenario, flow is maintained to the SSF Submersible Pump Condenser Circulating Water (CCW) piping providing that either the CCW pumps or the Essential Siphon Vacuum (ESV) provides flow. Thus, if modeled, the delta risk associated with the failure to deploy the submersible pump is expected to be epsilon (E) for fire events. This conclusion is further substantiated by insights gained from the internal events PRA and the associated expert panel reviews.
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| Note: The additional risk added because of these VFDRs, as determined from the FRE for this fire area, is provided in SE Table 3.5.
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| Recovery Actions (RAs)
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| By letter dated November 19, 2010 (Reference 52), the licensee identified one recovery action that is credited in this fire area for meeting the nuclear safety performance criteria, and is provided in the following table:
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| Component Component Name Description of Action 10 The deployment and operation of the SSF submersible pump is only credited for floods and is not currently credited for any fire SSF scenarios. In a fire scenario, flow is maintained to the Submersible SSF Submersible Pump Condenser Circulating Water (CCW) piping providing that Pump either the CCW pumps or the Essential Siphon Vacuum (ESV) provides flow. The submersible pump would onlv be required OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 240 in the fire scenario under the following 'set of conditions: CCW flow is insufficient only if the CCW pumps are not running, the lake level is too low to support backflow through the condensate coolers, and the ESV systems are unavailable to maintain adequate siphon.
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| Note: A bounding assessment of the additional risk being added because of this RA was determined by the NRC staff to be sufficiently small that the risk acceptance guidelines associated with pre-approved recovery actions have all been met. See SE Section 3.4.4 for a detailed discussion of the NRC staff's review of the estimated risk for this RA.
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| Fire Detection & Suppression Systems Required to Meet the Nuclear Safety Performance Criteria The licensee performed an evaluation of the fire detection and suppression systems in this area.
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| The results of the evaluation were documented in LAR Table 4-4, applicable portions included below. The identified fire detection system modifications are to improve plant fire detection and fire brigade response time.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 241 Suppression Required Detection Required Auto System? System?
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| Fire Fire Detection Zone Description Suppression Area Zone Provided?
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| Provided?
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| E R D S E R D S AB Auxiliary Buildina 48 Unit 3 LPI & RBSP No No No No No Yes No Yes No No 49 Unit 3 LPI & RBSP No No No No No Yes No Yes No No 50 Unit 3 HPI Pump Area No No No No No Yes No Yes No No Unit 3 HPI Pump, Spt Resin Xfr Pump 50A No No No No No Yes No Yes No No Waste Tank, Waste & CT Drain Pumps Unit 3 Purification & Deboration 51 No No No No No No No No No No Deminerilizers Unit 2 LPI Pumps & Valve Room (Inside 52 No No No No No Yes No Yes No No Room 63) 53 Units 1 & 2 LPI Pumps & RBSP No No No No No Yes No Yes No No Unit 1 LPI Pumps & Valve Room (Inside 54 No No No No No Yes No Yes No No Room 61)
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| Unit 1 RB Sump & Cmp Drain Pmp, HPI 55 No No No No No Yes No Yes No No Pmp, Spt Res Transfer 55A Units 1 & 2 HPI Pump Area No No No No No Yes No Yes No No Unit 2 Spt Res Transfer Pmp, HPI Pmp, 56 No No No No No Yes No Yes No No RB Sump & Cmp Drain Units 1 & 2 Purification & Deboration 57 No No No No No No No No No No Deminerilizers None Unit 3 Boric Acid Mix, Spt Res Storage, 58 No No No No No Yes No Yes No No RC BHUT CBAST, Misc Unit 3 Decay Heat Removal Coolers, Seal 59 No No No No No No No No No No Supply Filter/Pipe 60 Unit 3 LPI Room Hatch Area No No No No No Yes No Yes No No Yes 61 Unit 3 HPI Room Hatch Area No No No No No No No No No (MR)
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| Unit 3 Operators Panel/Chemical Sample 62 No No No No No Yes No Yes No No Hood 63 Unit 3 LDST, LD Filters, LD Filter Hatch No No No No No No No No No No 64 Unit 2 Emeraency Aux SW Pump No No No ,No No Yes No Yes No No Unit 2 MWHT, Misc Waste Evaporator, 65 No No No No No Yes No Yes No No CBAST, RC Bleed Xfer Pmp, Unit 2 Decay Heat Removal Coolers, Seal 66 No No No No No No No No No No Supply Filter/Pipe 67 Unit 2 LPI Room Hatch Area No No No No No Yes No Yes No No 68 Unit 2 HPI Room Hatch Area No No No No No No No Yes No No OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 242 Suppression Required Detection Required Auto System? System?
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| Fire Fire Detection Zone Description Suppression Area Zone Provided?
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| Provided?
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| E R D S E R D S (MR)
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| Unit 2 Operators Panel/Chemical Sample No 69 No No No No No Yes No Yes No Hood 70 Unit 1 LPI Room Hatch Area No No No No No Yes No Yes No No 71 Unit 2 LOST, LD Filters, LD Filter Hatch No No No No No No No No No No Yes 72 Unit 1 HPI Room Hatch Area No No No No No No No No No (MR) 73 Unit 1 LOST, LD Filters, LD Filter Hatch No No No No No No No No No No Units 1 & 2 Dress out Area for Units 1 & 2 74 No No No No No No No No No No HPI Hatch Areas Unit 1 Pipe Rooms, Seal Supply 75 No No No No No No No No No No Filter/Pipe Room Unit 1 RC HU Tanks, CBAST, RC Bleed 76 No No No No No Yes No Yes No No Xfr Pmp, Wst Dma, Fltr Room, SRST Yes 77 Unit 3 Storage, Chemistry Storage No No No No No No No No No (MR)
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| Unit 3 Spent Fuel Cooler Filters/Demin, 78 No No No No No No No No No No Spent Fuel Coolers 79 Unit 3 RB Component Coolers No No No No No Yes No Yes No No Unit 3 Waste Gas Decay Tanks, Waste 80 No No No No No No No No No No Gas Comp Room Unit 2 I&E Hot Shop, Misc Evaporator 81 No No No No No Yes No Yes No No Feedwater Tank, Chemical Storaae, Units 1 & 2 Spent Fuel Coolers, Spent 82 No No No No No No No No No No Fuel Cooler Filter/Demin 83 Units 1 & 2 RB Component Coolers No No No No No Yes No Yes No No Units 1 & 2 Waste Gas Decay Tanks, 84 No No No No No No No No No No Waste Gas Comp Unit 1 Chemistry Storage, High Level 85 No No No No No Yes No Yes No No Storaqe Unit 3 Hatch Area Chemistry Labs &
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| 86 No No No No No Yes No Yes No No Chanqe Room Unit 3 Fuel Receiving Area, SFP &
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| 88 No No No No No No No No No No Loadinq Area 89 Unit 3 Eauipment Room No No No No No Yes No Yes No No Unit 2 Hallway, Change Room, Laundry 90 No No No No No Yes No Yes No No Room, RP Lab, Chemistry Laboratory, OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 243 Suppression Required Detection Required Auto System? System?
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| Fire Fire Detection Zone Description Suppression Area Zone Provided?
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| Provided?
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| E R D S E R D S Medical Room, and Decontamination (DECON) Room 92 Unit 2 Eauipment Room No No No No No Yes No Yes No No Units 1 & 2 Fuel Receiving Area, SFP &
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| 93 No No No No No No No No No No Loadina Area Unit 1 Hallway, Hatch Area, Change Yes 94 No No No No No No No No No Room, and Tool (MR) 95 Unit 1 Eauipment Room No No No No No Yes No Yes No No Unit 2 Hot Machine Shop Tunnel, Hot 96 No No No No No No No No No No Machine Shop Yes 99 Unit 3 East Penetration Room No No No No No No No No No (MR) 100 Unit 3 Control Battery Room No No No No No Yes No Yes No No Ye 101 Unit 3 Cable Room Manual No No No Yes Yes Yes No No s
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| Yes 103 Unit 2 East Penetration Room No No No No No Yes No No No (MR) 104 Unit 2 Control Battery Room No No No No No Yes No Yes No No Ye 105 Unit 2 Cable Room Yes No No No Yes Yes Yes No No s
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| Ye 106 Unit 1 Cable Room Manual No No No Yes Yes Yes No No s
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| Yes 108 Unit 1 East Penetration Room No No No No No No No No No (MR) 109 Unit 1 Control Battery Room No No No No No Yes No Yes No No 109A Unit 1 Control Room Lobby/AHU Room No No No No No No No No No No Yes 110 Units 1 & 2 Control Room No No No No No Yes No No No (MR) 111 Unit 2 Control Room Lobby/AHU Room No No No No No No No No No No Yes 112 Unit 3 Control Room No No No No No Yes No No No (MR) 113 Unit 3 Control Room Lobby/AHU Room No No No No No No No No No No Yes 115 Unit 3 Main Purge Exhaust Room No No No No No No No No No (MR) 116 Unit 3 AHU Room R No No No No No Yes No Yes No No OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 244 Suppression Required Detection Required Auto System? System?
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| Fire Fire Detection Zone Description Suppression Area Zone Provided?
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| Provided?
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| E R D S E R D S Yes 118 Unit 2 Main Purge Exhaust Room No No No No No No No No No (MR) 119 Units 1 & 2 AHU Room No No No No No Yes No Yes No No Yes 121 Unit 1 Main Purge Exhaust Room No No No No No No No No No (MR)
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| Legend:
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| E - EEEE/LA: Systems required for acceptability of EEE Evaluations I NRC-approved Licensing Action (Section 2.2.7)
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| R - Risk: Systems required to meet the Risk Criteria for the PB Approach (Section 4.2.4)
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| D - Defense-in-Depth: Systems required to maintain adequate balance of Defense-in-Depth for a PB Approach (Section 4.2.4.2)
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| S - Separation Criteria: Systems required for NFPA 805, Chapter 4 Separation Criteria in (Section 4.2.3)
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| MR - Modification Required Systems are committed to be modified as indicated in Table 4-4 and Attachment S of LAR OFFICIAL USE O~JLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELl\TED INFORMATION 245 Fire Area AB Conclusion The licensee has utilized the FRE PB approach to demonstrate the ability to meet the I\IFPA 805 nuclear safety performance criteria for this fire area. A FRE in accordance with NFPA 805, Section 4.2.4.2, in conjunction with deterministic methods for simplifying assumptions, was used in applying this approach.
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| Based on the information provided in the LAR, as supplemented, the NRC staff finds Fire Area AB meets the nuclear safety goals, objectives, and performance criteria of NFPA 805. This conclusion is based on the following:
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| * Fire protection SSCs were evaluated in accordance with NFPA 805, Chapter 4, to determine which, if any, were required to meet the nuclear safety performance criteria.
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| This evaluation included:
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| On a fire zone basis, the fire protection detection systems required to meet the nuclear safety performance criteria were documented.
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| Fire Area boundaries were defined using three hour rated walls, ceilings and floors, including fire barriers, fire barrier penetrations and through penetration fire stops and spatial separation.
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| * Three exemptions and one other licensing action from the pre-transition fire protection requirements were evaluated and found to be valid and applicable under the NFPA 805 RI/PB FPP.
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| * Forty-two VFDRs were identified, evaluated through the performance of a FRE, and either found to meet the risk acceptance criteria, as well as the requirements for DID and SMs, or modifications were planned to address the issue. The acceptability of the risk for this fire area is contingent on the risk reduction from the planned PSW modification (see SE Section 3.4 for a detailed discussion of the NRC staff's review of the adequacy of the FRE method used at ONS).
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| * This fire area did not require the use of recovery actions to meet the nuclear safety performance criteria.
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| * The following modifications were identified to address VFDRs:
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| Fire detection upgrades and/or new installation in thirteen (13) fire zones to improve plant fire detection and fire brigade response time.
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| Fire barriers upgrades to provide three hour fire rated separation, as follows:
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| o Purge Inlet Rooms and Spent Fuel Poll Area for Units 1,2, and 3 [AB-40, AB-41 ,
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| and AB-42]
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| o AB / TB [AB-34]
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| o AB / West Penetration Room [AB-35]
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| Fire Area BH12, Units 1 and 2 Block House The licensee analyzed this fire area using the FRE approach in accordance with NFPA 805, Section 4.2.4.2, but also applied deterministic simplifying assumptions in order to credit those portions of the facility design that met the deterministic requirements of NFPA 805, Section OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 246 4.2.3. The licensee identified the SSCs necessary to meet the nuclear safety performance criteria in this fire area.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 247 Evaluation of Fire Suppression Effects on Nuclear Safety Performance Criteria The licensee stated in LAR Attachment C, "NEI 04-02, Table B-3, Fire Area Transition," that safe and stable conditions can be achieved and maintained using equipment and cables outside of the area of fire suppression activity. Flooding of the suppression areas and discharge of suppression water to adjacent compartments is controlled and will not jeopardize achievement of safe and stable conditions. Based on the information provided by the licensee in the NFPA 805 LAR, the NRC staff finds the licensee's evaluation of fire suppression effects on nuclear safety performance criteria acceptable because the results of the licensee's analysis indicate that fire suppression activities will not adversely affect achievement of the nuclear safety performance criteria.
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| Exemptions and Other Licensing Actions The licensee did not credit any previously approved licensing actions or exemptions from the existing fire protection requirements.
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| Variation from Deterministic Requirements (VFDRs)
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| Fire Area BH12 has a total of 15 VFDRs, which are provided in the table below. All but one of these VFDRs are variances from NFPA 805 Section 4.2.3 (separation issues) that were dispositioned with a FRE (SE Section 3.4.3). The licensee's FRE determined that these variances are acceptable based on 1) the change in CDF and LERF for the fire area and the total CDF and LERF for each unit meet the acceptance criteria of RG 1.174 (SE Section 3.4.6) and 2) adequate DID and SMs are maintained for each fire area (SE Section 3.4.2). This determination relies on the following fire protection systems and features to meet the acceptance criteria:
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| * General area and/or fire hazard detection for the fire area BH 12 is required to meet the DID criteria.
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| Based on the reliance on fire detectors in fire area BH12 to meet the DID criteria, the licensee has committed to make modifications to the fire detection system, which may include fire detector upgrades and/or new installation (SE Section 2.8.1).
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| One of the 15 VFDRs, BH12-02, is a variance from NFPA 805, Section 4.2.3 (degraded fire protection feature) that will be corrected with a plant modification. According to the LAR, the wall separating fire area BH 12 from the fire area YARD does not currently have a three hour rated wall as required by NFPA 805, Section 3.11.1, and all of the penetrations in the wall do not have a three hour fire resistance rating as required by NFPA 805, Section 3.11.3. This wall is credited for area separation using the deterministic approach of NFPA 805, Section 4.2.3.
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| The licensee has committed to make a modification to install hinged steel covers/shields to the exterior side of the tornado vents that will qualify the wall 'adequate for the hazard' (SE Section 2.8.1).
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 248 VFDR# VFDR Description Component (Cables)
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| Normally open valve 1HP VA0023 is in the flow path from the LOST to the suction of the credited HPI pump. Normally closed valve 1HP VA0939 isolates the flow path from the LOST to the containment sump. Recirculation flow to the LOST 1HP VA0023 - HPI during prolonged operation of the HPI pump at low flow conditions may result in an increase in temperature of LOST Normal Suction MaV, BH12-01 contents to the operability limit of the HPI pump. The contents of the LOST must be diverted to the containment sump by 1HP VA0939 - LOST to opening 1HP VA0939 and closing 1HP VA0023 prior to the operability limit of the HPI pump being exceeded to prevent Emergency Sump challenging the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria. Fire damage to cables MaV for electrical equipment supplvinq power to these valves mav prevent the valves from being operated.
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| The penetrations in the wall interfacing the east wall of Blockhouse 1/2 and the east yard do not have a fire resistance Tornado Vents in BH12-02 rating as required by NFPA 805, Section 3.11.3. This wall is credited for area separation in the deterministic approach of Blockhouse 1/2 NFPA 805, Section 4.2.3. Buildinq Wall 1MS VA0017, 1MS VA0024, 1MS VA0026, These normally open, required closed valves isolate flow paths from the Main Steam Headers. Fire damage to cables 1MS VA0033, 1MS for electrical equipment supplying power to these valves may prevent these valves from being closed and could result in VA0035, 1MS VA0036, BH12-03 overcooling and shrinkage of RC inventory. This could challenge the Oecay Heat Removal Nuclear Safety Performance 1MS VA0076, 1MS Criterion. VA0079, 1MS VA0082, 1MS VA0084 - SG Isolation MaVs 1RC SXTRN001, Pressurizer heaters are required for RC pressure control. The heaters receive non-credited power from Unit 1 and 1RC SXTRN002 credited power from the PSW system power supply. The transfer of credited power to the pressurizer heaters requires a BH12-04 Pressurizer Heaters recovery action. Failure to transfer credited power to the heaters could challenge the Pressure Control Nuclear Safety PSW Power Transfer Performance Criterion Switches Normally open valve 2HP VA0023 is in the flow path from the LOST to the suction of the credited HPI pump. Normally closed valve 2HP VA0939 isolates the flow path from the LOST to the containment sump. Recirculation flow to the LOST 2HP VA0023 during prolonged operation of the HPI pump at low flow conditions may result in an increase in temperature of LOST HPJ Normal Suction BH12-05 contents to the operability limit of the HPI pump. The contents of the LOST must be diverted to the containment sump by MaV, 2HP VA0939 opening 2HP VA0939 and closing 2HP VA0023 prior to the operability limit of the HPI pump being exceeded to prevent LOST to Emergency challenging the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria. Fire damage to cables Sump MaV for electrical equipment sUPPlving power to these valves may prevent the valves from beinQ operated.
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| 2MS VA0017, 2MS VA0024, 2MS VA0026, These normally open, required closed valves isolate flow paths from the Main Steam Headers. Fire damage to 'cables 2MS VA0033, 2MS for electrical equipment supplying power to these valves may prevent these valves from being closed and could result in VA0035, 2MS VA0036, BH12-07 overcooling and shrinkage of RC inventory. This could challenge the Oecay Heat Removal Nuclear Safety Performance 2MS VA0076, 2MS Criterion. VA0079, 2MS VA0082, 2MS VA0084 - SG Isolation MaVs 2RC SXTRN001, 2RC Pressurizer heaters are required for RC pressure control. The heaters receive non-credited power from Unit 2 and SXTRN002, 2RC credited power from the PSW system power supply. The transfer of credited power to the pressurizer heaters requires a BH12-08 SXTRN003 recovery action. Failure to transfer credited power to the heaters could challenge the Pressure Control Nuclear Safety Pressurizer Heaters Performance Criterion.
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| PSW Power Transfer OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 249 VFDR# VFDR Description Component (Cables)
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| Switches Normally open valve 3HP VA0023 is in the flow path from the LOST to the suction of the credited HPI pump. Normally closed valve 3HP VA0939 isolates the flow path from the LOST to the containment sump. Recirculation flow to the LOST 3HP VA0023 - HPI during prolonged operation of the HPI pump at low flow conditions may result in an increase in temperature of LOST Normal Suction MaV, BH12-09 contents to the operability limit of the HPI pump. The contents of the LOST must be diverted to the containment sump by 3HP VA0939 - LOST to opening 3HP VA0939 and closing 3HP VA0023 prior to the operability limit of the HPI pump being exceeded to prevent Emergency Sump challenging the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria. Fire damage to cables MaV for electrical equipment supplyinq power to these valves may prevent the valves from beinq operated.
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| 3MS VA0017, 3MS VA0024, 3MS VAOO26, These normally open, required closed valves isolate flow paths from the Main Steam Headers. Fire damage to cables 3MS VA0033, 3MS for electrical equipment supplying power to these valves may prevent these valves from being closed and could result in VA0035, 3MS VAOO36, BH12-11 overcooling and shrinkage of RC inventory. This could challenge the Oecay Heat Removal Nuclear Safety Performance 3MS VA0076, 3MS Criterion. VA0079, 3MS VA0082, 3MS VA0084 - SG Isolation MaVs 3RC SXTRN001, 3RC Pressurizer heaters are required for RC pressure control. The heaters receive non-credited power from Unit 3 and SXTRN002, 3RC credited power from a PSW system power supply. The transfer of credited power to the pressurizer heaters requires a SXTRN003 BH12-12 recovery action. Failure to transfer credited power to the heaters could challenge the Pressure Control Nuclear Safety Pressurizer Heaters Performance Criterion. PSW Power Transfer Switches Fire damage to cables for electrical equipment supplying power to the station HVAC system may result in the Units 1 & 2 Control BH12-14 temperature inside the Units 1 & 2 control complex exceeding the operability limit of SSO components and challenge the Complex Cooling Vital Auxiliaries Nuclear Safety Performance Criterion.
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| Fire damage to cables for electrical equipment supplying power to the station HVAC system may result in the Unit 3 Control Complex BH12-15 temperature inside the Unit 3 control complex exceeding the operability limit of SSO components and challenge the Vital Cooling Auxiliaries Nuclear Safety Performance Criterion.
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| Fire damage to cables for electrical equipment supplying power to the containment cooling system may result in the Unit 1 Containment BH12-16 temperature inside the Unit 1 RB exceeding the operability limit of SSO components and challenge the Vital Auxiliaries Cooling Nuclear Safety Performance Criterion.
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| Fire damage to cables for electrical equipment supplying power to the containment cooling system may result in the Unit 2 Containment BH12-17 temperature inside the Unit 2 RB exceeding the operability limit of SSO components and challenge the Vital Auxiliaries Cooling Nuclear Safety Performance Criterion.
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| Fire damage to cables for electrical equipment supplying power to the containment cooling system may result in the Unit 3 Containment BH12-18 temperature inside the Unit 3 RB exceeding the operability limit of SSO components and challenge the Vital Auxiliaries Cooling Nuclear Safety Performance Criterion.
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| Note: The additional risk added because of these VFDRs, as determined from the FRE for this fire area, is provided in SE Table 3.5.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 250 Recovery Actions (RAs)
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| The licensee did not identify any recovery actions required for this fire area.
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| Fire Detection & Suppression Systems Required to Meet the Nuclear Safety Performance Criteria The licensee performed an evaluation of the fire detection and suppression systems in this area.
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| The results were documented in LAR Table 4-4 in the LAR, and the applicable portions have been included below. Partial detection is installed, so the existing detection requires an engineering evaluation. The identified fire detection system modifications are required to improve plant fire detection and fire brigade response time.
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| Suppression and detection systems were identified in LAR Table B-3 as elements required for the fire resistance qualification of the three hour fire rated wall. Suppression is limited to fire brigade capability and provides adequate suppression capability given the hazards in the BH12 fire area.
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| Suppression Detection Required System?
| |
| Auto Required System?
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| Fire Fire Detection Zone Description Suppression I Area Zone Provided?
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| Provided?
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| E R D S E R D S I
| |
| Units 1 & 2 Block Yes Yes BH12 45 No No No No No Yes No No House (MR) (MR)
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| Legend:
| |
| E - EEEElLA: Systems required for acceptability of EEE Evaluations I NRC approved Licensing Action (Section 2.2.7)
| |
| R - Risk: Systems required to meet the Risk Criteria for the PB Approach (Section 4.2.4)
| |
| D - Defense-in-Depth: Systems required to maintain adequate balance of Defense-in-Depth for a PB Approach (Section 4.2.4.2)
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| S - Separation Criteria: Systems required for NFPA 805, Chapter 4, Separation Criteria in (Section 4.2.3)
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| MR-Modification Required Systems are committed to be modified as indicated in Table 4-4 and Attachment S Fire Area BH12 Conclusion The licensee has utilized the FRE PB approach to demonstrate the ability to meet the NFPA 805 nuclear safety performance criteria for this fire area. A FRE in accordance with NFPA 80S, Section 4.2.4.2, in conjunction with deterministic methods for simplifying assumptions, was used in applying this approach.
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| Based on the information provided in the LAR, as supplemented, the NRC staff finds Fire Area BH12 meets the nuclear safety goals, objectives, and performance criteria of NFPA 805. This conclusion is based on the following:
| |
| * Fire protection SSCs were evaluated in accordance with NFPA 805 Chapter 4 to determine which, if any, were required to meet the nuclear safety performance criteria.
| |
| This evaluation included:
| |
| The fire protection detection systems required to meet the nuclear safety performance criteria were documented.
| |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 251 Fire Area boundaries were defined using three hour rated walls, ceilings and floors, including fire barriers, fire barrier penetrations and through penetration fire stops and spatial separation.
| |
| * No exemptions or licensing actions from the pre-transition fire protection requirements were required for transition to the NFPA 805 RI/PB FPP.
| |
| * Fifteen VFDRs were identified, evaluated through the performance of a FRE, and either found to meet the risk acceptance criteria, as well as the requirements for DID and SMs, or modifications were planned to address the issue. The acceptability of the risk for this fire area is contingent on the risk reduction from the planned PSW modification (see SE Section 3.4 for a detailed discussion of the NRC staff's review of the adequacy of the FRE method used at ONS).
| |
| This fire area did not require the use of recovery actions to meet the nuclear safety performance criteria.
| |
| * The following modifications were identified to address VFDRs:
| |
| * Modification to improve general area and/or hazard detection for fire area BH12 were identified as required. These detection modifications are to improve plant fire detection and fire brigade response time.
| |
| * Modifications to install hinged steel covers/shields to the exterior side of the tornado vents that will qualify the wall as adequate for the hazard.
| |
| Fire Area BH3, Unit 3 Block House The licensee analyzed this fire area using the FRE approach in accordance with NFPA 805, Section 4.2.4.2, but also used deterministic simplifying assumptions in order to credit those portions of the facility design that met the deterministic requirements of NFPA 805, Section 4.2.3. The licensee identified the SSCs necessary to meet the nuclear safety performance criteria in this fire area.
| |
| Evaluation of Fire Suppression Effects on Nuclear Safety Performance Criteria The licensee stated in Attachment C, "NEI 04-02, Table B-3, Fire Area Transition," that safe and stable conditions can be achieved and maintained using equipment and cables outside of the area of fire suppression activity. Flooding of the suppression areas and discharge of suppression water to adjacent compartments is controlled and will not jeopardize achievement of safe and stable conditions.
| |
| Based on the information provided by the licensee in the NFPA 805 LAR, the NRC staff finds the licensee's evaluation of fire suppression effects on nuclear safety performance criteria acceptable because the results of the licensee's analysis indicate that fire suppression activities will not adversely affect achievement of the nuclear safety performance criteria.
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| OFFICI,I\L USE ONLY SECURITY RELATED INFORM,I\TION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 252 Exemptions and Other Licensing Actions The licensee did not credit any previously approved licensing actions or exemptions from the existing fire protection requirements.
| |
| Variation from Deterministic Requirements (VFDRs)
| |
| Fire Area BH3 has a total of 14 VFDRs, which are provided in the table below. All of these VFDRs are variances from NFPA 805, Section 4.2.3 (separation issues) that were dispositioned with a FRE (SE Section 3.4.3). The licensee's FRE determined that these variances are acceptable based on 1) the change in CDF and LERF for the 'fire area and the total CDF and LERF for each unit meet the acceptance criteria of RG 1.174 (SE Section 3.4.6) and 2) adequate DID and SMs are maintained for each fire area (SE Section 3.4.2). This determination relies on the following fire protection systems and features to meet the acceptance criteria:
| |
| * General area and/or fire hazard detection for the fire area BH3 is required to meet the DID criteria.
| |
| Based on the reliance on fire detectors in fire area BH3 to meet the DID criteria, the licensee has committed to make modifications to the fire detection system, which may include fire detector upgrades and/or new installation (SE Section 2.8.1).
| |
| Component VFDR# VFDR Description (Cables)
| |
| Normally open valve 1HP VA0023 is in the flow path from the LOST to the suction of the credited HPJ pump. Normally closed valve 1HP VA0939 isolates the flow path from the LOST to the containment sump. Recirculation flow to the LOST during 1HP VAOO23 prolonged operation of the HPI pump at low flow conditions may result in an HPI Normal increase in temperature of LOST contents to the operability limit of the HPI pump. Suction MaV, BH3-01 The contents of the LOST must be diverted to the containment sump by opening 1HP VA0939 1HP VA0939 and closing 1HP VA0023 prior to the operability limit of the HPJ pump LOST to being exceeded to prevent challenging the Reactivity, Inventory and Pressure Emergency Sump Control Nuclear Safety Performance Criteria. Fire damage to cables for electrical MaV equipment supplying power to these valves may prevent the valves from being operated.
| |
| 1MS VAOO17, 1MS VAOO24, 1MS VAOO26, 1MS VAOO33, These normally open, required closed valves isolate flow paths from the Main Steam 1MS VAOO35, Headers. Fire damage to cables for electrical equipment supplying power to these 1MS VAOO36, BH3-03 valves may prevent these valves from being closed and could result in overcooling 1MS VAOO76, and shrinkage of RC inventory. This could challenge the Decay Heat Removal 1MS VAOO79, Nuclear Safety Performance Criterion.
| |
| 1MS VAOO82, 1MS VAOO84 SG Isolation MaVs 1RC SXTRNOO1, Pressurizer heaters are required for RC pressure control. The heaters receive non-1RC SXTRN002 credited power from Unit 1 and credited power from a PSW system power supply.
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| Pressurizer BH3-04 The transfer of credited power to the pressurizer heaters requires a recovery action.
| |
| Heaters PSW Failure to transfer credited power to the heaters could challenge the Pressure Power Transfer Control Nuclear Safety Performance Criterion.
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| Switches Normally open valve 2HP VA0023 is in the flow path from the LOST to the suction of 2HP VAOO23 BH3-05 the credited HPI pump. Normally closed valve 2HP VA0939 isolates the flow path HPI Normal from the LOST to the containment sump. Recirculation flow to the LOST durinQ Suction MaV, OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 253 Component VFDR# VFDR Description (Cables) prolonged operation of the HPI pump at low flow conditions may result in an 2HP VA0939 increase in temperature of LDST contents to the operability limit of the HPI pump. LDST to The contents of the LDST must be diverted to the containment sump by opening Emergency Sump 2HP VA0939 and closing 2HP VA0023 prior to the operability limit of the HPI pump MaV being exceeded to prevent challenging the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria. Fire damage to cables for electrical equipment supplying power to these valves may prevent the valves from being operated.
| |
| 2MS VAOO17, 2MS VAOO24, 2MS VAOO26, 2MS VAOO33, These normally open, required closed valves isolate flow paths from the Main Steam 2MS VAOO35, Headers. Fire damage to cables for electrical equipment supplying power to these 2MS VAOO36, BH3-07 valves may prevent these valves from being closed and could result in overcooling 2MS VAOO76, and shrinkage of RC inventory. This could challenge the Decay Heat Removal 2MS VAOO79, Nuclear Safety Performance Criterion.
| |
| 2MS VAOO82, 2MS VAOO84 SG Isolation MaVs 2RC SXTRN001 ,
| |
| Pressurizer heaters are required for RC pressure control. The heaters receive non- 2RC SXTRNOO2, credited power from Unit 2 and credited power from the PSW system power 2RC SXTRN003 BH3-08 supply. The transfer of credited power to the pressurizer heaters requires a Pressurizer recovery action. Failure to transfer credited power to the heaters could challenge Heaters PSW the Pressure Control Nuclear Safety Performance Criterion. Power Transfer Switches Normally open valve 3HP VA0023 is in the flow path from the LDST to the suction of the credited HPI pump. Normally closed valve 3HP VA0939 isolates the flow path from the LDST to the containment sump. Recirculation flow to the LDST during 3HP VAOO23 prolonged operation of the HPI pump at low flow conditions may result in an HPI Normal increase in temperature of LDST contents to the operability limit of the HPI pump. Suction MaV, BH3-09 The contents of the LDST must be diverted to the containment sump by opening 3HP VA0939 3HP VA0939 and closing 3HP VA0023 prior to the operability limit of the HPI pump LDST to being exceeded to prevent challenging the Reactivity, Inventory and Pressure Emergency Sump Control Nuclear Safety Performance Criteria. Fire damage to cables for electrical MaV equipment supplying power to these valves may prevent the valves from being operated.
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| 3MS VAOO17, 3MS VAOO24, 3MS VAOO26, 3MS VAOO33, These normally open, required closed valves isolate flow paths from the MSHs .
| |
| 3MS VAOO35, Fire damage to cables for electrical equipment supplying power to these valves may 3MS VAOO36, BH3-11 prevent these valves from being closed and could result in overcooling and 3MS VAOO76, shrinkage of RC inventory. This could challenge the Decay Heat Removal Nuclear 3MS VAOO79, Safety Performance Criterion.
| |
| 3MS VAOO82, 3MS VAOO84 SG Isolation MaVs 3RC SXTRN001 ,
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| Pressurizer heaters are required for RC pressure control. The heaters receive non- 3RC SXTRNOO2, credited power from Unit 3 and credited power from the PSW system power supply. 3RC SXTRN003 BH3-12 The transfer of credited power to the pressurizer heaters requires a recovery action. Pressurizer Failure to transfer credited power to the heaters could challenge the Pressure Heaters PSW Control Nuclear Safety Performance Criterion. Power Transfer Switches Fire damage to cables for electrical equipment supplying power to the station HVAC Units 1 & 2 BH3-14 system may result in the temperature inside the Units 1 & 2 control complex Control Complex OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 254 Component VFDR# VFDR Description (Cables) exceeding the operability limit of SSD components and challenge the Vital Cooling Auxiliaries Nuclear Safety Performance Criterion.
| |
| Fire damage to cables for electrical equipment supplying power to the station HVAC system may result in the temperature inside the Unit 3 control complex exceeding Unit 3 Control BH3-15 the operability limit of SSD components and challenge the Vital Auxiliaries Nuclear Complex Cooling Safety Performance Criterion.
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| Fire damage to cables for electrical equipment supplying power to the containment Unit 1 cooling system may result in the temperature inside the Unit 1 RB exceeding the BH3-16 Containment operability limit of SSD components and challenge the Vital Auxiliaries Nuclear Cooling Safety Performance Criterion.
| |
| Fire damage to cables for electrical equipment supplying power to the containment Unit 2 cooling system may result in the temperature inside the Unit 2 RB exceeding the BH3-17 Containment operability limit of SSD components and challenge the Vital Auxiliaries Nuclear Cooling Safety Performance Criterion.
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| Fire damage to cables for electrical equipment supplying power to the containment Unit 3 cooling system may result in the temperature inside the Unit 3 RB exceeding the BH3-18 Containment operability limit of SSD components and challenge the Vital Auxiliaries Nuclear Cooling Safety Performance Criterion.
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| Note: The additional risk added because of these VFoRs, as determined from the FRE for this fire area, is provided in SE Table 3.5.
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| Recovery Actions (RAs)
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| The licensee did not identify any recovery actions required for this fire area.
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| Fire Detection & Suppression Systems Required to Meet the Nuclear Safety Performance Criteria The licensee performed an evaluation of the fire detection and suppression systems in this area.
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| The results were documented in LAR Table 4-4 and the applicable portions have been included below. Partial detection is installed in BH3 and modification is required to improve general area and/or hazard detection for DID. The identified fire detection systems modifications are required to improve plant fire detection and fire brigade response time.
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| Suppression Detection Required Auto Required System? System?
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| Fire Fire Detection Zone Description Suppression Area Zone Provided?
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| Provided?
| |
| E R D S E R D S I
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| Yes BH3 47 Unit 3 Block House No No No No No Yes No No No I (MR)
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| Legend:
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| E - EEEE/LA: Systems required for acceptability of EEE Evaluations I NRC approved Licensing Action (Section 2.2.7)
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| R - Risk: Systems required to meet the Risk Criteria for the PB Approach (Section 4.2.4)
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| D - Defense-in-Depth: Systems required to maintain adequate balance of Defense-in-Depth for a PB Approach (Section 4.2.4)
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| S - Separation Criteria: Systems required for NFPA 805, Chapter 4, Separation Criteria in (Section 4.2.3)
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| MR - Modification Required Systems are committed to be modified as indicated in LAR Table 4-4 and Attachment S Conclusion The licensee has utilized the FRE PB approach to demonstrate the ability to meet the NFPA 805 nuclear safety performance criteria for this fire area. A FRE in accordance with NFPA 805, OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY REL,A,TED INFORMATION 255 Section 4.2.4.2, in conjunction with deterministic methods for simplifying assumptions, was used in applying this approach.
| |
| Based on the information provided in the LAR, as supplemented, the NRC staff finds Fire Area BH3 meets the nuclear safety goals, objectives, and performance criteria of NFPA 805. This conclusion is based on the following:
| |
| * Fire protection SSCs were evaluated in accordance with NFPA 805 Chapter 4 to determine which, if any, were required to meet the nuclear safety performance criteria. This evaluation included:
| |
| : a. The fire protection detection systems required to meet the nuclear safety performance criteria were documented.
| |
| : b. Fire Area boundaries were defined using three hour rated walls, ceilings and floors, including fire barriers, fire barrier penetrations and through penetration fire stops and spatial separation.
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| * No exemptions or licensing actions from the pre-transition fire protection requirements were required for transition to the NFPA 805 RI/PB FPP.
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| * Fourteen VFDRs were identified, evaluated through the performance of a FRE, and either found to meet the risk acceptance criteria, as well as the requirements for DID and SMs, or modifications were planned to address the issues. The acceptability of the risk for this fire area is contingent on the risk reduction from the planned PSW modification (see SE Section 3.4 for a detailed discussion of the NRC staff's review of the adequacy of the FRE method used at ONS).
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| * This fire area did not require the use of recovery actions to meet the nuclear safety performance criteria.
| |
| * The following modifications were identified to address VFDRs:
| |
| a Modification to improve general area and/or hazard detection for fire area BH3 were identified as required. These detection modifications are to improve plant fire detection and fire brigade response time.
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| Fire Area RB1. Unit 1 Reactor Building The licensee analyzed this fire area using the FRE approach in accordance with NFPA 805, Section 4.2.4.2, but also used deterministic simplifying assumptions in order to credit those portions of the facility design that met the deterministic requirements of NFPA 805, Section 4.2.3. The licensee identified the SSCs necessary to meet the nuclear safety performance criteria in this fire area.
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| Evaluation of Fire Suppression Effects on Nuclear Safety Performance Criteria The licensee stated in Attachment C, "NEI 04-02, Table B-3, Fire Area Transition," that safe and stable conditions can be achieved and maintained using equipment and cables outside of the area of fire suppression activity. Flooding of the suppression areas and discharge of OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 256 suppression water to adjacent compartments is controlled and will not jeopardize achievement of safe and stable conditions. Based on the information provided by the licensee in the NFPA 805 LAR, the NRC staff finds the licensee's evaluation of fire suppression effects on nuclear safety performance criteria acceptable because the results of the licensee's analysis indicate that fire suppression activities will not adversely affect achievement of the nuclear safety performance criteria.
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| Exemptions and Other Licensing Actions Based on the information provided in the LAR, the licensee credited two previously approved exemptions from the existing fire protection requirements. The licensee used the process described in LAR Section 4.2.3, "Licensing Action Transition," and Attachment K, "Licensing Action Review," to carry forward these exemptions, which requires a determination of the basis of acceptability and a determination that the basis of the acceptability is still valid. The NRC staff's evaluation of each exemption is provided in the table below.
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| Exemption I Licensing Licensee's Statement on Basis and Continuing Validity NRC Staff Evaluation Action Appendix R Presented justification for the lack of 20 feet horizontal Based on the previous NRC staff Exemption, distance separation between SSD circuits with no approval of the portion of this RB 20 feet intervening combustibles. For 20 feet separation with exemption having to do with lack of 20 separation intervening combustibles: feet horizontal distance separation w/o
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| * More than 20 feet separation. between SSD circuits and the intervening
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| * Low concentration of cables in cable trays. statement by the licensee that the combustibles
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| * Cable insulation is comparable to IEEE-383 qualified basis remains valid, the NRC staff finds cables which burn slowly with an initial low rate of heat this portion of this acceptable.
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| release.
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| * Fire brigade response would be adequate. However, for the pressurizer level instrumentation 15 feet separation For pressurizer level instrumentation 15 feet separation exemption, the NRC staff disagrees (RB1 Only): that the bases for previous acceptance remains valid because the separation distance was reported by the licensee
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| * No intervening combustibles.
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| to be less than 15 feet described in the
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| * Low combustible loading in general area.
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| exemption documentation. The NRC
| |
| * Administrative controls to limit transient combustibles staff does not therefore find this portion in area.
| |
| of this acceptable. The licensee,
| |
| * Inspections prior to starting the unit after an outage.
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| however, further evaluated this as a
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| * RB is a huge structure to dissipate heat from a fire.
| |
| VFDR (VFDR RB1-11) and determined
| |
| * Fire brigade response would be adequate. it to have negligible risk with a recovery action (1 RC P 0233) to provide DID.
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| The bases for previous acceptance remain valid. Incorporating this recovery action in the SSD procedures is an implementation item (SE Section 2.8, Table 2.8.1-1, Item 14).
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| Appendix R Provides the following justification for the lack of three hour Based on the previous NRC staff Exemption, fire rated pipe penetrations: approval of this exemption and the RB Unrated
| |
| * RB walls serve as a substantial heat sink. statement by the licensee that the Containment
| |
| * Combustible loading near penetrations is low. basis remains valid, the NRC staff finds Mechanical
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| * Mechanical pipe penetrations are designed to meet this acceptable.
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| Penetrations multiple containment integrity criteria and are substantial.
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| * Large room volumes on both sides dissipate heat from a fire away from penetration area.
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| The bases for previous acceptance remain valid.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 257 Variation from Deterministic Requirements (VFDRs)
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| Fire Area RB'I has a total of 17 VFDRs, which are provided in the table below. All of these VFDRs are variances from NFPA 805 Section 4.2.3 (separation issues) that were dispositioned with a FRE (SE Section 3.4.3). The licensee's FRE determined that these variances are acceptable based on 1) the change in CDF and LERF for the fire area and the total CDF and LERF for each unit meet the acceptance criteria of RG 1.174 (SE Section 3.4.6) and 2) adequate DID and SMs are maintained for each fire area (SE Section 3.4.2). This determination relies on the following fire protection systems and features to meet the acceptance criteria:
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| * All VFDRs require reliance on the general area and/or hazard detection associated with existing fire detection in the RB1 to meet the DID criteria.
| |
| * For VFDRs RB1-02, RB1-09, RB1-11, and RB1-12, in addition to the reliance on existing fire detection, recovery actions are identified to monitor alternative instrumentation as stated in LAR Attachment G Table G-2 and are identified as relied upon to meet DID criteria (SE Section 2.9, Table 2.9-1, Item 14).
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| * ForVFDRs RB1-10 and RB1-16, in addition to reliance on existing fire detection, operator guidance will be inserted into shutdown procedures for operation of RC high point vent valves for RC letdown in the event that head vent valve flow path becomes inoperable (SE Section 2.9; Table 2.9-1 ,Item 30).
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| VFDR# VFDR Description Component (Cables)
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| This normally closed, required closed valve provides train separation by isolating the cross connect header between the 1A and 1B trains of Emergency Feedwater. Fire induced cable damage may result in 1CCWVA0269 - SG A FOW RB1-01 spurious opening of this valve, a diversion of flow to either the 1A or Control MOV 1B SGs, and a challenge to the Oecay Heat Removal Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 damaqe.
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| SG level indication is required for process monitoring and diagnosis of plant transients. Fire induced cable damage may result in loss of SG level indication resulting in the inability of the operator to monitor 1FOWP 0270, 1FOWP 0271 - SG RB1-02 and control level in either the 1A or 1B SGs from the MCR and Level Indications challenge the Process Monitoring Nuclear Safety Performance Criterion.
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| This normally open, required open valve is located in the EFW flow path to the 1B SG. Fire induced cable damage may result in spurious closing of this valve, isolating Protected Service Water flow to the 1B SG. The subsequent decrease in SG shell temperature RB1-03 may result in 1B SG exceeding its tube to shell differential 1FOWVA0347 - SG B Inlet MOV temperature limit. This could challenge the Inventory Control and Oecay Heat Removal Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 damage.
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| Normally open valve 1HP VA0023 is in the flow path from the LOST to the suction of the credited HPI pump. Normally closed valve 1HP VA0939 isolates the flow path from the LOST to the containment sump. RecirCUlation flow to the LOST during prolonged operation of 1HP VA0023 - HPJ Normal Suction the HPI pump at low flow conditions may result in an increase in RB1-04 MOV, 1HP VA0939 - LOST to temperature of LOST contents to the operability limit of the HPI Emergency Sump MOV pump. The contents of the LOST must be diverted to the U n t a l n m e n t ,"mp by opening 1HP VA0939 and Closing 1HP VA0023 prior to the operability limit of the HPI pump being exceeded to prevent challenging the Reactivity, Inventory and Pressure Control OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 258
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| -VFDR# VFDR Description Component (Cables)
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| Nuclear Safety Performance Criteria. Although unaffected by fire, the power supplies for these valves are not credited following a fire in this fire area and a loss of power may prevent these valves from being repositioned.
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| These normally open, required closed valves isolate the flow path from the BWST to the LPI Pumps, RBS Pumps, and containment sump. Although unaffected by fire, the power supplies to these valves are not credited following a fire in this fire area and a loss of power may prevent these valves from being closed and result in a 1LP VA0021, 1LP VA0022 - BWST RB1-06 diversion of BWST inventory to the containment sump via the LPI Suction MOVs system. In addition, an inadvertent ES actuation could result in a diversion of BWST inventory to the containment sump via the RBS system. A loss of BWST inventory could challenge the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria.
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| Source range flux indication is required for process monitoring and control. Fire induced cable damage may result in loss of source 1RPSP 1007, 1RPSP 1008 RB1-07 range flux indications resulting in the inability of the operator to Source Range Flux monitor this parameter from the MCR and challenge the Process MonitorinQ Nuclear Safety Performance Criterion.
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| These normally open, required closed valves isolate flow paths from 1MS VA0017, 1MS VA0024, 1MS the MSHs. Although unaffected by fire, the power supplies to these VA0026, 1MS VA0033, 1MS valves are not credited following a fire in this fire area and a loss of VA0035, 1MS VA0036, 1MS RB1-08 power may prevent these valves from being closed and could result VA0076, 1MS VA0079, 1MS in overcooling and shrinkage of RC inventory. This could challenge VA0082, 1MS VA0084 - SG the Decav Heat Removal Nuclear Safety Performance Criterion. Isolation MOVs RC pressure indication is required for process monitoring and diagnosis of plant transients. Fire induced cable damage may result in loss of RC pressure indications resulting in the inability of the 1RC CR0045 - RC Pressure RB1-09 operator to monitor and control this parameter from the MCR and Recorder challenge the Process Monitoring Nuclear Safety Performance Criterion.
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| Operation of the pressurizer heaters is required to maintain control of 1RC HE0001, 1RC HE0002, 1RC RC pressure. Fire damage to cables may result in a loss of the RB1-10 HE0003, 1RC HE0004 pressurizer heaters and could challenge the Pressure Control Pressurizer Heaters Nuclear Safety Performance Criterion.
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| Pressurizer level indication is required for process monitoring and diagnosis of plant transients. Fire impingement on instrument sensing lines or fire induced cable damage may result in loss of 1RC P 0365 - Pressurizer Level RB1-11 pressurizer level indication resulting in the inability of the operator to Indication monitor and control this parameter from the MCR and challenge the Process MonitorinQ Nuclear Safety Performance Criterion.
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| RC temperature indication is required for process monitoring and diagnosis of plant transients. Fire induced cable damage may result in loss of RC temperature indication resulting in the inability of the 1RC P 0376 - RC Temperature RB1-12 operator to monitor and control this parameter from the MCR and Indication challenge the Process Monitoring Nuclear Safety Performance Criterion.
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| Pressurizer heaters are required for RC pressure control. The heaters receive non-credited power from Unit 1 and credited power 1RC SXTRN001, 1RC SXTRN002 from the PSW system power supply. The transfer of credited power RB1-13 - Pressurizer Heaters PSW Power to the pressurizer heaters requires a recovery action. Failure to Transfer Switches transfer credited power to the heaters could challenge the Pressure Control Nuclear Safety Performance Criterion.
| |
| This normally closed, required closed valve isolates the flow path from the RCS to the Quench Tank. Fire induced cable damage may 1RC VA0066 - Pressurizer Power RB1-14 result in spurious opening of the PORV causing a loss of inventory Operated Relief Valve and RC subcooling. This could challenge the Inventory and Pressure Control Nuclear Safety Performance Criteria.
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| These normally closed, required closed valves isolate flow paths from 1RC VA0155, 1RC VA0157, 1RC RB1-15 the RCS to containment. Potential hot shorts within the electrical VA0159 - RC Hot Leg and Head OFFICIAL USE ONLY SECURITY RELATED INFORMl\TION
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| | |
| OFFICIAL USE ONLY SECURITY RELJ\TED INFORMATION 259 VFDR# VFDR Description Component (Cables) penetration box may spuriously open the reactor head vent and hot Vent Valves leg vent valves. The spurious opening of these valves may result in a loss of RC inventory and challenge the Inventory and Pressure Control Nuclear Safety Performance Criteria.
| |
| These normally closed valves isolate the flow path from the RCS to containment. These valves are required opened to provide an RC letdown flow path. Fire induced cable damage may prevent these 1RC VA0159, 1RC VA0160 - RC RB1-16 valves from being opened resulting in the lifting of the pressurizer Head Vent Valves safety relief valves and a challenge to the Inventory Control Nuclear Safety Performance Criterion.
| |
| Although unaffected by fire, the power supplies for the station HVAC system are not credited following a fire in this fire area and a loss of Units 1 & 2 Control Complex RB1-18 power may result in the temperature inside the Units 1 & 2 control Cooling complex exceeding the operability limit of SSD components and challenQe the Vital Auxiliaries Nuclear Safety Performance Criterion.
| |
| Fire damage to cables may result in a loss of power to the containment cooling system and may result in the temperature inside RB1-19 the Unit 1 RB exceeding the operability limit of SSD components. Unit 1 Containment Cooling This could challenge the Vital Auxiliaries Nuclear Safety Performance Criterion.
| |
| Note: The additional risk added because of these VFDRs, as determined from the FRE for this fire area, is provided in SE Table 3.5.
| |
| Recovery Actions (RAs)
| |
| Recovery actions credited in this fire area to satisfy the DID requirements of NFPA 805, Section 4.2.4.2, are provided in the following table:
| |
| Component Component Name Description of Action ID U1 SSF SG 1B LEVEL For a fire in the east side of containment, dispatch an operator 1FDWP 0232 INDICATION to the SSF to monitor instrument 1FDWP 0232.
| |
| For a fire in the east side of containment, monitor 1RC P 0365 U1 SSF PRESSURIZER 1RCP0233 from the control room if available; if not dispatch an operator to LEVEL INDICATION the SSF to monitor instrument 1RC P 0233.
| |
| U1 SSF RC LOOP B For a fire in the east side of containment, dispatch an operator 1RC P 0238 PRESSURE INDICATION to the SSF to monitor instrument 1RC P 0238.
| |
| For a fire in the east side of containment, dispatch an operator 1RC P 0315 REACTOR OUTLET LOOP B to the SSF to monitor instrument 1RC P 0315.
| |
| Note: The FRE for this fire area determined that the additional risk being added because of these RAs was negligible for both change in CDF and change in LERF. See SE Section 3.4.2 for a detailed discussion of the NRC staff's review of the FREs.
| |
| Fire Detection & Suppression Systems Required to Meet the Nuclear Safety Performance Criteria The licensee performed an evaluation of the fire detection and suppression systems in this area.
| |
| The results were documented in LAR Table 4-4 and the applicable portion has been included below.
| |
| OFFICIAL USE O~JLY SECURITY RELJ\TED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELJ\TED INFORMATION 260 Suppression Detection Required Auto Required System? System?
| |
| Fire Fire Zone Detection Suppression Area Zone Description Provided?
| |
| Provided?
| |
| E R D S E R D S Un it 1 Reactor RB1 122 No No No No No Yes No No Yes No Buildinq Legend:
| |
| E - EEEE/lA Systems required for acceptability of EEE Evaluations I NRC approved Licensing Action (Section 2.2.7)
| |
| R - Risk: Systems required to meet the Risk Criteria for the PB Approach (Section 4.2.4) o - Defense-in-Depth: Systems required to maintain adequate balance of Defense-in-Depth for a PB Approach (Section 4.2.4.2)
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| S - Separation Criteria: Systems required for Chapter 4 Separation Criteria in (Section 4.2.3)
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| MR - Modification Required Systems are committed to be modified as indicated in LAR Table 4-4 and Attachment S Fire Area RB 1 Conclusion The licensee has utilized the FRE PB approach to demonstrate the ability to meet the NFPA 805 nuclear safety performance criteria for this fire area. A FRE in accordance with NFPA 805, Section 4.2.4.2, in conjunction with deterministic methods for simplifying assumptions, was used in applying this approach.
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| Based on the information provided in the LAR, as supplemented, the NRC staff finds Fire Area RB1 meets the nuclear safety goals, objectives, and performance criteria of NFPA 805. This conclusion is based on the following:
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| * Fire protection SSCs were evaluated in accordance with NFPA 805 Chapter 4 to determine which, if any, were required to meet the nuclear safety performance criteria. This evaluation included:
| |
| : a. The fire protection detection systems required to meet the nuclear safety performance criteria were documented.
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| : b. Fire Area boundaries were defined using three hour rated walls, ceilings and floors, including fire barriers, fire barrier penetrations and through penetration fire stops and spatial separation.
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| * Two exemptions from the pre-transition fire protection requirements were evaluated and found to be valid and applicable under the NFPA 805 RI/PB FPP.
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| * Seventeen VFDRs were identified, evaluated through the performance of a FRE, and found to meet the risk acceptance criteria, as well as the requirements for DID and SMs. The acceptability of the risk for this fire area is contingent on the risk reduction from the planned PSW modification (see SE Section 3.4 for a detailed discussion of the NRC staff's review of the adequacy of the FRE method used at ONS).
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| * Four recovery actions were identified and evaluated for the additional risk (change in CDF and change in LERF) each poses. The additional risk of each recovery action was conservatively estimated to be taken as the change in CDF and change in LERF associated with the VFDR that resulted in the need for the recovery action. The change in CDF and change in LERF for each recovery action was determined to be negligible.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 261 Fire Area RB2, Unit 2 Reactor BUilding The licensee analyzed this fire area using the FRE approach in accordance with NFPA 805 Section 4.2.4.2, but also used deterministic simplifying assumptions in order to credit those portions of the facility design that met the deterministic requirements of NFPA 805 Section 4.2.3. The licensee identified the SSCs necessary to meet the nuclear safety performance criteria in this fire area.
| |
| Evaluation of Fire Suppression Effects on Nuclear Safety Performance Criteria The licensee stated in Attachment C, "NEI 04-02, Table B-3, Fire Area Transition," that safe and stable conditions can be achieved and maintained using equipment and cables outside of the area of fire suppression activity. Flooding of the suppression areas and discharge of suppression water to adjacent compartments is controlled and will not jeopardize achievement of safe and stable conditions. Based on the information provided by the licensee in the NFPA 805 LAR, the NRC staff finds the licensee's evaluation of fire suppression effects on nuclear safety performance criteria acceptable because the results of the licensee's analysis indicate that fire suppression activities will not adversely affect achievement of the nuclear safety performance criteria.
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| Exemptions and Other Licensing Actions Based on the information provided in the LAR, the licensee credited one previously approved exemption from the existing fire protection requirements. The licensee used the process described in LAR Section 4.2.3, "Licensing Action Transition," and Attachment K, "Licensing Action Review," to carry forward this exemption, which requires a determination of the basis of acceptability and a determination that the basis of the acceptability is still valid. The NRC staff's evaluation of the exemption is provided in the table below.
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| Exemption I Licensing Licensee's Statement on Basis and Continuing
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| | |
| ===NRC Staff Evaluation===
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| Action Validity Appendix R Exemption, RB Provides the following justification for the lack of three Based on the previous Unrated Containment hour fire rated pipe penetrations: NRC staff approval of this Mechanical Penetrations
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| * RB walls serve as a substantial heat sink. exemption and the
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| * Combustible loading near penetrations is low. statement by the licensee
| |
| * Mechanical pipe penetrations are designed to meet that the basis remains mUltiple containment integrity criteria and are valid, the NRC staff finds substantial. this acceptable.
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| * Large room volumes on both sides dissipate heat from a fire away from penetration area.
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| The bases for previous acceptance remain valid.
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| Variation from Deterministic Requirements (VFDRs)
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| Fire Area RB2 has a total of 17 VFDRs, which are provided in the table below. All of these VFDRs are variances from NFPA 805 Section 4.2.3 (separation issues) that were dispositioned with a FRE (SE Section 3.4.3). The licensee's FRE determined that these variances are acceptable based on 1) the change in CDF and LERF for the fire area and the total CDF and LERF for each unit meet the acceptance criteria of RG 1.174 (SE Section 3.4.6) and 2) adequate DID and SMs are maintained for each fire area (SE Section 3.4.2). This OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 262 determination relies on the following fire protection systems and features to meet the acceptance criteria:
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| * All VFDRs require reliance on the general area and/or hazard detection associated with existing fire detection in the RB2 to meet the DID criteria.
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| * For VFDRs RB2-02, RB2-09, RB2-11, and RB2-12, in addition to the reliance on existing fire detection, recovery actions are identified to monitor alternative instrumentation as stated in Attachment G Table G-2 of the LAR and are identified as relied upon to meet DID criteria (SE Section 2.9, Table 2.9-1, Item 14).
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| * For VFDRs RB2-10 and RB2-16, in addition to reliance on existing fire detection, operator guidance will be inserted into shutdown procedures for operation of RC high point vent valves for RC letdown in the event that head vent valve flow path becomes inoperable (SE Section 2.9; Table 2.9-1, Item 30).
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| Component VFDR# VFDR Description (Cables)
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| This normally closed, required closed valve provides train separation by isolating the cross connect header between the 2A and 2B trains of Emergency Feedwater.
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| 2CC'NVA0269 Fire induced cable damage may result in spurious opening of this valve, a RB2-01 SGAFDW diversion of flow to either the 2A or 2B SGs, and a challenge to the Decay Heat Control MaV Removal Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 damaqe.
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| SG level indication is required for process monitoring and diagnosis of plant transients. Fire induced cable damage may result in loss of SG level indication 2FDWP 0270, RB2-02 resulting in the inability of the operator to monitor and control level in either the 2A 2FDWP 0271 - SG or 2B SGs from the MCR and challenge the Process Monitoring Nuclear Safety Level Indications Performance Criterion.
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| This normally open, required open valve is located in the EFW flow path to the 2B SG. Fire induced cable damage may result in spurious closing of this valve, isolating Protected Service Water flow to the 2B SG. The subsequent decrease in 2FD'NVA0347 RB2-03 SG shell temperature may result in 2B SG exceeding its tube to shell differential SG B Inlet MaV temperature limit. This could challenge the Inventory Control and Decay Heat I Removal Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 damaae.
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| Normally open valve 2HP VA0023 is in the flow path from the LOST to the suction of the credited HPI pump. Normally closed valve 2HP VA0939 isolates the flow path from the LOST to the containment sump. Recirculation flow to the LOST 2HP VA0023 - HPI during prolonged operation of the HPI pump at low flow conditions may result in an Normal Suction increase in temperature of LOST contents to the operability limit of the HPI pump.
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| MaV,2HP RB2-04 The contents of the LOST must be diverted to the containment sump by opening VA0939 - LOST to 2HP VA0939 and closing 2HP VA0023 prior to the operability limit of the HPI pump Emergency Sump being exceeded to prevent challenging the ReactiVity, Inventory and Pressure MaV I Control Nuclear Safety Performance Criteria. Although unaffected by fire, the power supplies for these valves are not credited follOWing a fire in this fire area and a loss of power may prevent these valves from beinq repositioned.
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| These normally open, required closed valves isolate the flow path from the BWST to the LPI Pumps, RBS Pumps, and containment sump. Although unaffected by fire, the power supplies to these valves are not credited following a fire in this fire area and a loss of power may prevent these valves from being closed and result in 2LP VA0021, 2LP RB2-06 a diversion of BWST inventory to the containment sump via the LPI system. In VA0022 - BWST addition, an inadvertent ES actuation could result in a diversion of BWST inventory Suction MaVs to the containment sump via the RBS system. A loss of BWST inventory could challenge the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria.
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| Source range flux indication is required for process monitoring and control. Fire 2RPSP 1007, RB2-07 induced cable damaae may result in loss of source ranqe flux indications resultinq 2RPSP 1008 OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 263 Component VFDR# VFDR Description (Cables) in the inability of the operator to monitor this parameter from the MCR and Source Range challenqe the Process Monitoring Nuclear Safety Performance Criterion. Flux 2MS VA0017, 2MS VA0024, 2MS VA0026, These normally open, required closed valves isolate flow paths from the MSHs . 2MS VA0033, Although unaffected by fire, the power supplies to these valves are not credited 2MS VA0035, RB2-08 following a fire in this fire area and a loss of power may prevent these valves from 2MS VA0036, being closed and could result in overcooling and shrinkage of RC inventory. This 2MS VA0076, could challenge the Decay Heat Removal Nuclear Safety Performance Criterion. 2MS VA0079, 2MS VA0082, 2MS VA0084 - SG Isolation MOVs RC pressure indication is required for process monitoring and diagnosis of plant transients. Fire induced cable damage may result in loss of RC pressure 2RC CR0046 - RC RB2-09 indications resulting in the inability of the operator to monitor and control this Pressure parameter from the MCR and challenge the Process Monitoring Nuclear Safety Recorder Performance Criterion.
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| 2RC HEOO01, 2RC HEOO02, Operation of the pressurizer heaters is required to maintain control of RC 2RC HEOO03, RB2-10 pressure. Fire damage to cables may result in a loss of the pressurizer heaters 2RC HEOO04 and could challenge the Pressure Control Nuclear Safety Performance Criterion.
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| Pressurizer I
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| Heaters Pressurizer level indication is required for process monitoring and diagnosis of plant transients. Fire impingement on instrument sensing lines or fire induced 2RC P 0365 RB2-11 cable damage may result in loss of pressurizer level indication resulting in the Pressurizer Level inability of the operator to monitor and control this parameter from the MCR and Indication challenQe the Process MonitorinQ Nuclear Safety Performance Criterion.
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| RC temperature indication is required for process monitoring and diagnosis of plant transients. Fire induced cable damage may result in loss of RC temperature 2RC P 0376 - RC RB2-12 indication resulting in the inability of the operator to monitor and control this Temperature parameter from the MCR and challenge the Process Monitoring Nuclear Safety Indication Performance Criterion.
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| 2RC SXTRN001, Pressurizer heaters are required for RC pressure control. The heaters receive 2RC SXTRN002, non-credited power from Unit 2 and credited power from a PSW system power 2RC SXTRN003 RB2-13 supply. The transfer of credited power to the pressurizer heaters requires a Pressurizer recovery action. Failure to transfer credited power to the heaters could challenge Heaters PSW the Pressure Control Nuclear Safety Performance Criterion. Power Transfer Switches This normally closed, required closed valve isolates the flow path from the RCS to 2RC VA0066 the Quench Tank. Fire induced cable damage may result in spurious opening of Pressurizer Power RB2-14 the PORV causing a loss of inventory and RC subcooling. This could challenge Operated Relief the Inventorv and Pressure Control Nuclear Safety Performance Criteria. Valve These normally closed, required closed valves isolate flow paths from the RCS to 2RC VA0155, containment. Potential hot shorts within the electrical penetration box may 2RC VA0157, RB2-15 spuriously open the reactor head vent and hot leg vent valves. The spurious 2RC VA0159 - RC opening of these valves may result in a loss of RC inventory and challenge the Hot Leg and Head Inventorv and Pressure Control Nuclear Safety Performance Criteria. Vent Valves These normally closed valves isolate the flow path from the RCS to containment.
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| These valves are required opened to provide an RC letdown flow path. Fire 2RC VA0159, RB2-16 induced cable damage may prevent these valves from being opened resulting in 2RC VA0160 - RC the lifting of the pressurizer safety relief valves and a challenge to the Inventory Head Vent Valves Control Nuclear Safety Performance Criterion.
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| Although unaffected by fire, the power supplies for the station HVAC system are Units 1 & 2 not credited following a fire in this fire area and a loss of power may result in the RB2-18 Control Complex temperature inside the Units 1 & 2 control complex exceeding the operability limit Cooling of SSD components and challenge the Vital Auxiliaries Nuclear Safety I OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 264 Component VFDR# VFDR Description (Cables)
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| Performance Criterion.
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| Fire damage to cables may result in a loss of power to the containment cooling Unit 2 system and may result ;n the temperature inside the Unit 2 RB exceeding the RB2-19 Containment operability limit of SSD components. This could challenge the Vital Auxiliaries Cooling Nuclear Safety Performance Criterion.
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| Note: The additional risk added because of these VFDRs, as determined from the FRE for this fire area, is provided in SE Table 3.
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| Recovery Actions (RAs)
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| Recovery actions credited in this fire area to satisfy the DID requirements of NFPA 805, Section 4.2.4.2, are provided in the following table:
| |
| Component Component Name Description of Action 10 U2 SSF SG 1B LEVEL For a fire in the east side of containment, dispatch an operator to the 2FDWP0232 INDICATION SSF to monitor instrument 2FDWP 0232.
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| For a fire in the east side of containment, monitor 2RC P 0365 from U2 SSF PRESSURIZER 2RC P 0233 the control room if available; if not dispatch an operator to the SSF to LEVEL INDICATION monitor instrument 2RC P 0233.
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| U2 SSF RC LOOP B For a fire in the east side of containment, dispatch an operator to the 2RC P 0238 PRESSURE INDICATION SSF to monitor instrument 2RC P 0238.
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| REACTOR OUTLET LOOP For a fire in the east side of containment, dispatch an operator to the 2RC P 0315 B SSF to monitor instrument 2RC P 0315.
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| Note: The FRE for this fire area determined that the additional risk being added because of these RAs was negligible for both change in CDF and change in LERF. See SE Section 3.4.2 for a detailed discussion of the NRC staff's review of the FREs.
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| Fire Detection & Suppression Systems Required to Meet the Nuclear Safety Performance Criteria The licensee performed an evaluation of the fire detection and suppression systems in this area.
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| The results were documented in LAR Table 4-4 and the applicable portion has been included below.
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| Suppression Detection Detection Required Auto Required System? Provided? System?
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| Fire Fire Zone Description Suppression Area Zone Provided?
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| E R D S E R D S Unit 2 Reactor RB2 123 No No No No No Yes No No Yes No Buildinq Legend:
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| E - EEEElLA: Systems required for acceptability of EEE Evaluations I NRC-approved Licensing Action (Section 2.2.7)
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| R - Risk: Systems required to meet the Risk Criteria for the PB Approach (Section 4.2.4)
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| D - Defense-in-Depth: Systems required to maintain adequate balance of Defense-in-Depth for a PB Approach (Section 4.2.4.2)
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| S - Separation Criteria: Systems required for Chapter 4 Separation Criteria in (Section 4.2.3)
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| MR - Modification Required Systems are committed to be modified as indicated in LAR Table 4-4 and Attachment S OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 265 Fire Area RB2 Conclusion The licensee has utilized the FRE PB approach to demonstrate the ability to meet the NFPA 805 nuclear safety performance criteria for this fire area. A FRE in accordance with NFPA 805, Section 4.2.4.2, in conjunction with deterministic methods for simplifying assumptions, was used in applying this approach.
| |
| * Based on the information provided in the LAR, as supplemented, the NRC staff finds Fire Area RB2 meets the nuclear safety goals, objectives, and performance criteria of NFPA 805.
| |
| This conclusion is based on the following: fire protection SSCs were evaluated in accordance with NFPA 805, Chapter 4, to determine which, if any, were required to meet the nuclear safety performance criteria. This evaluation included:
| |
| The fire protection detection systems required to meet the nuclear safety performance criteria were documented.
| |
| : a. Fire Area boundaries were defined using three hour rated walls, ceilings and floors, including fire barriers, fire barrier penetrations, and through penetration fire stops.
| |
| * One exemption from the pre-transition fire protection requirements was evaluated and found to be valid and applicable under the NFPA 805 RI/PB FPP.
| |
| * Seventeen VFDRs were identified, evaluated through the performance of a FRE, and found to meet the risk acceptance criteria, as well as the requirements for DID and SMs. The acceptability of the risk for this fire area is contingent on the risk reduction from the planned PSW modification (see SE Section 3.4 for a detailed discussion of the NRC staff's review of the adequacy of the FRE method used at ONS).
| |
| * Four recovery actions were identified and evaluated for the additional risk (change in CDF and change in LERF) each poses. The additional risk of each was conservatively estimated to be taken as the change in CDF and change in LERF associated with the VFDR that resulted in the need for the recovery action. The change in CDF and change in LERF for each recovery action was determined to be negligible.
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| Fire Area RB3. Unit 3 Reactor Building The licensee analyzed this fire area using the FRE approach in accordance with NFPA 805 Section 4.2.4.2, but also used deterministic simplifying assumptions in order to credit those portions of the facility design that met the deterministic requirements of NFPA 805 Section 4.2.3. The licensee identified the SSCs necessary to meet the nuclear safety performance criteria in this fire area.
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| Evaluation of Fire Suppression Effects on Nuclear Safety Performance Criteria The licensee stated in Attachment C, "NEI 04-02, Table B-3, Fire Area Transition," that safe and stable conditions can be achieved and maintained using equipment and cables outside of the area of fire suppression activity. Flooding of the suppression areas and discharge of suppression water to adjacent compartments is controlled and will not jeopardize achievement of safe and stable conditions. Based on the information provided by the licensee in the NFPA 805 LAR, the NRC staff finds the licensee's evaluation of fire suppression effects on nuclear safety performance criteria acceptable because the results of the licensee's analysis indicate OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 266 that fire suppression activities will not adversely affect achievement of the nuclear safety performance criteria.
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| Exemptions and Other Licensing Actions Based on the information provided in the LAR, the licensee credited one previously approved exemption from the existing fire protection requirements. The licensee utilized the process described in LAR Section 4.2.3, "Licensing Action Transition," and Attachment K, "Licensing Action Review," to carry forward this exemption, which requires a determination of the basis of acceptability and a determination that the basis of the acceptability is still valid. The NRC staff's evaluation of the exemption is provided in the table below.
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| Exemption I Licensing Licensee's Statement on Basis and Continuing
| |
| | |
| ===NRC Staff Evaluation===
| |
| Action Validity Appendix R Exemption, RB Provides the following justification for the lack of three Based on the previous NRC staff Unrated Containment hour fire rated pipe penetrations: approval of this exemption and Mechanical Penetrations the statement by the licensee that the basis remains valid, the NRC
| |
| * RB walls serve as a substantial heat sink.
| |
| staff finds this acceptable.
| |
| * Combustible loading near penetrations is low.
| |
| * Mechanical pipe penetrations are designed to meet multiple containment integrity criteria and are substantial.
| |
| * Large room volumes on both sides dissipate heat from a fire away from penetration area.
| |
| The bases for previous acceptance remain valid.
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| Variation from Deterministic Requirements (VFDRs)
| |
| Fire Area RB3 has a total of 17 VFDRs, which are provided in the table below. All of these VFDRs are variances from NFPA 805 Section 4.2.3 (separation issues) that were dispositioned with a FRE (SE Section 3.4.3). The licensee's FRE determined that these variances are acceptable based on 1) the change in CDF and LERF for the fire area and the total CDF and LERF for each unit meet the acceptance criteria of RG 1.174 (SE Section 3.4.6) and 2) adequate DID and SMs are maintained for each fire area (SE Section 3.4.2). This determination relies on the following fire protection systems and features to meet the acceptance criteria:
| |
| * All VFDRs require reliance on the general area and/or hazard detection associated with existing fire detection in the RB3 to meet the DID criteria.
| |
| * For VFDRs RB3-02, RB3-09, RB3-11, and RB3-12, in addition to the reliance on existing fire detection, recovery actions are identified to monitor alternative instrumentation as stated in Attachment G Table G-2 of the LAR and are identified as relied upon to meet DID criteria (SE Section 2.9; Item 14).
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| * For VFDRs RB3-10 and RB3-16, in addition to reliance on existing fire detection, operator guidance will be inserted into shutdown procedures for operation of RC high point vent valves for RC letdown in the event that head vent valve flow path becomes inoperable (SE Section 2.9; Item 30).
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 267 Component VFDR# VFDR Description (Cables)
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| This normally closed, required closed valve provides train separation by isolating the cross connect header between the 3A and 3B trains of Emergency Feedwater. 3CCWVA0269 - SG RB3-01 Fire induced cable damage may result in spurious opening of this valve, a A FDW Contml I diversion of flow to either the 3A or 3B SGs, and a challenge to the DHR Nuclear MOV I Safety Performance Criteria. This valve may suffer IN 92-18 damaQe.
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| SG level indication is required for process monitoring and diagnosis of plant transients. Fire induced cable damage may result in loss of SG level indication 3FDWP 0270, RB3-02 resulting in the inability of the operator to monitor and contro/level in either the 3A 2FOWP 0271 - SG or 3B SGs from the MCR and challenge the Process Monitoring Nuclear Safety Level Indications I
| |
| Performance Criterion.
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| This normally open, required open valve is located in the EFW flow path to the 3B SG. Fire induced cable damage may result in spurious closing of this valve, isolating Protected Service Water flow to the 3B SG. The SUbsequent decrease in 3FDWVA0347 - SG RB3-03 I SG shell temperature may result in 3B SG exceeding its tube to shell differential B Inlet MOV temperature limit. This could challenge the Inventory Control and OHR Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 damaQe.
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| Normally open valve 3HP VA0023 is in the flow path from the LOST to the suction of the credited HPI pump. Normally closed valve 2HP VA0939 isolates the flow path from the LOST to the containment sump. Recirculation flow to the LOST 3HP VA0023 - HPJ during prolonged operation of the HPI pump at low flow conditions may result in an Normal Suction increase in temperature of LOST contents to the operability limit of the HPI pump.
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| MOV, 3HP VA0939 RB3-04 The contents of the LOST must be diverted to the containment sump by opening
| |
| - LOST to 3HP VA0939 and closing 3HP VA0023 prior to the operability limit of the HPI pump Emergency Sump being exceeded to prevent challenging the Reactivity, Inventory and Pressure MOV Control Nuclear Safety Performance Criteria. Although unaffected by fire, the power supplies for these valves are not credited following a fire in this fire area and a loss of power may prevent these valves from beina repositioned.
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| These normally open, required closed valves isolate the flow path from the BWST to the LPI Pumps, RBS Pumps, and containment sump. Although unaffected by fire, the power supplies to these valves are not credited following a fire in this fire area and a loss of power may prevent these valves from being closed and result in 3LP VA0021, 3LP I
| |
| RB3-06 a diversion of BWST inventory to the containment sump via the LPI system. In VA0022 - BWST addition, an inadvertent ES actuation could result in a diversion of BWST inventory Suction MOVs to the containment sump via the RBS system. A loss of BWST inventory could challenge the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria.
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| Source range flux indication is required for process monitoring and control. Fire 3RPSP 1007, induced cable damage may result in loss of source range flux indications resulting RB3-07 3RPSP 1008 in the inability of the operator to monitor this parameter from the MCR and I Source Range Flux challenQe the Process MonitorinQ Nuclear Safety Performance Criterion.
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| 3MS VA0017, 3MS VA0024,3MS VA0026,3MS These normally open, required closed valves isolate flow paths from the MSHs . VA0033, 3MS Although unaffected by fire, the power supplies to these valves are not credited VA0035,3MS RB3-08 following a fire in this fire area and a loss of power may prevent these valves from VA0036,3MS being closed and could result in overcooling and shrinkage of RC inventory. This VA0076,3MS could challenge the OHR Nuclear Safety Performance Criterion. VA0079,3MS VA0082, 3MS VA0084- SG Isolation MOVs RC pressure indication is required for process monitoring and diagnosis of plant transients. Fire induced cable damage may result in loss of RC pressure 3RC CR0045 - RC RB3-09 indications resulting in the inability of the operator to monitor and control this Pressure Recorder parameter from the MCR and challenge the Process Monitoring Nuclear Safety I I Performance Criterion.
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| Operation of the pressurizer heaters is required to maintain control of RC pressure. 3RC HE0001, 3RC RB3-10 Fire damage to cables may result in a loss of the pressurizer heaters and could HEOO02,3RC challenae the Pressure Control Nuclear Safety Performance Criterion. HEOO03,3RC OFFICIAL USE ONLY SECURITY RELP.TED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 268 Component VFDR# VFDR Description (Cables)
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| HE0004 Pressurizer Heaters Pressurizer level indication is required for process monitoring and diagnosis of plant transients. Fire impingement on instrument sensing lines or fire induced 3RC P 0365 RB3-11 cable damage may result in loss of pressurizer level indication resulting in the Pressurizer Level inability of the operator to monitor and control this parameter from the MCR and Indication challenQe the Process MonitorinQ Nuclear Safety Performance Criterion.
| |
| RC temperature indication is required for process monitoring and diagnosis of plant transients. Fire induced cable damage may result in loss of RC temperature 3RC P 0376 - RC RB3-12 indication resulting in the inability of the operator to monitor and control this Temperature parameter from the MCR and challenge the Process Monitoring Nuclear Safety Indication Performance Criterion.
| |
| 3RC SXTRN001, Pressurizer heaters are required for RC pressure control. The heaters receive 3RC SXTRN002, non-credited power from Unit 3 and credited power from a PSW system power 3RC SXTRN003 RB3-13 supply. The transfer of credited power to the pressurizer heaters requires a Pressurizer Heaters recovery action. Failure to transfer credited power to the heaters could challenge PSW Power the Pressure Control Nuclear Safety Performance Criterion.
| |
| Transfer Switches This normally closed, required closed valve isolates the flow path from the RCS to 3RC VA0066 the Quench Tank. Fire induced cable damage may result in spurious opening of Pressurizer Power RB3-14 the PORV causing a loss of inventory and RC sUbcooling. This could challenge Operated Relief the Inventory and Pressure Control Nuclear Safety Performance Criteria. Valve These normally closed, required closed valves isolate flow paths from the RCS to 3RC VA0155, 3RC containment. Potential hot shorts within the electrical penetration box may VA0157,3RC RB3-15 spuriously open the reactor head vent and hot leg vent valves. The spurious VA0159 - RC Hot opening of these valves may result in a loss of RC inventory and challenge the Leg and Head Vent Inventory and Pressure Control Nuclear Safety Performance Criteria. Valves These normally closed valves isolate the flow path from the RCS to containment.
| |
| These valves are required opened to provide an RC letdown flow path. Fire 3RC VA0159, 3RC RB3-16 induced cable damage may prevent these valves from being opened resulting in VA0160 - RC Head the lifting of the pressurizer safety relief valves and a challenge to the Inventory Vent Valves Control Nuclear Safety Performance Criterion.
| |
| Although unaffected by fire, the power supplies for the station HVAC system are not credited following a fire in this fire area and a loss of power may result in the Unit 3 Control RB3-18 temperature inside the Unit 3 control complex exceeding the operability limit of Complex Cooling SSD components and challenge the Vital Auxiliaries Nuclear Safety Performance Criterion.
| |
| Fire damage to cables may result in a loss of power to the containment cooling system and may result in the temperature inside the Unit 3 RB exceeding the Unit 3 Containment RB3-19 operability limit of SSD components. This could challenge the Vital Auxiliaries Cooling Nuclear Safety Performance Criterion.
| |
| Note: The additional risk added because of these VFDRs, as determined from the FRE for this fire area, is provided in SE Table 3.5.
| |
| Recovery Actions (RAs)
| |
| Recovery actions credited in this fire area to satisfy the DID requirements of NFPA 805, Section 4.2.4.2, are provided in the following table:
| |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 269 Component Component Name Description of Action 10 U3 SSF SG 1B LEVEL For a fire in the east side of containment, dispatch an operator to the SSF 3FDWP 0232 INDICATION to monitor instrument 3FDWP 0232.
| |
| U3 SSF For a fire in the east side of containment, monitor 3RC P 0365 from the 3RC P 0233 PRESSURIZER control room if available; if not dispatch an operator to the SSF to monitor LEVEL INDICATION instrument 3RC P 0233.
| |
| U3 SSF RC LOOP B For a fire in the east side of containment, dispatch an operator to the SSF 3RC P 0238 PRESSURE to monitor instrument 3RC P 0238.
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| INDICATION REACTOR OUTLET For a fire in the east side of containment, dispatch an operator to the SSF 3RC P 0315 LOOP B to monitor instrument 3RC P 0315.
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| Note: The FRE for this fire area determined that the additional risk being added because of these RAs was negligible for both change in CDF and change in LERF. See SE Section 3.4.2 for a detailed discussion of the NRC staff's review of the FREs.
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| Fire Detection & Suppression Systems Required to Meet the Nuclear Safety Performance Criteria The licensee performed an evaluation of the fire detection and suppression systems in this area.
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| The results of the evaluation were documented in LAR Table 4-4 and the applicable portion has been included below.
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| Suppression Detection Detection Required Required Auto Provided? System?
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| Fire Fire Zone System?
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| Suppression Area Zone Description Provided?
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| E R 0 S E R 0 S Unit 3 Reactor RB3 124 No No No No No Yes No No Yes No Building Legend:
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| E - EEEElLA: Systems required for acceptability of EEE Evaluations I NRC-approved Licensing Action (Section 2.2.7)
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| R - Risk: Systems required to meet the Risk Criteria for the PB Approach (Section 4.2.4) o - Defense-in-Depth: Systems required to maintain adequate balance of Defense-in-Depth for a PB Approach (Section 4.2.4.2)
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| S - Separation Criteria: Systems required for Chapter 4 Separation Criteria in (Section 4.2.3)
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| MR - Modification Required Systems are committed to be modified as indicated in Table 4-4 and Attachment S of LAR Fire Area RB3 Conclusion The licensee has utilized the FRE PB approach to demonstrate the ability to meet the NFPA 805 nuclear safety performance criteria for this fire area. A FRE in accordance with NFPA 805, Section 4.2.4.2, in conjunction with deterministic methods for simplifying assumptions, was used in applying this approach.
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| Based on the information provided in the LAR, as supplemented, the NRC staff finds Fire Area RB3 meets the nuclear safety goals, objectives, and performance criteria of NFPA 805. This conclusion is based on the following:
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| * Fire protection SSCs were evaluated in accordance with NFPA 805, Chapter 4 to determine which, if any, were required to meet the nuclear safety performance criteria. This evaluation included:
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 270
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| : a. The fire protection detection systems required to meet the nuclear safety performance criteria were documented.
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| : b. Fire Area boundaries were defined using three hour rated walls, ceilings and floors, including fire barriers, fire barrier penetrations, and through penetration fire stops and spatial separation.
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| * One exemption from the pre-transition fire protection requirements was evaluated and found to be valid and applicable under the NFPA 805 RIIPB FPP.
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| * Seventeen VFDRs were identified, evaluated through the performance of a FRE, and found to meet the risk acceptance criteria, as well as the requirements for DID and SMs. The acceptability of the risk for this fire area is contingent on the risk reduction from the planned PSW modification (see SE Section 3.4 for a detailed discussion of the NRC staff's review of the adequacy of the FRE method used at ONS).
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| * Four recovery actions were identified and evaluated for the additional risk (change in CDF and change in LERF) each poses. The additional risk of each recovery action was conservatively estimated to be taken as the change in CDF and change in LERF associated with the VFDR that resulted in the need for the recovery action. The change in CDF and change in LERF for each recovery action was determined to be negligible.
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| * No modifications were identified as necessary for meeting requirements of NFPA 805.
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| * Two VFDR dispositions, in addition to reliance on existing fire detection, require operator guidance to be inserted into shutdown procedures for operation of RC high point vent valves for RC letdown in the event that head vent valve flow path becomes inoperable.
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| Fire Area SSF, Standby Shutdown Facility The licensee analyzed this fire area using the FRE approach in accordance with NFPA 805 Section 4.2.4.2, but also used deterministic simplifying assumptions in order to credit those portions of the facility design that met the deterministic requirements of NFPA 805 Section 4.2.3. The licensee identified the SSCs necessary to meet the nuclear safety performance criteria in this fire area.
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| Evaluation of Fire Suppression Effects on Nuclear Safety Performance Criteria The licensee stated in Attachment C, "NEI 04-02, Table B-3, Fire Area Transition," that safe and stable conditions can be achieved and maintained using equipment and cables outside of the area of fire suppression activity. Flooding of the suppression areas and discharge of suppression water to adjacent compartments is controlled and will not jeopardize achievement of safe and stable conditions. Based on the information provided by the licensee in the NFPA 805 LAR, the NRC staff finds the licensee's evaluation of fire suppression effects on nuclear safety performance criteria acceptable because the results of the licensee's analysis indicate that fire suppression activities will not adversely affect achievement of the nuclear safety performance criteria.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATIO~J 271 Exemptions and Other Licensing Actions The licensee did not credit any previously approved licensing actions or exemptions from the existing fire protection requirements.
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| Variation from Deterministic Requirements (VFDRs)
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| Fire Area SSF has a total of 32 VFDRs, which are provided in the table below. All of these VFDRs are variances from NFPA 805, Section 4.2.3 (separation issues) that were dispositioned with a FRE (SE Section 3.4.3). The licensee's FRE determined that these variances are acceptable based on 1) the change in CDF and LERF for the fire area and the total CDF and LERF for each unit meet the acceptance criteria of RG 1.174 (SE Section 3.4.6) and 2) adequate DID and SMs are maintained for each fire area (SE Section 3.4.2). This determination relies on the following fire protection systems and features to meet the acceptance criteria:
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| * General area and/or hazard detection associated with the existing fire detection panel is required to meet the DID criteria.
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| VFDR# VFDR Description Component (Cables)
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| This normally closed, required closed valve provides train separation by isolating the cross connect header between the 1A and 1B trains of Emergency Feedwater.
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| 1CCWVA0269 - SG A FDW SSF-01 Fire induced cable damage may result in spurious opening of this valve, a diversion Control MaV of flow to either the 1A or 1B SGs, and a challenge to the OHR Nuclear Safety Performance Criterion. This valve may suffer IN 92-18 damaQe.
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| This normally closed, required closed valve isolates the flow path between the SSF ASW Pump and the 1B EFW header. Fire induced cable damage may result in 1CCWVA0287 - SSF ASW SSF-02 spurious opening of this valve resulting in a diversion of PSW flow from the 1B Pump to SG Supply MaV EFW header and challenge the OHR Nuclear Safety Performance criterion. This valve may suffer IN 92-18 damage.
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| This normally open, required open valve is located in the EFW flow path to the 1B SG. Fire-induced cable damage may result in spurious closing of this valve, isolating Protected Service Water flow to the 1B SG. The subsequent decrease in 1FDWVA0347 - SG B Inlet SSF-03 SG shell temperature may result in 1B SG exceeding its tube to shell differential MaV temperature limit. This could challenge the Inventory Control and OHR Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 damaQe.
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| Normally open valve 1HP VA0023 is in the flow path from the LOST to the suction of the credited HPI pump. Normally closed valve 1HP VA0939 isolates the flow path from the LOST to the containment sump. Recirculation flow to the LOST during prolonged operation of the HPI pump at low flow conditions may result in an 1HP VA0023 - HPI Normal increase in temperature of LOST contents to the operability limit of the HPI pump.
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| Suction MaV, 1HP VA0939 SSF-04 The contents of the LOST must be diverted to the containment sump by opening LOST to Emergency Sump 1HP VA0939 and closing 1HP VA0023 prior to the operability limit of the HPI pump MaV being exceeded to prevent challenging the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria. Although unaffected by fire, the power supplies for these valves are not credited following a fire in this fire area and a loss of power may prevent these valves from beinQ repositioned.
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| This normally closed, required closed valve isolates the RC letdown flow path to the SFP. Fire induced cable damage may result in spurious opening of this valve 1HP VA0426 - RC Letdown to SSF-05 resulting in a loss of RC inventory from the RCS to the SFP. This could challenge SFP MaV the Reactivity and Inventory Control Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 damaQe.
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| These normally open, required closed valves isolate the flow path from the BWST to the LPI Pumps, RBS Pumps, and containment sump. Although unaffected by 1LP VA0021, 1LP VA0022 SSF-07 fire, the power supplies to these valves are not credited following a fire in this fire BWST Suction MaVs area and a loss of power may prevent these valves from being closed resulting in excess BWST inventory loss to the containment sump and challenqe the Reactivity, OFFICIAL USE ONLY SECURITY RELJ\TED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 272 VFDR# VFDR Description Component (Cables)
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| Inventory and Pressure Control Nuclear Safety Performance Criteria.
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| 1MS VA0017, 1MS VAOO24, These normally open, required closed valves isolate flow paths from the MSHs .
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| 1MS VA0026, 1MS VAOO33, Although unaffected by fire, the power supplies to these valves are not credited 1MS VA0035, 1MS VAOO36, SSF-08 following a fire in this fire area, and a loss of power may prevent these valves from 1MS VA0076, 1MS VAOO79, being closed and could result in overcooling and shrinkage of RC inventory. This 1MS VA0082, 1MS VAOO84 could challenge the OHR Nuclear Safety Performance Criterion.
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| SG Isolation MaVs Pressurizer heaters are required for RC pressure control. Groups C & 0 of pressurizer heater bank 2 can be controlled from the SSF. Fire damage to cables 1RC HE0002 - Pressurizer may result in the spurious operation of these heaters resulting in an increase in RC SSF-09 Heater Bank 2 (Groups C &
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| pressure which may challenge the Pressure Control Nuclear Safety Performance 0)
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| Criterion.
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| Pressurizer heaters are required for RC pressure control. The heaters receive non-1RC SXTRN001, 1RC credited power from Unit 1 and credited power from a PSW system power supply.
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| SXTRN002 - Pressurizer SSF-11 The transfer of credited power to the pressurizer heaters requires a recovery action.
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| Heaters PSW Power Transfer Failure to transfer credited power to the heaters could challenge the Pressure Switches Control Nuclear Safety Performance Criterion This normally closed, required closed valve provides train separation by isolating the cross connect header between the 2A and 2B trains of Emergency Feedwater.
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| 2CCWVA0269 - SG A FDW SSF-12 Fire induced cable damage may result in spurious opening of this valve, a diversion Control MaV of flow to either the 2A or 2B SGs, and a challenge to the OHR Nuclear Safety Performance Criterion. This valve may suffer IN 92-18 damaQe.
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| This normally closed, required closed valve isolates the flow path between the SSF auxiliary service water (ASW) Pump and the 2B EFW header. Fire induced cable 2CCWVA0287 - SSF ASW SSF-13 damage may result in spurious opening of this valve resulting in a diversion of PSW Pump to SG Supply MaV flow from the 2B EFW header and challenge the OHR Nuclear Safety Performance criterion. This valve may suffer IN 92-18 damage.
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| This normally open, required open valve is located in the EFW flow path to the 2B SG. Fire induced cable damage may result in spurious closing of this valve, isolating Protected Service Water flow to the 2B SG. The subsequent decrease in 2FDWVA0347 - SG B Inlet SSF-14 SG shell temperature may result in 2B SG exceeding its tube to shell differential MaV temperature limit. This could challenge the Inventory Control and OHR Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 damaQe.
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| Normally open valve 2HP VA0023 is in the flow path from the LOST to the suction of the credited HPI pump. Normally closed valve 2HP VA0939 isolates the flow path from the LOST to the containment sump. Recirculation flow to the LOST during prolonged operation of the HPI pump at low flow conditions may result in an 2HP VA0023 - HPI Normal increase in temperature of LOST contents to the operability limit of the HPI pump.
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| Suction MaV, 2HP VA0939 SSF-15 The contents of the LOST must be diverted to the containment sump by opening LOST to Emergency Sump 2HP VA0939 and closing 2HP VA0023 prior to the operability limit of the HPJ pump MaV being exceeded to prevent challenging the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria. Although unaffected by fire, the power supplies for these valves are not credited following a fire in this fire area and a loss of power may prevent these valves from beinQ repositioned.
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| This normally closed, required closed valve isolates the RC letdown flow path to the SFP. Fire induced cable damage may result in spurious opening of this valve 2HP VA0426 - RC Letdown to SSF-16 resulting in a loss of RC inventory from the RCS to the SFP. This could challenge SFP MaV the Reactivity and Inventory Control Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 damaQe.
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| These normally open, required closed valves isolate the flow path from the BWST to the LPI Pumps, RBS Pumps, and containment sump. Although unaffected by fire, the power supplies to these valves are not credited following a fire in this fire 2LP VA0021, 2LP VAOO22 SSF-18 area and a loss of power may prevent these valves from being closed resulting in BSWT Suction MaVs excess BWST inventory loss to the containment sump and challenge the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria.
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| These normally open, required closed valves isolate flow paths from the MSHs . 2MS VA0017, 2MS VAOO24, Although unaffected by fire, the power supplies to these valves are not credited 2MS VA0026, 2MS VAOO33, SSF-19 following a fire in this fire area, and a loss of power may prevent these valves from 2MS VA0035, 2MS VAOO36, beinQ closed and could result in overcoolinQ and shrinkaQe of RC inventory. This 2MS VA0076, 2MS VAOO79, OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 273 VFDR# VFDR Description Component (Cables) could challenge the DHR Nuclear Safety Performance Criterion. 2MS VA0082, 2MS VAOO84 SG Isolation MaVs Pressurizer heaters are required for RC pressure control. Groups C & D of pressurizer heater bank 2 can be controlled from the SSF. Fire damage to cables 2RC HE0002 - Pressurizer SSF-20 may result in the spurious operation of these heaters resulting in an increase in RC Heater Bank 2 (Groups C &
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| pressure which may challenge the Pressure Control Nuclear Safety Performance D)
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| Criterion.
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| Pressurizer heaters are required for RC pressure control. The heaters receive non-credited power from Unit 2 and credited power from a PSW system power supply. 2RC SXTRN001, 2RC The transfer of credited power to the pressurizer heaters requires a recovery action. SXTRN002, 2RC SXTRNOO3 SSF-22 Failure to transfer credited power to the heaters could challenge the Pressure - Pressurizer Heaters PSW Control Nuclear Safety Performance Criterion. Power Transfer Switches This normally closed, required closed valve provides train separation by isolating the cross connect header between the 3A and 3B trains of Emergency Feedwater.
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| 3CCWVA0269 - SG A FDW SSF-23 Fire induced cable damage may result in spurious opening of this valve, a diversion Control MaV of flow to either the 3A or 3B SGs, and a challenge to the DHR Nuclear Safety Performance Criterion. This valve may suffer IN 92-18 damaoe.
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| This normally closed, required closed valve isolates the flow path between the SSF ASW Pump and the 3B EFW header. Fire induced cable damage may result in 3CCWVA0287 - SSF ASW SSF-24 spurious opening of this valve resulting in a diversion of PSW flow from the 3B Pump to SG Supply MaV EFW header and challenge the DHR Nuclear Safety Performance criterion. This valve may suffer IN 92-18 damaQe.
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| This normally open, required open valve is located in the EFW flow path to the 3B SG. Fire induced cable damage may result in spurious closing of this valve, isolating Protected Service Water flow to the 3B SG. The subsequent decrease in 3FDWVA0347 - SG B Inlet SSF-25 SG shell temperature may result in 3B SG exceeding its tube to shell differential MaV temperature limit. This could challenge the Inventory Control and DHR Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 damaQe Normally open valve 3HP VA0023 is in the flow path from the LDST to the suction of the credited HPI pump. Normally closed valve 3HP VA0939 isolates the flow path from the LDST to the containment sump. Recirculation flow to the LDST during prolonged operation of the HPI pump at low flow conditions may result in an 3HP VA0023 - HPI Normal increase in temperature of LDST contents to the operability limit of the HPI pump.
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| Suction MaV, 3HP VA0939 SSF-26 The contents of the LDST must be diverted to the containment sump by opening LDST to Emergency Sump 3HP VA0939 and closing 3HP VA0023 prior to the operability limit of the HPJ pump MaV being exceeded to prevent challenging the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria. Although unaffected by fire, the power supplies for these valves are not credited following a fire in this fire area and a loss of power may prevent these valves from being repositioned.
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| This normally closed, required closed valve isolates the RC letdown flow path to the SFP. Fire induced cable damage may result in spurious opening of this valve 3HP VA0426 - RC Letdown to SSF-27 resulting in a loss of RC inventory from the RCS to the SFP. This could challenge SFP MaV the Reactivity and Inventory Control Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 damaoe.
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| These normally open, required closed valves isolate the flow path from the BWST to the LPI Pumps, RBS Pumps, and containment sump. Although unaffected by fire, the power supplies to these valves are not credited following a fire in this fire 3LP VA0021, 3LP VAOO22 SSF-29 area and a loss of power may prevent these valves from being closed resulting in BSWT Suction MaVs excess BWST inventory loss to the containment sump and challenge the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria.
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| 3MS VA0017, 3MS VAOO24, These normally open, required closed valves isolate flow paths from the MSHs .
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| 3MS VA0026, 3MS VAOO33, Although unaffected by fire, the power supplies to these valves are not credited 3MS VA0035, 3MS VAOO36, SSF-30 following a fire in this fire area, and a loss of power may prevent these valves from 3MS VA0076, 3MS VAOO79, being closed and could result in overcooling and shrinkage of RC inventory. This 3MS VA0082, 3MS VAOO84 could challenge the DHR Nuclear Safety Performance Criterion.
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| SG Isolation MaVs Pressurizer heaters are required for RC pressure control. Groups C & D of 3RC HE0002 - Pressurizer SSF-31 pressurizer heater bank 2 can be controlled from the SSF. Fire damage to cables Heater Bank 2 (Groups C &
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| may result in the spurious operation of these heaters resultino in an increase in RC D)
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 274 VFDR# VFDR Description Component (Cables) pressure which may challenge the Pressure Control Nuclear Safety Performance Criterion.
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| Pressurizer heaters are required for RC pressure control. The heaters receive non-credited power from Unit 3 and credited power from a PSW system power supply. 3RC SXTRN001 , 3RC The transfer of credited power to the pressurizer heaters requires a recovery action. SXTRN002, 3RC SXTRNOO3 SSF-33 Failure to transfer credited power to the heaters could challenge the Pressure - Pressurizer Heaters PSW Control Nuclear Safety Performance Criterion. Power Transfer Switches Although unaffected by fire, the power supplies for the station HVAC system are not credited following a fire in this fire area and a loss of power may result in the Units 1 & 2 Control Complex SSF-35 temperature inside the Units 1 & 2 control complex exceeding the operability limit of Cooling SSD components and challenge the Vital Auxiliaries Nuclear Safety Performance Criterion.
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| Although unaffected by fire, the power supplies for the station HVAC system are not credited following a fire in this fire area and a loss of power may result in the Unit 3 Control Complex SSF-36 temperature inside the Unit 3 control complex exceeding the operability limit of SSD Cooling I components and challenge the Vital Auxiliaries Nuclear Safety Performance Criterion.
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| Fire damage to cables may result in a loss of power to the containment cooling system and may result in the temperature inside the Unit 3 RB exceeding the SSF-37 Unit 1 Containment Cooling operability limit of SSD components. This could challenge the Vital Auxiliaries Nuclear Safety Performance Criterion Fire damage to cables may result in a loss of power to the containment cooling system and may result in the temperature inside the Unit 3 RB exceeding the SSF-38 Unit 2 Containment Cooling operability limit of SSD components. This could challenge the Vital Auxiliaries Nuclear Safety Performance Criterion Fire damage to cables may result in a loss of power to the containment cooling system and may result in the temperature inside the Unit 3 RB exceeding the SSF-39 Unit 3 Containment Cooling operability limit of SSD components. This could challenge the Vital Auxiliaries Nuclear Safety Performance Criterion.
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| Note: The additional risk added because of these VFDRs, as determined from the FRE for this fire area, is provided in SE Table 3.5.
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| Recovery Actions (RAs)
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| The licensee did not identify any recovery actions required for this fire area.
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| Fire Detection & Suppression Systems Required to Meet the Nuclear Safety Performance Criteria The licensee performed an evaluation of the fire detection and suppression systems in this area.
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| The results of the evaluation were documented in LAR Table 4-4 and the applicable portion has been included below.
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| OFFICIAL USE ONLY SECURITY RELATED INfORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 275 Suppression Detection Detection Required Required Auto Provided? System?
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| Fire Fire System?
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| Zone Description Suppression Area Zone Provided?
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| E R D S E R D S Standby Shutdown SSF SSF Yes No No No No Yes No No Yes No Facilitv Legend:
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| E - EEEE/LA: Systems required for acceptability of EEE Evaluations I NRC approved Licensing Action (Section 2.2.7)
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| R - Risk: Systems required to meet the Risk Criteria for the PB Approach (Section 4.2.4)
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| D - Defense-in-Depth: Systems required to maintain adequate balance of Defense-in-Depth for a PB Approach (Section 4.2.4.2)
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| S - Separation Criteria: Systems required for Chapter 4 Separation Criteria in (Section 4.2.3)
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| MR - Modification Required Systems are committed to be modified as indicated in LAR Table 4-4 and Attachment S Fire Area SSF Conclusion The licensee has utilized the FRE PB approach to demonstrate the ability to meet the NFPA 805 nuclear safety performance criteria for this fire area. A FRE in accordance with NFPA 805, Section 4.2.4.2, in conjunction with deterministic methods for simplifying assumptions, was used in applying this approach.
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| Based on the information provided in the LAR, as supplemented, the NRC staff finds Fire Area SSF meets the nuclear safety goals, objectives, and performance criteria of NFPA 805. This conclusion is based on the following:
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| * Fire protection SSCs were evaluated in accordance with NFPA 805 Chapter 4 to determine which, if any, were required to meet the nuclear safety performance criteria. This evaluation included:
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| : a. The fire protection detection systems required to meet the nuclear safety performance criteria were documented.
| |
| : b. Fire Area boundaries were defined using three hour rated walls, ceilings and floors, including fire barriers, fire barrier penetrations, and through penetration fire stops and spatial separation.
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| * No exemptions or licensing actions from the pre-transition fire protection requirements were required.
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| * Thirty-two VFDRs were identified, evaluated through the performance of a FRE, and found to meet the risk acceptance criteria, as well as the requirements for DID and SMs. The acceptability of the risk for this fire area is contingent on the risk reduction from the planned PSW modification (see SE Section 3.4 for a detailed discussion of the NRC staff's review of the adequacy of the FRE method used at ONS).
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| * This fire area did not require the use of recovery actions to meet the nuclear safety performance criteria.
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| * No modifications were identified as necessary for meeting requirements of NFPA 805.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 276 Fire Area TB, Turbine Building The licensee analyzed this fire area using the FRE approach in accordance with NFPA 805, Section 4.2.4.2, but also used deterministic simplifying assumptions in order to credit those portions of the facility design that met the deterministic requirements of NFPA 805, Section 4.2.3. The licensee identified the SSCs necessary to meet the nuclear safety performance criteria in this fire area.
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| Evaluation of Fire Suppression Effects on Nuclear Safety Performance Criteria The licensee stated in Attachment C, "NEI 04-02, Table B-3, Fire Area Transition," that safe and stable conditions can be achieved and maintained using equipment and cables outside of the area of fire suppression activity. Flooding of the suppression areas and discharge of suppression water to adjacent compartments is controlled and will not jeopardize achievement of safe and stable conditions. Based on the information provided by the licensee in the NFPA 805 LAR, the NRC staff finds the licensee's evaluation of fire suppression effects on nuclear safety performance criteria acceptable because the results of the licensee's analysis indicate that fire suppression activities will not adversely affect achievement of the nuclear safety performance criteria.
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| Exemptions and Other Licensing Actions The licensee did not credit any previously approved licensing actions or exemptions from the existing fire protection requirements.
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| Variation from Deterministic Requirements (VFDRs)
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| Fire Area TB has a total of 45 VFDRs, which are provided in the table below. All but one of these VFDRs are variances from NFPA 805, Section 4.2.3, (separation issues) that were dispositioned with a FRE (SE Section 3.4.3). The licensee's FRE determined that these variances are acceptable based on 1) the change in CDF and LERF for the fire area and the total CDF and LERF for each unit meet the acceptance criteria of RG 1.174 (SE Section 3.4.6) and 2) adequate DID and SMs are maintained for each fire area (SE Section 3.4.2). This determination relies on the following fire protection systems and features to meet the acceptance criteria:
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| * General area and/or hazard detection for the TB Fire Area is required to meet the risk acceptance criteria. The Fire PRA makes assumptions regarding the time of fire discovery, fire brigade notification, and brigade manual suppression. These assumptions determine the impact of the fire, including the likelihood of a HGL being formed in the compartment. Specifically, the Fire PRA is based on a fire brigade response time of 20 minutes or less. The existing fire zone detection system coverage of the general area and/or hazard necessary for this assumption to be valid was not considered sufficient to conservatively meet the risk criteria. Therefore, modifications to the fire detection system in the TB Fire Area are required to support the fire risk analysis assumption of 20 minute brigade response time.
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| * General area and/or hazard detection associated with the Auxiliary Shutdown Panels FZ 39 and 41 is required to meet the DI D criteria.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 277 Based on the reliance on fire detectors in the TB Fire Area to meet the risk and DID criteria, the licensee has committed to make modifications to the fire detection system, which may include fire detector upgrades and/or new installation. Improvements to the following TB Fire Zones for general area and/or fire hazard detection are required: 3, 6, 12, 15, 19, 24, 25, 28, 29, 32, 33, 33A, 34, 34A, 35, 37, 39, 39A, and 41 (SE Section 2.8.1).
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| One of the 45 VFDRs, TB-06, is a variance from the deterministic requirements of NFPA 805, Section 4.2.3 (separation issue) that will be corrected with a plant modification. According to the LAR, the wall separatrng the TB and the AB is not currently a three hour rated wall as required by NFPA 805, Section 3.11.1, and all of the penetrations in the wall do not have a fire resistance rating as required by NFPA 805, Section 3.11.3. This wall is credited for area separation using the deterministic approach of NFPA 805, Section 4.2.3. The licensee has committed to make modifications to the wall to bring it into compliance with the requirements of NFPA 805 (SE Section 2.8.1).
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| VFDR# VFDR Description Component (Cables)
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| The Main Feedwater (MFW) Pumps are required to be off to isolate MFW to the SGs. Fire damage to cables may result in an inability to 1FOWPUOO01, secure the pumps or may result in a spurious pump start. Spurious TB-01 1FOWPU0002 - Main operation of the MFW Pumps could result in overfill of the SGs, Feedwater Pumps overcooling of the RCS and a challenge to the OHR Nuclear Safety Performance Criterion.
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| The EFW Pumps are required to be off to isolate EFW to the SGs. Fire 1FOWPUOO03, damage to cables may result in an inability to secure the pumps or may 1FOWPUOO04, TB-02 result in a spurious pump start. Spurious operation of the EFW Pumps 1FOWPU0005 - EFW could result in overfill of the SGs, overcooling of the RCS and a Pumps challenQe to the OHR Nuclear Safety Performance Criterion.
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| Normally open valve 1HP VA0023 is in the flow path from the LOST to the suction of the credited HPI pump. Normally closed valve 1HP VA0939 isolates the flow path from the LOST to the containment sump.
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| Recirculation flow to the LOST during prolonged operation of the HPI pump at low flow conditions may result in an increase in temperature of LOST contents to the operability limit of the HPI pump. The contents of 1HP VA0023 - HPI Normal the LOST must be diverted to the containment sump by opening 1HP Suction MOV, 1HP TB-03 VA0939 and closing 1HP VA0023 prior to the operability limit of the HPI VA0939 - LOST to pump being exceeded to prevent challenging the Reactivity, Inventory Emergency Sump MOV and Pressure Control Nuclear Safety Performance Criteria. Fire damage to cables for electrical equipment supplying power to these valves may prevent the valves from being repositioned. Fire damage to cables for
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| 'IHP VA0939 may cause this valve to spuriously open prematurely, fail to open or spuriously close once opened. Valve 1HP VA0939 may suffer IN 92-18 damaQe.
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| This normally closed valve isolates the flow path from the BWST to the suction of the HPI Pumps. This valve is required to be open to supply borated water from the BWST to the HPI pump for RC boration and inventory control, and seal injection to the RCPs. Prior to the transfer of 1HP VA0024 - HPI BWST TB-04 the power supply to the PSW system, fire damage to cables may prevent Suction MOV this valve from opening or cause the valve to spuriously close and challenge the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 damage. I This normally throttled, required closed valve provides normal RC makeup flow for pressurizer level control. This valve is normally 1HP VA0120 - RC Volume TB-05 controlled from the main control room, but an alternate control station is Control AOV provided in the Auxiliary Shutdown Panel located on the operatinQ floor OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 278 VFDR# VFDR Description Component (Cables) of the TB fire area. Fire induced cable damage may result in spurious opening of this valve, resulting in an uncontrolled increase in RC inventory and challenge the Inventory and Pressure Control Nuclear Safety Performance Criteria.
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| The wall separating the TB and AB is not three hour rated as required by NFPA 805, Section 3.11.1 and all the penetrations in the wall do not TB-06 have a fire resistance rating as required by NFPA 805, Section 3.11.3. TB / AB Wall This wall is credited for area separation in the deterministic approach of NFPA 805, Section 4.2.3.
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| This normally closed, required closed valve isolates the flow path from the LOST to the containment sump. This valve is required to remain closed to prevent diverting BWST inventory to the containment sump via HPI pump recirculation to the LOST. Fire damage to cables for electrical 1HP VA0940 - LOST to TB-07 equipment supplying power to this valve may cause this valve to Emergency Sump MOV spuriously open and challenge the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 damage.
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| The 1C High Pressure Injection Pump is required to be off to prevent an uncontrolled increase in RC inventory. Fire damage to cables may result 1HPIPU0003 - HPI Pump TB-08 in spurious pump start and challenge the Inventory and Pressure Control 1C Nuclear Safety Performance Criteria.
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| These normally open, required closed valves isolate the flow path from the BWST to the LPI Pumps, RBS Pumps, and containment sump. Fire damage to cables may prevent these valves from being closed or cause 1LP VA0021, 1LP VA0022 TB-09 them to spuriously open resulting in excess BWST inventory loss to the
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| - BWST Suction MOVs containment sump and a challenge to the Reactivity and Inventory Control Nuclear Safety Performance Criteria. These valves may suffer IN 92-18 damaqe.
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| 1MS VA0017 1MS VA0024 These normally open, required closed valves isolate flow paths from the 1MS VA0026 MSHs. Fire damage to cables may prevent these valves from being 'IMS VA0033 closed or may result in spurious opening of the valves. The failure to 1MS VA0035 TB-10 close these valves or the spurious opening of the valves could result in 1MS VA0036 overcooling and shrinkage of RC inventory and challenge the OHR 1MS VA0076 I Nuclear Safety Performance Criterion. These valves may suffer IN 92 1MS VA0079 18 damage 1MS VA0082 1MS VA0084 SG Isolation MOVs Pressurizer heaters are required for RC pressure control. Pressurizer heater bank 2 is normally controlled from the main control room, but an alternate control station is provided in the Auxiliary Shutdown Panel 1RC HE0002 - Pressurizer TB-11 located on the operating floor of the TB fire area. Fire damage to cables Heater Bank 2 (Groups B may result in the spurious operation of this heater reSUlting in an &0) uncontrolled increase in RC pressure which may challenge the Pressure Control Nuclear Safety Performance Criterion.
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| The Reactor Coolant Pumps (RCPs) are required off when SSO is being 1RC PUOO01 accomplished by the PSW system. Fire damage to cables may result in 1RC PUOO02 TB-13 an inability to secure the RCPs or result in a spurious pump start. This 1RC PUOO03 will place the unit in an unanalyzed condition and challenge the OHR 1RC PU0004 - RCPs Nuclear Safety Performance Criterion.
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| Pressurizer heaters are required for RC pressure control. The heaters 1RC SXTRN001, 1RC TB-14 receive non-credited power from Unit 1 and credited power from a PSW SXTRN002 - Pressurizer system power supply. The transfer of credited power to the pressurizer Heaters PSW Power OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 279 VFDR# VFDR Description Component (Cables) heaters requires a recovery action. Failure to transfer credited power to Transfer Switches the heaters could challenge the Inventory and Pressure Control Nuclear Safety Performance Criteria.
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| These normally closed, required closed valves isolate flow paths from 1S0 VA0027, 1S0 the MSHs. Fire damage to cables may result in spurious opening of the VA0290, 1S0 VA0418, valves. The spurious opening of the valves could result in overcooling TB-15 1S0 VA0419, 1S0 and shrinkage of RC inventory. This could challenge the OHR Nuclear VA0420, 1S0 VA0421 -
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| Safety Performance Criterion. These valves may suffer IN 92-18 SG Isolation MOVs damage.
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| The Main Feedwater (MFW) Pumps are required to be off to isolate MFW to the SGs. Fire damage to cables may result in an inability to 2FOWPUOOO1, secure the pumps or may result in a spurious pump start. Spurious TB-16 2FOWPU0002 - Main operation of the MFW Pumps could result in overfill of the SGs, Feedwater Pumps overcooling of the RCS and a challenge to the OHR Nuclear Safety Performance Criterion.
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| The EFW Pumps are required to be off to isolate EFW to the SGs. Fire 2FOWPUOOO3, damage to cables may result in an inability to secure the pumps or may 2FOWPUOOO4, TB-17 result in a spurious pump start. Spurious operation of the EFW Pumps 2FOWPU0005 - EFW could result in overfill of the SGs, overcooling of the RCS and a Pumps challenqe to the OHR Nuclear Safety Performance Criterion.
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| Normally open valve 2HP VA0023 is in the flow path from the LOST to the suction of the credited HPI pump. Normally closed valve 2HP VA0939 isolates the flow path from the LOST to the containment sump.
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| Recirculation flow to the LOST during prolonged operation of the HPI pump at low flow conditions may result in an increase in temperature of LOST contents to the operability limit of the HPI pump. The contents of 2HP VA0023 - HPI Normal the LOST must be diverted to the containment sump by opening 2HP Suction MOV, 2HP TB-18 VA0939 and closing 2HP VA0023 prior to the operability limit of the HPI VA0939 - LOST to pump being exceeded to prevent challenging the ReactiVity, Inventory Emergency Sump IVIOV and Pressure Control Nuclear Safety Performance Criteria. Fire damage to cables for electrical equipment supplying power to these valves may prevent the valves from being repositioned. Fire damage to cables for 2HP VA0939 may cause this valve to spuriously open prematurely, fail to open or spuriously close once opened. Valve 2HP VA0939 may suffer IN 92-18 damaqe.
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| This normally closed valve isolates the flow path from the BWST to the suction of the HPI Pumps. This valve is required to be open to supply borated water from the BWST to the HPI pump for RC boration and inventory control, and seal injection to the RCPs. Prior to the transfer of 2HP VA0024 - HPI BWST TB-19 the power supply to the PSW system, fire damage to cables may prevent Suction MOV this valve from opening or cause the valve to spuriously close and challenge the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 damage.
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| This normally throttled, required closed valve provides normal RC makeup flow for pressurizer level control. This valve is normally controlled from the main control room, but an alternate control station is provided in the Auxiliary Shutdown Panel located on the operating floor 2HP VA0120 - RC Volume TB-20 of the TB fire area. Fire induced cable damage may result in spurious Control AOV opening of this valve, resulting in an uncontrolled increase in RC inventory and challenge the Inventory and Pressure Control Nuclear Safety Performance Criteria.
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| This normally closed, required closed valve isolates the flow path from 2HP VA0940 - LOST to TB-22 the LOST to the containment sump. This valve is required to remain Emergency Sump MOV closed to prevent divertinq BWST inventory to the containment sump via OFFICIAL USE ONLY SECURITY RELATED INFORMJ\TION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 280 VFDR# VFDR Description Component (Cables)
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| HPI pump recirculation to the LDST. Fire damage to cables for electrical equipment supplying power to this valve may cause this valve to spuriously open and challenge the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 damage.
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| The 2C High Pressure Injection Pump is required to be off to prevent an uncontrolled increase in RC inventory. Fire damage to cables may result 2HPIPU0003 - HPI Pump TB-23 in spurious pump start and challenge the Inventory and Pressure Control 2C Nuclear Safety Performance Criteria.
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| These normally open, required closed valves isolate the flow path from the BWST to the LPI Pumps, RBS Pumps, and containment sump. Fire damage to cables may prevent these valves from being closed or cause 2LP VA0021, 2LP VA0022 TB-24 them to spuriously open resulting in excess BWST inventory loss to the
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| - BWST Suction MOVs containment sump and a challenge to the Reactivity and Inventory Control Nuclear Safety Performance Criteria. These valves may suffer IN 92-18 damage.
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| 2MS VA0017, 2MS These normally open, required closed valves isolate flow paths from the VA0024, 2MS VA0026, MSHs. Fire damage to cables may prevent these valves from being 2MS VA0033, 2MS closed or may result in spurious opening of the valves. The failure to VA0035, 2MS VA0036, TB-25 close these valves or the spurious opening of the valves could result in 2MS VA0076, 2MS overcooling and shrinkage of RC inventory and challenge the DHR VA0079, 2MS VA0082, Nuclear Safety Performance Criterion. These valves may suffer IN 92-18 2MS VA0084 - SG damage.
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| Isolation MOVs Pressurizer heaters are required for RC pressure control. Pressurizer heater bank 2 is normally controlled from the main control room, but an alternate control station is provided in the Auxiliary Shutdown Panel 2RC HE0002 - Pressurizer TB-26 located on the operating floor of the TB fire area. Fire damage to cables Heater Bank 2 (Groups B may result in the spurious operation of this heater resulting in an &D) uncontrolled increase in RC pressure which may challenge the Pressure Control Nuclear Safety Performance Criterion.
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| The Reactor Coolant Pumps (RCPs) are required off when SSD is being accomplished by the PSW system. Fire damage to cables may result in 2RC PU0001, 2RC TB-28 an inability to secure the RCPs or result in a spurious pump start. This PU0002, 2RC PUOO03, will place the unit in an unanalyzed condition and challenge the DHR 2RC PU0004 - RCPs Nuclear Safety Performance Criterion.
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| Pressurizer heaters are required for RC pressure control. The heaters 2RC SXTRN001, 2RC receive non-credited power from Unit 2 and credited power from a PSW SXTRN002, 2RC system power supply. The transfer of credited power to the pressurizer TB-29 SXTRN003 - Pressurizer heaters requires a recovery action. Failure to transfer credited power to Heaters PSW Power the heaters could challenge the Inventory and Pressure Control Nuclear Transfer Switches Safety Performance Criteria.
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| These normally closed, required closed valves isolate flow paths from 2SD VA0027, 2SD the MSHs. Fire damage to cables may result in spurious opening of the VA0290, 2SD VA0418, valves. The spurious opening of the valves could result in overcooling TB-30 2SD VA0419, 2SD and shrinkage of RC inventory. This could challenge the DHR Nuclear VA0420, 2SD VA0421 -
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| Safety Performance Criterion. These valves may suffer IN 92-18 SG Isolation MOVs damage.
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| The Main Feedwater (MFW) Pumps are required to be off to isolate MFW to the SGs. Fire damage to cables may result in an inability to 3FDWPUOO01, secure the pumps or may result in a spurious pump start. Spurious TB-31 3FDWPU0002 - Main operation of the MFW Pumps could result in overfill of the SGs, Feedwater Pumps overcooling of the RCS and a challenge to the DHR Nuclear Safety Performance Criterion.
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| OFFICIAL USE O~JLY SECURITY. RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 281 VFDR# VFDR Description Component (Cables)
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| The EFW Pumps are required to be off to isolate EFW to the SGs. Fire damage to cables may result in an inability to secure the pumps or may 3FOWPUOOO3, result in a spurious pump start. Spurious operation of the EFW Pumps 3FOWPUOOO4, TB-32 could result in overfill of the SGs, overcooling of the RCS and a 3FOWPU0005 - EFW challenge to the OHR Nuclear Safety Performance Criterion. Pumps Normally open valve 3HP VA0023 is in the flow path from the LOST to the suction of the credited HPI pump. Normally closed valve 3HP VA0939 isolates the flow path from the LOST to the containment sump.
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| Recirculation flow to the LOST during prolonged operation of the HPI pump at low flow conditions may result in an increase in temperature of 3HP VA0023 - HPI Normal LOST contents to the operability limit of the HPI pump. The contents of Suction MOV, 3HP TB-33 the LOST must be diverted to the containment sump by opening 3HP VA0939 - LOST to VA0939 and closing 3HP VA0023 prior to the operability limit of the HPI Emergency Sump MOV pump being exceeded to prevent challenging the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria. Fire damage to cables for electrical equipment supplying power to these valves may prevent the valves from beinq repositioned.
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| This normally throttled, required closed valve provides normal RC makeup flow for pressurizer level control. This valve is normally controlled from the main control room, but an alternate control station is provided in the Auxiliary Shutdown Panel located on the operating floor 3HP VA0120 - RC Volume TB-35 of the TB fire area. Fire induced cable damage may result in spurious Control AOV opening of this valve, resulting in an uncontrolled increase in RC inventory and challenge the Inventory and Pressure Control Nuclear Safety Performance Criteria.
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| The 3C High Pressure Injection Pump is required to be off to prevent an uncontrolled increase in RC inventory. Fire damage to cables may result 3HPIPU0003 - HPI Pump TB-37 in spurious pump start and challenge the Inventory and Pressure Control 3C Nuclear Safety Performance Criteria.
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| These normally open, required closed valves isolate the flow path from the BWST to the LPI Pumps, RBS Pumps, and containment sump. Fire damage to cables for electrical equipment supplying power to these 3LP VA0021, 3LP VAOO22 TB-38 valves may prevent these valves from being closed resulting in excess - BWST Suction MOVs BWST inventory loss to the containment sump and a challenge to the Inventory and Reactivity Control Nuclear Safety Performance Criteria.
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| 3MS VA0017, 3MS These normally open, required closed valves isolate flow paths from the VA0024, 3MS VAOO26, MSHs. Fire damage to cables may prevent these valves from being 3MS VA0033, 3MS closed or may result in spurious opening of the valves. The failure to VA0035, 3MS VAOO36, TB-39 close these valves or the spurious opening of the valves could result in 3MS VA0076, 3MS overcooling and shrinkage of RC inventory and challenge the OHR VA0079, 3MS VAOO82, Nuclear Safety Performance Criterion. These valves may suffer IN 92-18 3MS VA0084 - SG damage.
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| Isolation MOVs Pressurizer heaters are required for RC pressure control. Pressurizer heater bank 2 is normally controlled from the main control room, but an alternate control station is provided in the Auxiliary Shutdown Panel 3RC HE0002 - Pressurizer TB-40 located on the operating floor of the TB fire area. Fire damage to cables Heater Bank 2 (Groups B may result in the spurious operation of this heater resulting in an &0) uncontrolled increase in RC pressure which may challenge the Pressure Control Nuclear Safety Performance Criterion.
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| The Reactor Coolant Pumps (RCPs) are required off when SSO is being 3RC PU0001, 3RC TB-42 accomplished by the PSW system. Fire damage to cables may result in PU0002, 3RC PUOOO3, an inability to secure the RCPs or result in a spurious pump start. This 3RC PU0004 - RCPs OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 282 VFDR# VFDR Description Component (Cables) will place the unit in an unanalyzed condition and challenge the DHR Nuclear Safety Performance Criterion.
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| Pressurizer heaters are required for RC pressure control. The heaters 3RC SXTRN001, 3RC receive non-credited power from Unit 3 and credited power from a PSW SXTRN002, 3RC system power supply. The transfer of credited power to the pressurizer TB-43 SXTRN003 - Pressurizer heaters requires a recovery action. Failure to transfer credited power to Heaters PSW Power the heaters could challenge the Inventory and Pressure Control Nuclear Transfer Switches Safety Performance Criteria.
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| These normally closed, required closed valves isolate flow paths from 3SD VA0027, 3SD the MSHs. Fire damage to cables may result in spurious opening of the VA0418, 3SD VA0419, valves. The spurious opening of the valves could result in overcooling TB-44 3SD VA0420, 3SD I and shrinkage of RC inventory. This could challenge the DHR Nuclear VA0421 - SG Isolation Safety Performance Criterion. These valves may suffer IN 92-18 MOVs damaqe.
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| Fire damage to cables for electrical eqUipment supplying power to the station HVAC system may result in the temperature inside the Units 1 &
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| Units 1 & 2 Control TB-45 2 control complex exceeding the operability limit of SSD components Complex Cooling located within the control complex and challenge the Vital Auxiliaries Nuclear Safety Performance Criterion.
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| Fire damage to cables for electrical equipment supplying power to the station HVAC system may result in the temperature inside the Unit 3 Unit 3 Control Complex TB-46 control complex exceeding the operability limit of SSD components Cooling located within the control complex and challenge the Vital Auxiliaries Nuclear Safety Performance Criterion.
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| Fire damage to cables for electrical equipment supplying power to the containment cooling system may result in the temperature inside the Unit Unit 1 Containment TB-47 1 RB exceeding the operability limit of SSD components located within Cooling containment and challenge the Vital Auxiliaries Nuclear Safety Performance Criterion.
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| Fire damage to cables for electrical equipment supplying power to the containment cooling system may result in the temperature inside the Unit Unit 2 Containment TB-48 2 RB exceeding the operability limit of SSD components located within Cooling containment and challenge the Vital Auxiliaries Nuclear Safety Performance Criterion.
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| Fire damage to cables for electrical equipment supplying power to the containment cooling system may result in the temperature inside the Unit Unit 3 Containment TB-49 3 RB exceeding the operability limit of SSD components located within Cooling containment and challenge the Vital Auxiliaries Nuclear Safety Performance Criterion.
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| This normally closed valve isolates the flow path from the discharge of the 2A HPI pump to the reactor coolant system (RCS). This valve is required to be open to supply borated water from the BWST to the RCS for RC boration and inventory control. Prior to the transfer of the power 2HP VA0026 - 2A HP TB-51 supply to the PSW system, fire damage to cables may prevent this valve Injection MOV from opening or cause the valve to spuriously close and challenge the Reactivity, Inventory and Pressure Control Nuclear Safety Performance I Criteria. This valve may suffer IN 92-18 damage.
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| BWST level indication is required to monitor the performance of the reactivity and inventory control systems. Fire induced cable damage 3LPIP 0345 - BWST Level TB-52 may result in loss of BWST level indication and challenge the Process Indication Monitoring Nuclear Safety Performance Criterion.
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| Note: The additional risk added because of these VFDRs, as determined from the FRE for this fire area, is provided in SE Table 3.5.
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| OFFICIAL USE ONLY SECURITY RELJ\TED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 283 Recovery Actions (RAs)
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| The licensee did not identify any recovery actions required for this fire area.
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| Fire Detection & Suppression Systems Required to Meet the Nuclear Safety Performance Criteria The licensee performed an evaluation of the fire detection and suppression systems in this area.
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| The results of the evaluation were documented in LAR Table 4-4 and the applicable portions have been included below. The identified fire detection system modifications are to improve plant fire detection and fire brigade response time. The existing detectors within the TB are required with improvements to the following fire zones for general area and/or fire hazard detection: 3, 6,12,15,19,24,25,28,29,32,33, 33A, 34, 34A, 35,37,39, 39A, 41.
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| Suppression Detection Required Auto Required System? System?
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| Fire Fire Detection Zone Description Suppression Area Zone Provided?
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| Provided?
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| E R D S E R D S TB Turbine Building Unit 3 Lube Oil No No No Yes No Yes No Yes No No 1 Purifier Area Unit 3 Electro-Hydraulic Control No No No No No Yes No Yes No No 2 (EHC) Area Unit 3 Heater Bay Yes No No No No No Yes No No No 3 Area (MR)
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| Unit 3 Turbine Driven EFDW No No No No No Yes No Yes No No 4 Pump Area Unit 3 Condensate No No No No No No No No No No 5 Booster Pump Area Unit 3 Main Yes Feedwater Pump Yes No Yes No No Yes No No No (MR) 6 Area Unit 3 Motor Driven No No No No No No No Yes No No 7 EFDW Pump Area Unit 3 Hotwell Pump & TB Sump No No No No No No No Yes No No 8 Area Unit 3 PowdexlLSPW No No No No No No No Yes No No 9 Pump Area Unit 2 Lube Oil No No No No No No No Yes No No 10 Purifier Area 11 Unit 2 EHC Area No No No No No No No Yes No No Unit 2 Heater Bay Yes No No No No No Yes No No No 12 Area (MR)
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| Unit 2 Turbine Driven EFDW No No No No No Yes No Yes No No 13 Pump Area Unit 2 Condensate No No No No No No No No No No 14 Booster Pump Area Unit 2 Main Yes Yes No Yes No No Yes No No No 15 Feedwater Pump (MR)
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| OFFICIAL USE ONLY SECURITY RELI\TED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 284 Suppression Detection Required Auto Required System? System?
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| Fire Fire Detection Zone Description Suppression Area Zone Provided?
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| Provided?
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| E R D S E R D S Area Unit 2 Motor Driven No No No No No No No Yes No No 16 EFDW Pump Area Unit 2 HW Pump, LPSW Pump - B No No No No No Yes No Yes No No 17 Area Unit 2 HPSW Pump No No No No No No No No No No 17A B Area Unit 2 Powdex, Backup IA No No No No No Yes No Yes No No 18 Compressors Unit 1 Main Yes Feedwater Pump Yes No Yes No No Yes No No No (MR) 19 Area Unit 1 Motor Driven EFDW Pump Area No No No No No Yes No Yes No No 20 R Unit 1 HW Pump, LPSW Pump-A No No No No No Yes No Yes No No 21 Area Unit 1 HPSW Pump No No No No No No No No No No 21A A Area 22 Unit 1 Powdex Area No No No No No Yes No Yes No No Unit 1 Lube Oil No No No No No No No No No No 22A StoraQe House Unit 1 Condensate No No No No No No No No No No 23 Booster Pump Area Unit 1 TDEFDW Yes Pump, EHC, Oil No No No No No Yes No No No (MR) 24 Purifier Unit 1 Feedwater Yes Heaters & Drain No No No No No Yes No No No (MR) 25 Pumps Unit 3 Moisture Separators (MS) B1 No No No No No No No No No No 26 & B2 Unit 3 Main Turbine (MT) Oil Tank and No No No No No Yes No Yes No No MS Stop & Control 27 Valves Unit 3 Heater Bay Area, Moisture Yes Separator No No No No No Yes No No No (MR)
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| Reheaters (MSRH) 28 A1 &A2 Unit 3 4160 Volt Yes No No No No No Yes No No No 29 SwitchQear (MR) 30 Unit 2 MSs B1 & B2 No No No No No No No No No No Unit 2 MT Oil Tank and MS Stop & No No No No No Yes No Yes No No 31 Control Valves Unit 2 Heater Bay Yes Area, MSRH A1 & No No No No No Yes No No No (MR) 32 A2 33 Unit 2 6900/4160 No No No No No Yes No Yes No No OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OrrlCIAL USE ONLY SECURITY RELATED INrORMATION 285 Suppression Detection Required Auto Required System? System?
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| Fire Fire Detection Zone Description Suppression Area Zone Provided?
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| Provided?
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| E R D S E R D S Volt SwitchQear (MR)
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| Unit 2 Power Yes No No No No No Yes No No No 33A Batteries (MR)
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| Unit 16900/4160 Yes No No No No No Yes No No No 34 Volt Switchoear (MR)
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| Unit 1 Power Yes No No No No No Yes No No No 34A Batteries (MR)
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| Unit 1 Heater Bay Yes Area, MSRH A 1 & No No No No No Yes No No No (MR) 35 A2 Unit 1 MT Oil Tank and MS Stop & No No No No No Yes No Yes No No 36 Control Valves Unit 2 2X11, Yes 2X11A, 3X5, 3X6 No No No No No Yes No No No (MR) 37 Area Unit 3 Main Turbine, Turbine No No No No No Yes No Yes No No 38 Fir, Offices Unit 3 Heater Bay & Yes No No No No No Yes No No No 39 Upper SurQe Tanks (MR)
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| Unit 3 Power Yes No No No No No Yes No No No 39A Batteries (MR)
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| Unit 2 Main Turbine, Turbine No No No No No Yes No Yes No No 40 Fir, Offices Unit 2 Heater Bay & Yes No No No No No Yes No No No 41 Upper Suroe Tanks (MR)
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| Unit 1 Main Turbine, Turbine No No No No No Yes No Yes No No 42 Fir, Offices Unit 1 Heater Bay &
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| No No No No No No No No No No 43 Upper Suroe Tanks Unit 1 TB Truck No No No No No No No No No No 44 Receivino Bav Legend:
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| E - EEEE/LA: Systems required for acceptability of EEE Evaluations 1 NRC approved Licensing Action (Section 2.2.7)
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| R - Risk: Systems required to meet the Risk Criteria for the PB Approach (Section 4.2.4)
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| D - Defense-in-Depth: Systems required to maintain adequate balance of Defense-in-Depth for a PB Approach (Section 4.2.4.2)
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| S - Separation Criteria: Systems required for Chapter 4 Separation Criteria in (Section 4.2.3)
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| MR-Modification Required Systems are committed to be modified as indicated in LAR Table 4-4 and Attachment S OrrlCIAL USE ONLY SECURITY RELATED INrORMATION
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| OFFICIAL USE ONLY SECURITY RElJ\TED INFORMATION 286 Fire Area TB Conclusion The licensee has utilized the FRE PB approach to demonstrate the ability to meet the NFPA 805 nuclear safety performance criteria for this fire area. A FRE in accordance with NFPA 805, Section 4.2.4.2, in conjunction with deterministic methods for simplifying assumptions, was used in applying this approach.
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| Based on the information provided in the LAR, as supplemented, the NRC staff finds Fire Area TB meets the nuclear safety goals, objectives, and performance criteria of NFPA 805. This conclusion is based on the following:
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| * Fire protection SSCs were evaluated in accordance with NFPA 805, Chapter 4 to determine which, if any, were required to meet the nuclear safety performance criteria. This evaluation included:
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| : a. The fire protection detection systems required to meet the nuclear safety performance criteria were documented.
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| : b. Fire Area boundaries were defined using three hour rated walls, ceilings and floors, including fire barriers, fire barrier penetrations, and through penetration fire stops and spatial separation.
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| * No exemptions or licensing actions from the pre-transition fire protection requirements were required.
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| * Forty-five VFDRs were identified, evaluated through the performance of a FRE, and either found to meet the risk acceptance criteria, as well as the requirements for DID and SMs, or modifications were planned/ implemented to address the issue. The acceptability of the risk for this fire area is contingent on the risk reduction from the planned PSW modification (see SE Section 3.4 for a detailed discussion of the NRC staff's review of the adequacy of the FRE method used at ONS).
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| * This fire area did not require the use of recovery actions to meet the nuclear safety performance criteria.
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| * The following modifications were identified to address VFDRs:
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| : a. Modifications for TB are required to the following zones for general area and/or fire hazard detection: 3, 6, 12, 15, 19,24,25,28,29,32,33, 33A, 34, 34A, 35, 37, 39, 39A, 41. These detection modifications are to improve plant fire detection and fire brigade response time.
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| : b. A modification (VFDR TB-06) to the wall separating the AB and the TB requires modification to upgrade to a three hour fire barrier. This wall is credited for area separation in the deterministic approach of NFPA 805, Section 4.2.3.
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| Fire Area WP1, Unit 1 West Penetration Room The licensee analyzed this fire area using the FRE approach in accordance with NFPA 805 Section 4.2.4.2, but also used deterministic simplifying assumptions in order to credit those OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 287 portions of the facility design that met the deterministic requirements of NFPA 805 Section 4.2.3. The licensee identified the SSCs necessary to meet the nuclear safety performance criteria in this fire area.
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| Evaluation of Fire Suppression Effects on Nuclear Safety Performance Criteria The licensee stated in Attachment C, "NEI 04-02, Table B-3, Fire Area Transition," that safe and stable conditions can be achieved and maintained using equipment and cables outside of the area of fire suppression activity. Flooding of the suppression areas and discharge of suppression water to adjacent compartments is controlled and will not jeopardize achievement of safe and stable conditions. Based on the information provided by the licensee in the NFPA 805 LAR, the NRC staff finds the licensee's evaluation of fire suppression effects on nuclear safety performance criteria acceptable because the results of the licensee's analysis indicate that fire suppression activities will not adversely affect achievement of the nuclear safety performance criteria.
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| Exemptions and Other Licensing Actions Based on the information provided in the LAR, the licensee credited two previously approved exemptions from the existing fire protection requirements. The licensee utilized the process described in LAR Section 4.2.3, "Licensing Action Transition," and Attachment K, "Licensing Action Review," to carry forward these exemptions, which requires a determination of the basis of acceptability and a determination that the basis of the acceptability is still valid. The NRC staff's evaluation of each exemption is provided in the table below.
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| Exemption I Licensee's Statement on Basis and Continuing Validity NRC Staff Evaluation Licensing Action Appendix R Provides the following justification for the lack of three hour Based on the previous Exemption, RB fire rated pipe penetrations: NRC staff approval of this Unrated Containment exemption and the Mechanical
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| * RB walls serve as a substantial heat sink. statement by the licensee Penetrations
| |
| * Combustible loading near penetrations is low. that the basis remains
| |
| * Mechanical pipe penetrations are designed to meet valid, the NRC staff finds multiple. this acceptable.
| |
| * Containment integrity criteria and are substantial.
| |
| * Large room volumes on both sides dissipate heat from a fire away from penetration area.
| |
| The bases for previous acceptance remain valid.
| |
| Appendix R Presented justification for the lack of three hour fire barriers Based on the previous Exemption, AB Lack of because: NRC staff approval of this three hour fire rated exemption and the barrier
| |
| * Low combustible loading in pipe tunnel access area. statement by the licensee
| |
| * Fire propagation path is circuitous, consisting of that the basis remains several unrated barriers and open areas. valid, the NRC staff finds
| |
| * If a fire were to occur, it would develop slowly. this acceptable.
| |
| * Fire brigade may use portable extinguishers, manual hose stations, or a fire hose supplied from a nearby fire hydrant.
| |
| In conclusion, although the exact number and configuration of combustibles may have changed over time, the bases for previous acceptance remain valid as substantiated by field walkdown.
| |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 288 Variation from Deterministic Requirements (VFDRs)
| |
| Fire Area WP1 has a total of 14 VFDRs, which are provided in the table below. All but two of these VFDRs are variances NFPA 805, Section 4.2.3 (separation issues) that were dispositioned with a FRE (SE Section 3.4.3). The licensee's FRE determined that these variances are acceptable based on 1) the change in CDF and LERF for the fire area and the total CDF and LERF for each unit meet the acceptance criteria of RG 1.174 (SE Section 3.4.6),
| |
| and 2) adequate DID and SMs are maintained for each fire area (SE Section 3.4.2). This determination relies on the following fire protection systems and features to meet the acceptance criteria:
| |
| * General area and/or hazard detection for the Unit 1 Purge Inlet Room (Fire Zone 120) is required to meet the risk acceptance criteria. The Fire PRA makes assumptions regarding the time of fire discovery, fire brigade notification, and brigade manual suppression. These assumptions determine the impact of the fire, including the likelihood of a HGL being formed in the compartment. Specifically, the Fire PRA is based on a fire brigade response time of 20 minutes or less. The existing room detection system coverage of the general area and/or hazard necessary for this assumption to be valid was not considered sufficient to conservatively meet the risk criteria. Therefore, modification to the fire detection system in the Purge Inlet Room is required to support the fire risk analysis assumption of 20 minute brigade response time.
| |
| * General area and/or hazard detection associated with the Unit 1 West Penetration Pen Room (Fire Zone 107) and the Unit 1 Cask Decon Tank Room (Fire Zone 97) are required to meet the DID criteria.
| |
| Based on the reliance on fire detectors in the Unit 1 West Penetration Room (Fire Area WP1) to meet the risk and DID criteria, the licensee has committed to make modifications to the fire detection system, which may include fire detector upgrades and/or new installation.
| |
| Improvements of general area and/or fire hazard detection are required for the Unit 1 West Penetration Pen Room (Fire Zone 107), the Unit 1 Cask Decon Tank Room (Fire Zone 97), and the Unit 1 Purge Inlet Room (Fire Zone 120) (SE Section 2.8.1).
| |
| Two of the 14 VFDRs, WP1-04 and WP1-10, are variances from NFPA 805 Section 4.2.3 (separation issue) that will be corrected with a plant modification. According to the LAR, the wall separating the Unit 1 Purge Inlet Room from the SFP Area and the wall separating the AB (Fire Area AB) from the Unit 1 West Penetration Room (Fire Area WP1) are not currently three hour fire rated as required by NFPA 805, Section 3.11.1, and all of the penetrations in the walls do not have a fire resistance rating as required by NFPA 805, Section 3.11.3. These walls are credited for area separation using the deterministic approach of NFPA 805, Section 4.2.3. The licensee has committed to make modifications to the wall to bring it into compliance with the requirements of NFPA 805, Section 2.8.1.
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| VFDR# VFDR Description I Component (Cables) I OFFICI/\L USE ONLY SECURITY RELATED INFORM,A,TION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 289 VFDR# VFDR Description Component (Cables)
| |
| RB pressure instrumentation is required for process monitoring and diagnosis of plant transients. Fire induced cable damage may result in a loss of RB 1BS P 0011 - RB WP1-01 pressure indication resulting in the inability of the operator to monitor and Pressure Indication control this parameter from the MCR and challenge the Process Monitoring Nuclear Safety Performance Criterion.
| |
| This normally closed, required closed valve provides train separation by isolating the cross connect header between the 1A and 1B trains of Emergency Feedwater. Fire-induced cable damage may result in the spurious opening of 1CCWVA0269 - SG A WP1-02 this valve, a diversion of flow to either the 1A or 1B SGs, and a challenge to the FOW Control MaV OHR Nuclear Safety Performance Criterion. This valve may suffer IN 92-18 damage.
| |
| This normally open, required open valve is located in the EFW flow path to the 1B SG. Fire induced cable damage may result in spurious closing of this valve, isolating Protected Service Water flow to the 1B SG. The subsequent decrease 1FDWVA0347 - SG B WP1-03 in SG shell temperature may result in 1B SG exceeding its tube to shell Inlet MaV differential temperature limit. This could challenge the Inventory Control and OHR Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 damaQe.
| |
| The areas separating the Unit 1 Purge Inlet Room and SFP area is not three hour rated as required by NFPA 805, Section 3.11.1 and the penetrations Purge Inlet Room I WP1-04 (seals and doors) do not have a fire resistance rating as required by NFPA 805, SFP Area Section 3.11.3. These barriers are credited for area separation in the deterministic approach of NFPA 805, Section 4.2.3.
| |
| Normally open valve 1HP VA0023 is in the flow path from the LOST to the suction of the credited HPI pump. Normally closed valve 1HP VA0939 isolates the flow path from the LOST to the containment sump. Recirculation flow to the LOST during prolonged operation of the HPI pump at low flow conditions may 1HP VA0023 - HPI result in an increase in temperature of LOST contents to the operability limit of Normal Suction MaV, the HPI pump. The contents of the LOST must be diverted to the containment WP1-05 1HP VA0939 - LOST sump by opening 1HP VA0939 and closing 1HP VA0023 prior to the operability to Emergency Sump limit of the HPI pump being exceeded to prevent challenging the Reactivity, MaV Inventory and Pressure Control Nuclear Safety Performance Criteria. Although unaffected by fire, the power supplies for these valves are not credited following a fire in this fire area and a loss of power may prevent these valves from being repositioned.
| |
| These normally open, required closed valves isolate the flow path from the BWST to the LPI Pumps, RBS Pumps, and containment sump. Although unaffected by fire, the power supplies to these valves are not credited following a fire in this fire area and a loss of power may prevent these valves from being 1LP VA0021, 1LP WP1-07 closed and result in a diversion of BWST inventory to the containment sump via VA0022 - BWST the LPI system. In addition, an inadvertent ES actuation could result in a Suction MaVs diversion of BWST inventory to the containment sump via the RBS system. A loss of BWST inventory could challenge the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria.
| |
| BWST level indication is required to monitor the performance of the reactivity and inventory control systems. Fire-induced cable damage may result in loss 1LPIP 0345 - BWST WP1-08 of BWST level indication and challenge the Process Monitoring Nuclear Safety Level Indication Performance criterion.
| |
| 1MS VA0017, 1MS VA0024,1MS VA0026,1MS These normally open, required closed valves isolate flow paths from the MSHs.
| |
| VA0033,1MS Although unaffected by fire, the power supplies to these valves are not credited VAOO35,1MS following a fire in this fire area and a loss of power may prevent these valves WP1-09 VAOO36,1MS from being closed and could result in overcooling and shrinkage of RC VAOO76,1MS inventory. This could challenge the OHR Nuclear Safety Performance VAOO79,1MS Criterion.
| |
| VAOO82,1MS VA0084 - SG Isolation MaVs WP1-10 The wall separating the AB and the West penetration room does not have a AB I West OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 290 VFDR# VFDR Description Component (Cables) fire-resistance rating required by NFPA 805, Section 3.11.2 and all the Penetration Room penetrations in the wall do not have a fire resistance rating as required by Separation NFPA 805, Section 3.11.3. This wall is credited for area separation in the deterministic approach of NFPA 805, Section 4.2.3.
| |
| 1RC SXTRN001 ,
| |
| Pressurizer heaters are required for RC pressure control. The heaters receive 1RC SXTRN002 non-credited power from Unit 1 and credited power from a PSW system power Pressurizer Heaters WP1-11 supply. The transfer of credited power to the pressurizer heaters requires a PSW Power Transfer recovery action. Failure to transfer credited power to the heaters could Switches challenge the Pressure Control Nuclear Safety Performance Criterion.
| |
| These normally closed, required closed valves isolate flow paths from the RCS to containment. A series of potential hot shorts within the terminal box of the 1RCVA0157,1RC electrical penetration may spuriously open the reactor head vent and hot leg VA0159 - RC Hot Leg WP1-12 vent valves. The spurious opening of these valves may result in a loss of RC and Head Vent inventory and challenge the Inventory and Pressure Control Nuclear Safety Valves Performance Criteria.
| |
| These normally closed valves isolate the flow path from the RCS to containment. These valves are required to open to provide an RC letdown flow 1RC VA0159, 1RC WP1-13 path. Fire induced cable damage may prevent these valves from being opened VA0160 - RC Head resulting in the lifting of the pressurizer safety relief valves and a challenge to Vent Valves the Inventory Control Nuclear Safety Performance Criterion.
| |
| Although unaffected by fire, the power supplies for the station HVAC system are not credited following a fire in this fire area and a loss of power may result Units 1 & 2 Control WP1-15 in the temperature inside the Units 1 & 2 control complex exceeding the Complex Cooling operability limit of SSD components and challenge the Vital Auxiliaries Nuclear Safety Performance Criterion.
| |
| Fire damage to cables may result in a loss of power to the containment cooling system and may result in the temperature inside the Unit 1 RB exceeding the Unit 1 Containment WP1-16 operability limit of SSD components. This could challenge the Vital Auxiliaries Cooling Nuclear Safety Performance Criterion.
| |
| Note: The additional risk added because of these VFDRs, as determined from the FRE for this fire area, is provided in SE Table 3.5.
| |
| Recovery Actions (RAs)
| |
| The licensee did not identify any recovery actions required for this fire area.
| |
| Fire Detection & Suppression Systems Required to Meet the Nuclear Safety Performance Criteria The licensee performed an evaluation of the fire detection and suppression systems in this area.
| |
| The results of the evaluation were documented in LAR Table 4-4 and the applicable portions have been included below. These identified fire detection system modifications are to improve plant fire detection and fire brigade response time.
| |
| Suppression Detection Required Required Auto System?
| |
| Fire Fire System? Detection Zone Description Suppression Area Zone Required?
| |
| Required?
| |
| E R D S E R D S Unit 1 West Penetration WP1 Room Unit 1 Cask Decon Tank WP1 97 No No No No No Yes Yes No No No Room Unit 1 West Penetration Yes WP1 107 No No No No No Yes No No No Pen Room (MR)
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 291 WP1 I 120 I Unit 1 Purge Inlet Room I No I No I No I No I No I Yes Legend:
| |
| E - EEEE/LA: Systems required for acceptability of EEE Evaluations I NRC approved Licensing Action (Section 2.2.7)
| |
| R - Risk: Systems required to meet the Risk Criteria for the PB Approach (Section 4.2.4)
| |
| D - Defense-in-Depth: Systems required to maintain adequate balance of Defense-in-Depth for a PB Approach (Section 4.2.4.2)
| |
| S - Separation Criteria: Systems required for Chapter 4 Separation Criteria in (Section 4.2.3)
| |
| MR - Modification Required Systems are committed to be modified as indicated in LAR Table 4-4 and Attachment S Fire Area WP1 Conclusion The licensee has utilized the FRE PB approach to demonstrate the ability to meet the NFPA 805 nuclear safety performance criteria for this fire area. A FRE in accordance with NFPA 805, Section 4.2.4.2, in conjunction with deterministic methods for simplifying assumptions, was used in applying this approach.
| |
| Based on the information provided in the LAR, as supplemented, the NRC staff finds Fire Area WP1 meets the nuclear safety goals, objectives, and performance criteria of NFPA 805. This conclusion is based on the following:
| |
| * Fire protection SSCs were evaluated in accordance with NFPA 805, Chapter 4 to determine which, if any, were required to meet the nuclear safety performance criteria. This evaluation included:
| |
| : a. The fire protection detection systems required to meet the nuclear safety performance criteria were documented.
| |
| : b. Fire Area boundaries were defined using three hour rated walls, ceilings and floors, including fire barriers, fire barrier penetrations, and through penetration fire stops and spatial separation.
| |
| * Two exemptions from the pre-transition fire protection requirements were evaluated and found to be valid and applicable under the NFPA 805 RIJPB FPP.
| |
| * Fourteen VFDRs were identified, evaluated through the performance of a FRE, and either found to meet the risk acceptance criteria, as well as the requirements for DID and SMs, or modifications were planned/ implemented to address the issue. The acceptability of the risk for this fire area is contingent on the risk reduction from the planned PSW modification (see SE Section 3.4 for a detailed discussion of the NRC staff's review of the adequacy of the FRE method used at ONS).
| |
| * This fire area did not require the use of recovery actions to meet the nuclear safety performance criteria.
| |
| * The following modifications were identified to address VFDRs:
| |
| : a. Improve general area and/or hazard detection for Unit 1 Purge Inlet Room, Unit 1 West Penetration Pen Room and Unit 1 Cask Decon Tank Room.
| |
| o In order to take credit for evaluations in the fire area, the following barrier modifications are required:
| |
| o AB / Unit 1 West Penetration Room separation o Unit 1 Purge Inlet Room / SFP Area separation OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 292 Fire Area WP2, Unit 2 West Penetration Room The licensee analyzed this fire area using the FRE approach in accordance with NFPA 805 Section 4.2.4.2, but also used deterministic simplifying assumptions in order to credit those portions of the facility design that met the deterministic requirements of NFPA 805 Section 4.2.3. The licensee identified the SSCs necessary to meet the nuclear safety performance criteria in this fire area.
| |
| Evaluation of Fire Suppression Effects on Nuclear Safety Performance Criteria The licensee stated in Attachment C, "NEI 04-02, Table B-3, Fire Area Transition," that safe and stable conditions can be achieved and maintained using equipment and cables outside of the area of fire suppression activity. Flooding of the suppression areas and discharge of suppression water to adjacent compartments is controlled andwill not jeopardize achievement of safe and stable conditions.
| |
| Based on the information provided by the licensee in the NFPA 805 LAR, the NRC staff finds the licensee's evaluation of fire suppression effects on nuclear safety performance criteria acceptable because the results of the licensee's analysis indicate that fire suppression activities will not adversely affect achievement of the nuclear safety performance criteria.
| |
| Exemptions and Other Licensing Actions Based on the information provided in the LAR, the licensee credited two previously approved exemptions from the existing fire protection requirements. The licensee utilized the process described in LAR Section 4.2.3, "Licensing Action Transition," and Attachment K, "Licensing Action Review," to carry forward these exemptions, which requires a determination of the basis of acceptability and a determination that the basis of the acceptability is still valid. The NRC staff's evaluation of each exemption is provided in the table below.
| |
| Exemption I Basis and Continuing Validity NRC Staff Evaluation Licensinq Action Provides the following justification for the lack of three hour Based on the previous fire rated pipe penetrations: NRC staff approval of this exemption and the
| |
| * RB walls serve as a substantial heat sink. statement by the licensee Appendix R Exemption, that the basis remains RB Unrated
| |
| * Combustible loading near penetrations is low.
| |
| * Mechanical pipe penetrations are designed to meet valid, the NRC staff finds Containment this acceptable.
| |
| multiple containment integrity criteria and are Mechanical substantial.
| |
| Penetrations
| |
| * Large room volumes on both sides dissipate heat from a fire away from penetration area.
| |
| The bases for previous acceptance remain valid.
| |
| Presented justification for the lack of three hour fire barriers Based on the previous because: NRC staff approval of this exemption and the Appendix R Exemption,
| |
| * Low combustible loading in pipe tunnel access statement by the licensee AB Lack of three hour area. that the basis remains fire rated barrier
| |
| * Fire propagation path is circuitous, consisting of valid, the NRC staff finds several unrated barriers and open areas. this acceptable.
| |
| * If a fire were to occur, it would develop slowly.
| |
| * Fire briqade may use portable extinquishers, OFFICIAL USE ONLY SECURITY RELATED INFORMATION
| |
| | |
| OFFICIAL USE ONLY SECURITY REU\TED INFORMATION 293 manual hose stations, or a fire hose supplied from a nearby fire hydrant.
| |
| In conclusion, although the exact number and configuration of combustibles may have changed over time, the bases for previous acceptance remain valid as substantiated by field walkdown.
| |
| Variation from Deterministic Requirements (VFDRs)
| |
| Fire Area WP2 has a total of 14 VFDRs, which are provided in the table below. All but two of these VFDRs are variances from NFPA 805 Section 4.2.3 (separation issues) that were dispositioned with an FRE (SE Section 3.4.3). The licensee's FRE determined that these variances are acceptable based on 1) the change in CDF and LERF for the fire area and the total CDF and LERF for each unit meet the acceptance criteria of RG 1.174 (SE Section 3.4.6) and 2) adequate DID and SMs are maintained for each fire area (SE Section 3.4.2). This determination relies on the following fire protection systems and features to meet the acceptance criteria:
| |
| * General area and/or hazard detection for the Unit 2 Purge Inlet Room (Fire Zone 117) is required to meet the risk acceptance criteria. The Fire PRA makes assumptions regarding the time of fire discovery, fire brigade notification, and brigade manual suppression. These assumptions determine the impact of the fire, including the likelihood of an HGL being formed in the compartment. Specifically, the Fire PRA is based on a fire brigade response time of 20 minutes or less. The existing room detection system coverage of the general area and/or hazard necessary for this assumption to be valid was not considered sufficient to conservatively meet the risk criteria. Therefore, modification to the fire detection system in the Purge Inlet Room is required to support the fire risk analysis assumption of a 20-minute brigade response time.
| |
| * General area and/or hazard detection associated with the Unit 2 West Penetration Pen Room (Fire Zone 102) and the Unit 2 Cask Decon Tank Room (Fire Zone 91) are required to meet the DID criteria.
| |
| Based on the reliance on fire detectors in the West Penetration Room WP2 Fire Area to meet the risk and DID criteria, the licensee has committed to make modifications to the fire detection system, which may include fire detector upgrades and/or new installation. Improvements of general area and/or fire hazard detection are required for the Unit 2 West Penetration Pen Room (Fire Zone 102), the Unit 2 Purge Inlet Room (Fire Zone 117), and the Unit 2 Cask Decon Tank Room (, 2.8.1).
| |
| Two of the 14 VFDRs, WP2-04 and WP2-1 0, are variances from NFPA 805 Section 4.2.3, (separation issue) that will be corrected with a plant modification. According to the LAR, the wall separating the Unit 2 Purge Inlet Room from the SFP Area and the wall separating the AB (Fire Area AB) from the Unit 2 West Penetration Room (Fire Area WP2) are not currently a three-hour rated wall as required by NFPA 805, Section 3.11.1, and all of the penetrations in the walls do not have a fire resistance rating as required by NFPA 805, Section 3.11.3. These walls are credited for area separation using the deterministic approach of NFPA 805, Section 4.2.3.
| |
| The licensee has committed to make modifications to the wall to bring it into compliance with the requirements of NFPA 805, Section 2.8.1.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL UgE ONLY gECURITY RELATED INFORMATION 294 Component VFDR# VFDR Description (Cables)
| |
| This normally closed, required closed valve provides train separation by isolating the cross connect header between the 2A and 2B trains of Emergency 2CCWVA0269 - SG Feedwater. Fire-induced cable damage may result in the spurious opening of WP2-02 A FOW Control this valve, a diversion of flow to either the 2A or 2B SGs, and a challenge to the MOV OHR Nuclear Safety Performance Criterion. This valve may suffer IN 92-18 damaQe.
| |
| This normally open, required open valve is located in the EFW flow path to the 2B SG. Fire-induced cable damage may result in spurious closing of this valve, isolating Protected Service Water flow to the 2B SG. The subsequent decrease 2FDWVA0347 - SG WP2-03 in SG shell temperature may result in 2B SG exceeding its tube to shell B Inlet MOV differential temperature limit. This could challenge the Inventory Control and OHR Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 damaqe.
| |
| The areas separating the Unit 2 Purge Inlet Room and SFP area is not three-hour rated as required by NFPA 80S, Section 3.11.1 and the penetrations (seals Purge Inlet Room I WP2-04 and doors) do not have a fire resistance rating as required by NFPA 805, SFP Area Section 3.11.3. These barriers are credited for area separation in the deterministic approach of NFPA 805, Section 4.2.3.
| |
| Normally open valve 2HP VA0023 is in the flow path from the LOST to the suction of the credited HPI pump. Normally closed valve 2HP VA0939 isolates the flow path from the LOST to the containment sump. Recirculation flow to the LOST during prolonged operation of the HPI pump at low flow conditions may 2HP VA0023 - HPI result in an increase in temperature of LOST contents to the operability limit of Normal Suction the HPI pump. The contents of the LOST must be diverted to the containment MOV, 2HP VA0939 WP2-05 sump by opening 2HP VA0939 and closing 2HP VA0023 prior to the operability - LOST to limit of the HPI pump being exceeded to prevent challenging the Reactivity, Emergency Sump Inventory and Pressure Control Nuclear Safety Performance Criteria. Although MOV unaffected by fire, the power supplies for these valves are not credited following a fire in this fire area and a loss of power may prevent these valves from being repositioned.
| |
| These normally open, required closed valves isolate the flow path from the BWST to the LPI Pumps, RBS Pumps, and containment sump. Although unaffected by fire, the power supplies to these valves are not credited following a fire in this fire area and a loss of power may prevent these valves from being 2LP VA0021, 2LP WP2-07 closed and result in a diversion of BWST inventory to the containment sump via VA0022 - BWST the LPI system. In addition, an inadvertent ES actuation could result in a Suction MOVs diversion of BWST inventory to the containment sump via the RBS system. A loss of BWST inventory could challenge the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria.
| |
| BWST level indication is required to monitor the performance of the reactivity and inventory control systems. Fire-induced cable damage may result in loss of 2LPIP 0345 - BWST WP2-08 BWST level indication and challenge the Process Monitoring Nuclear Safety Level Indication Performance criterion.
| |
| 2MS VA0017, 2MS VA0024,2MS VA0026,2MS These normally open, required closed valves isolate flow paths from the MSHs . VA0033,2MS Although unaffected by fire, the power supplies to these valves are not credited VA0035,2MS WP2-09 following a fire in this fire area and a loss of power may prevent these valves VA0036,2MS from being closed and could result in overcooling and shrinkage of RC inventory. VA0076,2MS This could challenge the OHR Nuclear Safety Performance Criterion. VA0079,2MS VA0082,2MS VA0084 - SG Isolation MOVs The wall separating the AB and the West penetration room does not have a fire-resistance rating required by NFPA 805, Section 3.11.2 and all the penetrations AS I West WP2-10 in the wall do not have a fire resistance rating as required by NFPA 80S, Section Penetration Room 3.11.3. This wall is credited for area separation in the deterministic approach of Separation NFPA 805, Section 4.2.3.
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| OFFICIAL UgE ONLY gECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 295 Component VFDR# VFDR Description (Cables) 2RC SXTRN001 ,
| |
| Pressurizer heaters are required for RC pressure control. The heaters receive 2RC SXTRNOO2, non-credited power from Unit 1 and credited power from a PSW system power 2RC SXTRN003 WP2-11 supply. The transfer of credited power to the pressurizer heaters requires a Pressurizer Heaters recovery action. Failure to transfer credited power to the heaters could PSW Power challenge the Pressure Control Nuclear Safety Performance Criterion.
| |
| Transfer Switches This normally closed, required closed valve isolates the flow path from the RCS 2RC VAOO66 to the Quench Tank. Fire-induced cable damage may result in the spurious Pressurizer Power WP2-13 opening of the PORV causing a loss of RC inventory and RC subcooling. This Operated Relief could challenge the Inventory and Pressure Control Nuclear Safety Performance Valve Criteria.
| |
| These normally closed valves isolate the flow path from the RCS to containment.
| |
| These valves are required to open to provide an RC letdown flow path. Fire- 2RC VA0159, 2RC WP2-14 induced cable damage may prevent these valves from being opened resulting in VA0160 - RC Head the lifting of the pressurizer safety relief valves and a challenge to the Inventory Vent Valves Control Nuclear Safety Performance Criterion.
| |
| This normally closed, required closed valve isolates the flow path from the RCS to containment. A series of potential hot shorts within the terminal box of the 2RC VA0159 - RC WP2-15 electrical penetration may spuriously open the reactor head vent valve. The Head Vent Valve spurious opening of this valve may result in a loss of RC inventory and challenge the Inventory and Pressure Control Nuclear Safety Performance Criteria.
| |
| Although unaffected by fire, the power supplies for the station HVAC system are not credited following a fire in this fire area and a loss of power may result in the Units 1 & 2 Control WP2-17 temperature inside the Units 1 & 2 control complex exceeding the operability Complex Cooling limit of SSD components and challenge the Vital Auxiliaries Nuclear Safety Performance Criterion.
| |
| Fire damage to cables may result in a loss of power to the containment cooling system and may result in the temperature inside the Unit 2 RB exceeding the Unit 2 Containment WP2-18 operability limit of SSD components. This could challenge the Vital Auxiliaries Cooling Nuclear Safety Performance Criterion.
| |
| Note: The additional risk added because of these VFDRs, as determined from the FRE for this fire area, is provided in SE Table 3.5.
| |
| Recovery Actions (RAs)
| |
| The licensee did not identify any recovery actions required for this fire area.
| |
| Fire Detection & Suppression Systems Required to Meet the Nuclear Safety Performance Criteria The licensee performed an evaluation of the fire detection and suppression systems in this area.
| |
| The results of the evaluation were documented in LAR Table 4-4 and the applicable portions have been included below. These identified fire detection system modifications are to improve plant fire detection and fire brigade response time.
| |
| Suppression Detection Required Required Auto System?
| |
| Fire Fire System? Detection Zone Description Suppression Area Zone Required?
| |
| Provided?
| |
| E R D S E R D S Unit 2 West Penetration WP2 Room Unit 2 Cask Decon Tank WP2 91 No No No No No Yes Yes No No No Room OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 296 Unit 2 West Penetration Yes WP2 102 No No No No No Yes No No No Pen Room (MR)
| |
| Yes WP2 117 Unit 2 Purge Inlet Room No No No No No Yes No No No (MR)
| |
| Legend:
| |
| E - EEEE/lA Systems required for acceptability of EEE Evaluations I NRC approved Licensing Action (Section 2.2.7)
| |
| R - Risk: Systems required to meet the Risk Criteria for the PB Approach (Section 4.2.4)
| |
| D - Defense-in-Depth: Systems required to maintain adequate balance of Defense-in-Depth for a PB Approach (Section 4.2.4.2)
| |
| S - Separation Criteria: Systems required for Chapter 4 Separation Criteria in (Section 4.2.3)
| |
| MR- Modification Required Systems are committed to be modified as indicated in LAR Table 4-4 and Attachment S Fire Area WP2 Conclusion The licensee has utilized the FRE PB approach to demonstrate the ability to meet the NFPA 805 nuclear safety performance criteria for this fire area. An FRE in accordance with NFPA 805, Section 4.2.4.2, in conjunction with deterministic methods for simplifying assumptions, was used in applying this approach.
| |
| Based on the information provided in the LAR, as supplemented, the NRC staff finds Fire Area WP2 meets the nuclear safety goals, objectives, and performance criteria of NFPA 805. This conclusion is based on the following:
| |
| * Fire protection SSCs were evaluated in accordance with NFPA 805 Chapter 4 to determine which, if any, were required to meet the nuclear safety performance criteria. This evaluation included:
| |
| : a. The fire protection detection systems required to meet the nuclear safety performance criteria were documented.
| |
| : b. Fire Area boundaries were defined using three-hour rated walls, ceilings and floors, including fire barriers, fire barrier penetrations, and through penetration fire stops and spatial separation.
| |
| * Two exemptions from the pre-transition fire protection requirements were evaluated and found to be valid and applicable under the NFPA 805 RI/PB FPP.
| |
| * Fourteen VFDRs were identified, evaluated through the performance of an FRE, and either found to meet the risk acceptance criteria, as well as the requirements for DID and SMs, or modifications were planned/ implemented to address the issue. The acceptability of the risk for this fire area is contingent on the risk reduction from the planned PSW modification (see SE Section 3.4 for a detailed discussion of the NRC staff's review of the adequacy of the FRE method used).
| |
| * This fire area did not require the use of recovery actions to meet the nuclear safety performance criteria.
| |
| * The following modifications were identified to address VFDRs:
| |
| Improve general area and/or hazard fire detection for Unit 2 Purge Inlet Room, Unit 2 West Penetration Pen Room, and Unit 2 Cask Tank Decon Room.
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| : a. In order to take credit for evaluations in the fire area, the following barrier modifications are required:
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 297 AB / Unit 2 West Penetration Room separation
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| * Unit 2 Purge Inlet Room / SFP Area separation Fire Area WP3, Unit 3 West Penetration Room The licensee analyzed this fire area using the FRE approach in accordance with NFPA 805, Section 4.2.4.2, but also used deterministic simplifying assumptions in order to credit those portions of the facility design that met the deterministic requirements of NFPA 805, Section 4.2.3. The licensee identified the SSCs necessary to meet the nuclear safety performance criteria in this fire area.
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| Evaluation of Fire Suppression Effects on Nuclear Safety Performance Criteria The licensee stated in Attachment C, "NEI 04-02, Table B-3, Fire Area Transition," that safe and stable conditions can be achieved and maintained using equipment and cables outside of the area of fire suppression activity. Flooding of the suppression areas and discharge of suppression water to adjacent compartments is controlled and will not jeopardize achievement of safe and stable conditions. Based on the information provided by the licensee in the NFPA 805 LAR, the NRC staff finds the licensee's evaluation of fire suppression effects on nuclear safety performance criteria acceptable because the results of the licensee's analysis indicate that fire suppression activities will not adversely affect achievement of the nuclear safety performance criteria.
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| Exemptions and Other Licensing Actions Based on the information provided in the LAR, the licensee credited two previously approved exemptions from the existing fire protection requirements. The licensee utilized the process described in LAR Section 4.2.3, "Licensing Action Transition," and Attachment K, "Licensing Action Review," to carry forward these exemptions, which requires a determination of the basis of acceptability and a determination that the basis of the acceptability is still valid. The NRC staff's evaluation of each exemption is provided in the table below.
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| Exemption I Licensee's Statement on Basis and Continuing Validity NRC Staff Evaluation Licensing Action Provides the following justification for the lack of three-hour Based on the previous fire-rated pipe penetrations: NRC staff approval of this exemption and the
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| * RB walls serve as a substantial heat sink. statement by the licensee Appendix R that the basis remains Exemption, RB
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| * Combustible loading near penetrations is low.
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| * Mechanical pipe penetrations are designed to meet valid, the NRC staff finds Unrated Containment this acceptable.
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| multiple containment integrity criteria and are Mechanical substantial.
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| Penetrations
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| * Large room volumes on both sides dissipate heat from a fire away from penetration area.
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| The bases for previous acceptance remain valid.
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| Presented justification for the lack of three-hour fire barriers Based on the previous Appendix R because: NRC staff approval of this Exemption, AB Lack of exemption and the three hour fire rated barrier
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| * Low combustible loading in pipe tunnel access area. statement by the licensee
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| * Fire propagation path is circuitous, consisting of that the basis remains OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 298 several unrated barriers and open areas. valid, the NRC staff finds
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| * If a fire were to occur, it would develop slowly. this acceptable.
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| * Fire brigade may use portable extinguishers, manual hose stations, or a fire hose supplied from a nearby fire hydrant.
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| In conclusion, although the exact number and configuration of combustibles may have changed over time, the bases for previous acceptance remain valid as substantiated by field walkdown.
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| Variation from Deterministic Requirements (VFDRs)
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| Fire Area WP3 has a total of 12 VFDRs, which are provided in the table below. All but one of these VFDRs are variances from NFPA 805 Section 4.2.3 (separation issues) that were dispositioned with a FRE (SE Section 3.4.3). The licensee's FRE determined that these variances are acceptable based on 1) the change in CDF and LERF for the fire area and the total CDF and LERF for each unit meet the acceptance criteria of RG 1.174 (SE Section 3.4.6) and 2) adequate DID and SMs are maintained for each fire area (SE Section 3.4.2). This determination relies on the following fire protection systems and features to meet the acceptance criteria:
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| * General area and/or hazard detection for the Unit 3 Purge Inlet Room (Fire Zone 114) is required to meet the risk acceptance criteria. The Fire PRA makes assumptions regarding the time of fire discovery, fire brigade notification, and brigade manual suppression. These assumptions determine the impact of the fire, including the likelihood of a HGL being formed in the compartment. Specifically, the Fire PRA is based on a fire brigade response time of 20 minutes or less. The existing room detection system coverage of the general area and/or hazard necessary for this assumption to be valid was not considered sufficient to conservatively meet the risk criteria. Therefore, modification to the fire detection system in the Purge Inlet Room is required to support the fire risk analysis assumption of 20 minute brigade response time.
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| * General area and/or hazard detection associated with the Unit 3 West Penetration Pen Room (Fire Zone 98) and the Unit 3 Cask Decon Tank Room (Fire Zone 87) are required to meet the DID criteria.
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| Based on the reliance on fire detectors in the West Penetration Room WP3 Fire Area to meet the risk and DID criteria, the licensee has committed to make modifications to the fire detection system, which may include fire detector upgrades and/or new installation. Improvements of general area and/or fire hazard detection are required for the Unit 3 West Penetration Pen Room (Fire Zone 98), the Unit 3 Purge Inlet Room (Fire Zone 114), and the Unit 3 Cask Decon Tank Room (Fire Zone 87) (SE Section 2.8.1).
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| One of the 12 VFDRs, WP3-03 is a variance from NFPA 805, Section 4.2.3, (separation issue) that will be corrected with a plant modification. According to the LAR, the wall separating the Unit 3 Purge Inlet Room from the SFP Area is not currently a three hour rated wall as required by NFPA 805, Section 3.11.1, and all of the penetrations in the walls do not have a fire resistance rating as required by NFPA 805, Section 3.11.3. These walls are credited for area separation using the deterministic approach of NFPA 805, Section 4.2.3. The licensee has committed to make modifications to the wall to bring it into compliance with the requirements of NFPA 805, Section 2.8.1.
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| OFFICIAL USE ONLY SECURITY RELATED It>JFORMATION
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| OffiCIAL USE ONLY SECURITY RELATED INfORMATION 299 Component VFDR# VFDR Description (Cables)
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| This normally closed, required closed valve provides train separation by isolating the cross connect header between the 3A and 38 trains of EFW. Fire induced 3CCWVA0269 - SG WP3-01 cable damage may result in spurious opening of this valve, a diversion of flow to A FOW Control either the 3A or 38 SGs, and a challenge to the OHR Nuclear Safety Performance MOV Criterion. This valve may suffer IN 92-18 damage.
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| This normally open, required open valve is located in the EFW flow path to the 38 SG. Fire induced cable damage may result in spurious closing of this valve, isolating Protected Service Water flow to the 38 SG. The subsequent decrease in 3FDWVA0347 - SG WP3-02 SG shell temperature may result in 38 SG exceeding its tube to shell differential 8 Inlet MOV temperature limit. This could challenge the Inventory Control and OHR Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 damaqe.
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| The areas separating the Unit 3 Purge Inlet Room and SFP area is not three hour rated as required by NFPA 805, Section 3.11.1 and the penetrations (seals and Purge Inlet Room I WP3-03 doors) do not have a fire resistance rating as required by NFPA 805, Section SFP Area 3.11.3. These barriers are credited for area separation in the deterministic approach of NFPA 805, Section 4.2.3.
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| Normally open valve 3HP VA0023 is in the flow path from the LOST to the suction of the credited HPI pump. Normally closed valve 3HP VA0939 isolates the flow path from the LOST to the containment sump. Recirculation flow to the LOST 3HP VA0023 - HPI during prolonged operation of the HPI pump at low flow conditions may result in Normal Suction an increase in temperature of LOST contents to the operability limit of the HPI MOV, 3HP VA0939 WP3-04 pump. The contents of the LOST must be diverted to the containment sump by
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| - LOST to opening 3HP VA0939 and closing 3HP VA0023 prior to the operability limit of the Emergency Sump HPI pump being exceeded to prevent challenging the Reactivity, Inventory and MOV Pressure Control Nuclear Safety Performance Criteria. Although unaffected by fire, the power supplies for these valves are not credited following a fire in this fire area and a loss of power may prevent these valves from beinq repositioned.
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| These normally open, required closed valves isolate the flow path from the 8WST to the LPI Pumps, R8S Pumps, and containment sump. Although unaffected by fire, the power supplies to these valves are not credited following a fire in this fire area and a loss of power may prevent these valves from being closed and result 3LP VA0021, 3LP WP3-06 in a diversion of 8WST inventory to the containment sump via the LPI system. In VA0022 - 8WST addition, an inadvertent ES actuation could result in a diversion of 8WST Suction MOVs inventory to the containment sump via the R8S system. A loss of 8WST inventory could challenge the Reactivity, Inventory and Pressure Control Nuclear Safety Performance Criteria.
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| 3MS VA0017, 3MS VA0024, 3MS VA0026.3MS These normally open, required closed valves isolate flow paths from the MSHs . VA0033,3MS Although unaffected by fire, the power supplies to these valves are not credited VA0035, 3MS WP3-07 following a fire in this fire area and a loss of power may prevent these valves from VA0036, 3MS being closed and could result in overcooling and shrinkage of RC inventory. This VA0076, 3MS could challenge the OHR Nuclear Safety Performance Criterion. VA0079, 3MS VA0082,3MS VA0084 - SG Isolation MOVs 3RC SXTRN001, Pressurizer heaters are required for RC pressure control. The heaters receive 3RC SXTRN002, non-credited power from Unit 3 and credited power from a PSW system power 3RC SXTRN003 WP3-09 supply. The transfer of credited power to the pressurizer heaters requires a Pressurizer Heaters recovery action. Failure to transfer credited power to the heaters could challenge PSW Power the Pressure Control Nuclear Safety Performance Criterion.
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| Transfer Switches This normally closed, required closed valve isolates the flow path from the RCS to 3RC VA0066 the Quench Tank. Fire induced cable damage may result in spurious opening of Pressurizer Power WP3-11 the PORV causing a loss of RC inventory and RC subcooling. This could Operated Relief challenge the Inventory and Pressure Control Nuclear Safety Performance Valve Criteria.
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| OffiCIAL USE ONLY SECURITY RELATED INfORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 300 Component VFDR# VFDR Description (Cables)
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| These normally closed, required closed valves isolate flow paths from the RCS to containment. A series of potential hot shorts within the terminal box of the 3RC VA0157, 3RC electrical penetration may spuriously open the reactor head vent and hot leg vent VA0159 - RC Hot WP3-12 valves. The spurious opening of these valves may result in a loss of RC inventory Leg and Head Vent and challenge the Inventory and Pressure Control Nuclear Safety Performance Valves Criteria.
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| These normally closed valves isolate the flow path from the RCS to containment.
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| These valves are required to open to provide an RC letdown flow path. Fire 3RC VA0159, 3RC WP3-13 induced cable damage may prevent these valves from being opened resulting in VA0160 - RC Head the lifting of the pressurizer safety relief valves and a challenge to the Inventory Vent Valves Control Nuclear Safety Performance Criterion.
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| Although unaffected by fire, the power supplies for the station HVAC system are not credited following a fire in this fire area and a loss of power may result in the Unit 3 Control WP3-15 temperature inside the Unit 3 control complex exceeding the operability limit of Complex Cooling SSD components and challenge the Vital Auxiliaries Nuclear Safety Performance I Criterion.
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| Fire damage to cables may result in a loss of power to the containment cooling system and may result in the temperature inside the Unit 2 RB exceeding the Unit 3 Containment WP3-16 operability limit of SSD components. This could challenge the Vital Auxiliaries Cooling Nuclear Safety Performance Criterion.
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| Note: The additional risk added because of these VFDRs, as determined from the FRE for this fire area, is provided in SE Table 3.5.
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| Recovery Actions (RAs)
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| The licensee did not identify any recovery actions required for this fire area.
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| Detection & Suppression Systems Required to Meet the Nuclear Safety Performance Criteria The licensee performed an evaluation of the fire detection and suppression systems in this area.
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| The results of the evaluation were documented in LAR Table 4-4 and the applicable portions have been included below.
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| Partial detection is installed over electrical penetrations. Purge Inlet Room general area and/or hazard detection is required for risk criteria. West Penetration Room general area and/or hazard detection is required for DID. Modification is required to improve general area and/or hazard detection for Purge Inlet Room and West Penetration Room. These detection modifications are to improve plant fire detection and fire brigade response time.
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| F;~
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| Suppression Detection Required Required Auto System?
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| I Area Fire Zone Zone Description Suppression System? Detection Required?
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| Required?
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| E R D S E R D S Unit 3 West Penetration WP3 Room Unit 3 Cask Decon Tank WP3 87 No No No No No Yes Yes No No No Room Unit 3 West Penetration Yes WP3 98 No No No No No Yes No No No Pen Room (MR)
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| Yes WP3 114 Unit 3 Purge Inlet Room No No No No No Yes No No No (MR)
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| Leqend:
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 301 E - EEEE/LA: Systems required for acceptability of EEE Evaluations / NRC approved Licensing Action (Section 2.2.7)
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| R - Risk: Systems required to meet the Risk Criteria for the PB Approach (Section 4.2.4)
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| D - Defense-in-Depth: Systems required to maintain adequate balance of Defense-in-Depth for a PB Approach (Section 4.2.4.2)
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| S - Separation Criteria: Systems required for Chapter 4 Separation Criteria in (Section 4.2.3)
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| MR - Modification Required Systems are committed to be modified as indicated in LAR Table 4-4 and Attachment S Fire Area WP3 Conclusion The licensee has utilized the FRE PB approach to demonstrate the ability to meet the NFPA 805 nuclear safety performance criteria for this fire area. A FRE in accordance with NFPA 805, Section 4.2.4.2, in conjunction with deterministic methods for simplifying assumptions, was used in applying this approach.
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| Based on the information provided in the LAR, as supplemented, the NRC staff finds Fire Area WP3 meets the nuclear safety goals, objectives, and performance criteria of NFPA 805. This conclusion is based on the following:
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| * Fire protection SSCs were evaluated in accordance with NFPA 805, Chapter 4 to determine which, if any, were required to meet the nuclear safety performance criteria. This evaluation included:
| |
| : a. The fire protection detection systems required to meet the nuclear safety performance criteria were documented.
| |
| : b. Fire Area boundaries were defined using three hour rated walls, ceilings and floors, including fire barriers, fire barrier penetrations, and through penetration fire stops and spatial separation.
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| * Two exemptions from the pre-transition fire protection requirements were evaluated and found to be valid and applicable under the NFPA 805 RI/PB FPP.
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| * Twelve VFDRs were identified, evaluated through the performance of a FRE, and either found to meet the risk acceptance criteria, as well as the requirements for DID and SMs, or modifications were planned/ implemented to address the issue. The acceptability of the risk for this fire area is contingent on the risk reduction from the planned PSW modification (see SE Section 3.4 for a detailed discussion of the NRC staff's review of the adequacy of the FRE method used at ONS).
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| * This fire area did not require the use of recovery actions to meet the nuclear safety performance criteria.
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| * The following modifications were identified to address VFDRs:
| |
| : a. To improve general area and/or hazard detection for the Unit 3 Purge Inlet Room, Unit 3 West Penetration Pen Room and Unit 3 Cask Decon Tank Room.
| |
| : b. Fire barrier separating the Unit 3 Purge Inlet Room / SFP Area is required.
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| Fire Area YARD, Yard The licensee analyzed this fire area using the FRE approach in accordance with NFPA 805 Section 4.2.4.2, but also used deterministic simplifying assumptions in order to credit those OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 302 portions of the facility design that met the deterministic requirements of NFPA 805 Section 4.2.3. The licensee identified the SSCs necessary to meet the nuclear safety performance criteria in this fire area.
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| Evaluation of Fire Suppression Effects on Nuclear Safety Performance Criteria The licensee stated in Attachment C, "NEI 04-02, Table B-3, Fire Area Transition," that safe and stable conditions can be achieved and maintained using equipment and cables outside of the area of fire suppression activity. Flooding of the suppression areas and discharge of suppression water to adjacent compartments is controlled and will not jeopardize achievement of safe and stable conditions.
| |
| Based on the information provided by the licensee in the NFPA 805 LAR, the NRC staff finds the licensee's evaluation of fire suppression effects on nuclear safety performance criteria acceptable because the results of the licensee's analysis indicate that 'Fire suppression activities will not adversely affect achievement of the nuclear safety performance criteria.
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| Exemptions and Other Licensing Actions The licensee did not credit any previously approved licensing actions or exemptions from the existing fire protection requirements.
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| Variation from Deterministic Requirements (VFDRs)
| |
| Fire Area YARD has a total of 28 VFDRs, which are provided in the table below. All but one of these VFDRs are variances from NFPA 805, Section 4.2.3, (separation issues) that were dispositioned with a FRE (SE Section 3.4.3). The licensee determined that these variances are acceptable based on 1) the change in CDF and LERF for the fire area and the total CDF and LERF for each unit meet the acceptance criteria of RG 1.174 (SE Section 3.4.6) and 2) adequate DID and SMs are maintained for each fire area (SE Section 3.4.2). This determination relies on the following fire protection systems and features to meet the acceptance criteria:
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| * Fire suppression for Transformers CT-1, CT-2, and CT-3 is required to meet the DID criteria.
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| * Pre-fire plans will be updated to include fire brigade guidance for protection of the TB wall, combustible controls will be established, and vehicle traffic controls will be established in the vicinity of the Fire Area TB wall, transformers, and trenches to meet the DID criteria (SE Section 2.9, Table 2.9-1, Item 9).
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| The licensee does not require any system or barrier modifications because reliance on existing transformer suppression systems is sufficient to meet the criteria of DID. Fire brigade guidance update, combustible controls, and traffic control in the vicinity of the TB wall will be required to be in place to meet the criteria of DID.
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| One of the 28 VFDRs, YARD-04, is a variance from NFPA 805 Section 4.2.3 (separation issue) that will be corrected with a plant modification. According to the LAR, the wall separating the east YARD and the tornado vents of the Blockhouse 1 & 2 exterior wall currently is not "adequate for the hazard" as required by NFPA 805, Section 3.11.1, and all of the penetrations in the wall do not have a fire resistance rating as required by NFPA 805 ,Section 3.11.3. This OFFICIAL USE ONLY SECURITY RELATED INFORMATIO~J
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 303 wall is credited for area separation using the deterministic approach of NFPA 805, Section 4.2.3. The licensee has committed to make a modification to install hinged steel covers/shields to the exterior side of the tornado vents to qualify the wall 'adequate for the hazard' thereby bringing it into compliance with the requirements of NFPA 805 (SE Section 2.8.1).
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| Component VFDR# VFDR Description (Cables)
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| This normally closed, required closed valve provides train separation by isolating the cross connect header between the 1A and 1B trains of Emergency Feedwater. 1CCWVA0269 YARD-01 Fire induced cable damage may result in spurious opening of this valve, a diversion SGAFDW of flow to either the 1A or 1B SGs, and a challenge to the DHR Nuclear Safety Control MOV Performance Criterion. This valve may suffer IN 92-18 damage.
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| This normally open, required open valve is located in the EFW flow path to the 1B SG. Fire induced cable damage may result in spurious closing of this valve, isolating Protected Service Water flow to the 1B SG. The subsequent decrease in 1FDWVA0347 YARD-02 SG shell temperature may result in 1B SG exceeding its tube to shell differential SG B Inlet MOV temperature limit. This could challenge the Inventory Control and DHR Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 damaqe.
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| Normally open valve "I HP VA0023 is in the flow path from the LOST to the suction of the credited HPI pump. Normally closed valve 1HP VA0939 isolates the flow path from the LOST to the containment sump. Recirculation flow to the LOST during 1HP VA0023 - HPI prolonged operation of the HPI pump at low flow conditions may result in an Normal Suction increase in temperature of LOST contents to the operability limit of the HPI pump.
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| MOV,1HP YARD-03 The contents of the LOST must be diverted to the containment sump by opening VA0939 - LOST to 1HP VA0939 and closing 1HP VA0023 prior to the operability limit of the HPI pump Emergency Sump being exceeded to prevent challenging the Reactivity, Inventory and Pressure MOV Control Nuclear Safety Performance Criteria. Although unaffected by fire, the power supplies for these valves are not credited following a fire in this fire area and a loss of power may prevent these valves from being repositioned.
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| The penetrations in the wall interfacing the east wall of Blockhouse 1 & 2 and the Tornado Vents in east yard do not have a fire resistance rating as required by NFPA 805, Section YARD-04 Blockhouse 1 & 2 3.11.3. This wall is credited for area separation in the deterministic approach of Building Wall NFPA 805, Section 4.2.3.
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| These normally open, required closed valves isolate the flow path from the BWST to the LPI Pumps, RBS Pumps, and containment sump. Fire damage to cables for 1LP VA0021, 1LP electrical equipment supplying power to these valves may prevent these valves from YARD-05 VA0022 - BWST being closed resulting in excess BWST inventory loss to the containment sump and Suction MOVs a challenge to the Inventory and Reactivity Control Nuclear Safety Performance Criteria.
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| BWST level indication is required to monitor the performance of the reactivity and 1LPIP 0345 inventory control systems. Fire induced cable damage may result in loss of BWST YARD-06 BWST Level level indication and challenge the Process Monitoring Nuclear Safety Performance Indication Criterion 1MS VA0017, 1MS VA0024, 1MS VA0026, These normally open, required closed valves isolate flow paths from the MSHs . 1MS VA0033, Fire damage to cables for electrical equipment supplying power to these valves may 1MS VA0035, YARD-07 prevent these valves from being closed and could result in overcooling and 1MS VA0036, shrinkage of RC inventory. This could challenge the DHR Nuclear Safety 1MS VA0076, Performance Criterion. 1MS VA0079, 1MS VA0082, 1MS VA0084 - SG Isolation MOVs 1RC SXTRN001 ,
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| Pressurizer heaters are required for RC pressure control. The heaters receive non-1RC SXTRN002 credited power from Unit 1 and credited power from a PSW system power supply.
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| Pressurizer YARD-08 The transfer of credited power to the pressurizer heaters requires a recovery action.
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| Heaters PSW Failure to transfer credited power to the heaters could challenge the Pressure Power Transfer Control Nuclear Safety Performance Criterion.
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| Switches OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 304 Component VFDR# VFDR Description (Cables)
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| This normally closed, required closed valve provides train separation by isolating the cross connect header between the 2A and 28 trains of Emergency Feedwater. 2CCVVVA0269 YARD-10 Fire induced cable damage may result in spurious opening of this valve, a diversion SGA FDW of flow to either the 2A or 28 SGs, and a challenge to the DHR Nuclear Safety Control MOV Performance Criterion. This valve may suffer IN 92-18 damage This normally open, required open valve is located in the EFW flow path to the 28 SG. Fire induced cable damage may result in spurious closing of this valve, isolating Protected Service Water flow to the 28 SG. The subsequent decrease in 2FDVVVA0347 YARD-11 SG shell temperature may result in 28 SG exceeding its tube to shell differential SG 8 Inlet MOV temperature limit. This could challenge the Inventory Control and DHR Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 damaoe.
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| Normally open valve 2HP VA0023 is in the flow path from the LDST to the suction of the credited HPI pump. Normally closed valve 2HP VA0939 isolates the flow path from the LDST to the containment sump. Recirculation flow to the LDST during 2HP VA0023 - HPI prolonged operation of the HPI pump at low flow conditions may result in an Normal Suction increase in temperature of LDST contents to the operability limit of the HPI pump.
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| MOV,2HP YARD-12 The contents of the LDST must be diverted to the containment sump by opening VA0939 - LDST to 2HP VA0939 and closing 2HP VA0023 prior to the operability limit of the HPI pump Emergency Sump being exceeded to prevent challenging the Reactivity, Inventory and Pressure MOV Control Nuclear Safety Performance Criteria. Although unaffected by fire, the power supplies for these valves are not credited following a fire in this fire area and a loss of power may prevent these valves from being repositioned.
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| These normally open, required closed valves isolate the flow path from the 8WST to the LPI Pumps, R8S Pumps, and containment sump. Fire damage to cables for 2LP VA0021, 2LP electrical equipment supplying power to these valves may prevent these valves from YARD-14 VA0022 - 8WST being closed resulting in excess 8WST inventory loss to the containment sump and Suction MOVs a challenge to the Inventory and Reactivity Control Nuclear Safety Performance Criteria.
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| 8WST level indication is required to monitor the performance of the reactivity and 2LPIP 0345 inventory control systems. Fire induced cable damage may result in loss of 8WST YARD-15 8WST Level level indication and challenge the Process Monitoring Nuclear Safety Performance Indication Criterion.
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| 2MS VA0017, 2MS VA0024.
| |
| 2MS VA0026, These normally open, required closed valves isolate flow paths from the MSHs . 2MS VA0033, Fire damage to cables for electrical equipment supplying power to these valves may 2MS VA0035, YARD-16 prevent these valves from being closed and could result in overcooling and 2MS VA0036.
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| shrinkage of RC inventory. This could challenge the DHR Nuclear Safety 2MS VA0076, Performance Criterion. 2MS VA0079.
| |
| 2MS VAOO82, 2MS VA0084 - SG Isolation MOVs 2RC SXTRN001, Pressurizer heaters are required for RC pressure control. The heaters receive non- 2RC SXTRN002, credited power from Unit 2 and credited power from a PSW system power supply. 2RC SXTRN003 YARD-17 The transfer of credited power to the pressurizer heaters requires a recovery action. Pressurizer Failure to transfer credited power to the heaters could challenge the Pressure Heaters PSW Control Nuclear Safety Performance Criterion. Power Transfer Switches This normally closed, required closed valve provides train separation by isolating the cross connect header between the 3A and 38 trains of Emergency Feedwater. 3CCVVVA0269 YARD-19 Fire induced cable damage may result in spurious opening of this valve, a diversion SGA FDW of flow to either the 3A or 38 SGs, and a challenge to the DHR Nuclear Safety Control MOV Performance Criterion. This valve may suffer IN 92-18 damage.
| |
| This normally open, required open valve is located in the EFW flow path to the 38 SG. Fire induced cable damage may result in spurious closing of this valve, 3FDVVVA0347 YARD-20 isolating Protected Service Water flow to the 38 SG. The subsequent decrease in SG 8 Inlet MOV SG shell temperature may result in 38 SG exceeding its tube to shell differential OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 305 Component VFDR# VFDR Description (Cables) temperature limit. This could challenge the Inventory Control and DHR Nuclear Safety Performance Criteria. This valve may suffer IN 92-18 damaqe.
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| Normally open valve 3HP VA0023 is in the flow path from the LDST to the suction of the credited HPI pump. Normally closed valve 3HP VA0939 isolates the flow path from the LDST to the containment sump. Recirculation flow to the LDST during 3HP VA0023 - HPI prolonged operation of the HPI pump at low flow conditions may result in an Normal Suction increase in temperature of LDST contents to the operability limit of the HPI pump.
| |
| MOV,3HP YARD-21 The contents of the LDST must be diverted to the containment sump by opening VA0939 - LDST to 3HP VA0939 and closing 3HP VA0023 prior to the operability limit of the HPI pump Emergency Sump being exceeded to prevent challenging the Reactivity, Inventory and Pressure MOV Control Nuclear Safety Performance Criteria. Although unaffected by fire, the power supplies for these valves are not credited following a fire in this fire area and a loss of power may prevent these valves from beinq repositioned.
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| These normally open, required closed valves isolate the flow path from the BWST to the LPI Pumps, RBS Pumps, and containment sump. Fire damage to cables for 3LP VA0021, 3LP electrical equipment supplying power to these valves may prevent these valves from YARD-23 VA0022 - BWST being closed resulting in excess BWST inventory loss to the containment sump and Suction MOVs a challenge to the Inventory and Reactivity Control Nuclear Safety Performance Criteria.
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| BWST level indication is required to monitor the performance of the reactivity and 3LPIP 0345 inventory control systems. Fire induced cable damage may result in loss of BWST YARD-24 BWST Level level indication and challenge the Process Monitoring Nuclear Safety Performance Indication Criterion.
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| 3MS VA0017, 3MS VA0024, 3MS VA0026, These normally open, required closed valves isolate flow paths from the MSHs . 3MS VA0033, Fire damage to cables for electrical equipment supplying power to these valves may 3MS VA0035, YARD-25 prevent these valves from being closed and could result in overcooling and 3MS VA0036, shrinkage of RC inventory. This could challenge the DHR Nuclear Safety 3MS VA0076, Performance Criterion. 3MS VA0079, 3MS VA0082, 3MS VA0084 - SG Isolation MOVs The Reactor Coolant Pumps (RCPs) are required off when SSD is being 3RC PUOO01, accomplished by the PSW system. Unit 3 6900V RCP SWGR is located in Fire 3RC PUOO02, YARD-26 Area YARD. Fire damage to cables may result in an inability to secure the RCPs or 3RC PUOO03, result in a spurious pump start. This will place the unit in an unanalyzed condition 3RC PUOO04 and challenge the DHR Nuclear Safety Performance Criterion. RCPs 3RC SXTRN001, Pressurizer heaters are required for RC pressure control. The heaters receive non- 3RC SXTRNOO2, credited power from Unit 3 and credited power from a PSW system power supply. 3RC SXTRN003 YARD-27 The transfer of credited power to the pressurizer heaters requires a recovery action. Pressurizer Failure to transfer credited power to the heaters could challenge the Pressure Heaters PSW Control Nuclear Safety Performance Criterion. Power Transfer Switches Fire damage to cables for electrical equipment supplying power to the station HVAC system may result in the temperature inside the Units 1 & 2 control complex Units 1 & 2 Control YARD-29 exceeding the operability limit of SSD components and challenge the Vital Complex Cooling Auxiliaries Nuclear Safety Performance Criterion.
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| Fire damage to cables for electrical equipment supplying power to the station HVAC system may result in the temperature inside the Unit 3 control complex exceeding Unit 3 Control YARD-30 the operability limit of SSD components and challenge the Vital Auxiliaries Nuclear Complex Cooling Safety Performance Criterion.
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| Fire damage to cables for electrical equipment supplying power to the containment Unit 1 cooling system may result in the temperature inside the Unit 1 RB exceeding the YARD-31 Containment operability limit of SSD components and challenge the Vital Auxiliaries Nuclear Cooling Safety Performance Criterion.
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| YARD-32 Fire damaqe to cables for electrical equipment supplyinq power to the containment Unit 2 OFFICIAL USE O~JLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 306 Component VFDR# VFDR Description (Cables) cooling system may result in the temperature inside the Unit 2 RB exceeding the Containment operability limit of SSD components and challenge the Vital Auxiliaries Nuclear Cooling Safety Performance Criterion.
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| Fire damage to cables for electrical equipment supplying power to the containment Unit 3 cooling system may result in the temperature inside the Unit 3 RB exceeding the YARD-33 Containment operability limit of SSD components and challenge the Vital Auxiliaries Nuclear Cooling Safety Performance Criterion.
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| Note: The additional risk added because of these VFDRs, as determined from the FRE for this fire area, is provided in SE Table 3.5.
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| Recovery Actions (RAs)
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| The licensee did not identify any recovery actions required for this fire area.
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| Fire Detection & Suppression Systems Required to Meet the Nuclear Safety Performance Criteria The licensee performed an evaluation of the fire detection and suppression systems in this area.
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| The results of the evaluation were documented in LAR Table 4-4 and the applicable portions have been included below. Partial fire suppression is installed in the YARD fire area.
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| Suppression for transformers CT-1, CT-2, and CT-3 is required for DID. Suppression on transformers 1T, 2T, and 3T is required by Engineering Evaluation. Partial detection is installed in YARD fire area. Detection for Unit 3 RCP SWGR is required for DID.
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| Detection Suppression Required Auto Required System?
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| Fire Fire Detection System?
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| Zone Description Suppression Area Zone Provided?
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| Provided?
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| E R D S E R D S YARD Yard 230 KV Switchyard and YARD SYDYARD No No No No No No No No No No Relav House YARD TRENCH Cable Trench T-100 No No No No No No No No No No YARD Yard- East Yard Area - East Yes Yes No Yes No Yes No No Yes No Yard -
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| YARD Yard Area- West No No No No No No No No No No West Legend:
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| E - EEEE/LA: Systems required for acceptability of EEE Evaluations I NRC approved Licensing Action (Section 2.2.7)
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| R - Risk: Systems required to meet the Risk Criteria for the PB Approach (Section 4.2.4)
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| D - Defense-in-Depth: Systems required to maintain adequate balance of Defense-in-Depth for a PB Approach (Section 4.2.4.2)
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| S - Separation Criteria: Systems required for NFPA 805 Chapter 4 Separation Criteria in Section 4.2.3 MR- Modification Required Systems are committed to be modified as indicated in Table 4-4 and Attachment S of TR Fire Area YARD Conclusion The licensee has utilized the FRE PB approach to demonstrate the ability to meet the NFPA 805 nuclear safety performance criteria for this fire area. A FRE in accordance with NFPA 805, Section 4.2.4.2, in conjunction with deterministic methods for simplifying assumptions, was used in applying this approach.
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| OFFICIAL USE Ot>lLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 307 Based on the information provided in the LAR, as supplemented, the NRC staff finds Fire Area YARD meets the nuclear safety goals, objectives, and performance criteria of NFPA 805. This conclusion is based on the following:
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| * Fire protection SSCs were evaluated in accordance with NFPA 805 Chapter 4 to determine which, if any, were required to meet the nuclear safety performance criteria. This evaluation included:
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| : a. The fire protection detection systems required to meet the nuclear safety performance criteria were documented.
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| : b. Fire Area boundaries were defined using three hour rated walls, ceilings and floors, including fire barriers, fire barrier penetrations, and through penetration fire stops and spatial separation.
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| * No exemptions or licensing actions from the pre-transition fire protection requirements were required.
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| * Twenty-eight VFDRs were identified, evaluated through the performance of a FRE, and either found to meet the risk acceptance criteria, as well as the requirements for DID and SMs, or modifications were planned / implemented to address the issue. The acceptability of the risk for this fire area is contingent on the risk reduction from the planned PSW modification (see SE Section 3.4 for a detailed discussion of the NRC staff's review of the adequacy of the FRE method used at ONS).
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| * This fire area did not require the use of recovery actions to meet the nuclear safety performance criteria.
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| * The following modifications were identified to address VFDRs:
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| : a. Modification to install hinged steel covers/shields to exterior side of tornado vents to support FRE of separation between Units 1 & 2 Block House and the East Yard.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY - SECURITY-RELATED INFORMATION 308 Attachment E, Radioactive Release Tables In order to assess whether the ONS FPP to be implemented under NFPA 805 meets the radioactive release performance criteria, the licensee reviewed the existing ONS pre-fire plans and fire brigade training materials. Pre-fire plans that address fire areas/zones where there is no possibility of radioactive materials being present were screened from further review. All other pre-fire plans were reviewed to ascertain whether existing engineering controls are adequate to ensure that radioactive materials (contamination) generated as a direct result of fire suppression activities are contained and monitored before release to unrestricted areas, such that the release would meet the NFPA 805 radioactive release performance criteria.
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| The licensee's review determined that existing engineering controls, such as drains and forced air ventilation, supplemented by pre-fire plans and fire brigade training, were adequate to meet the NFPA 805 radioactive release requirements. In addition, the licensee identified the need for monitoring and control of potentially contaminated run-off into non-contaminated areas and developed a new fire brigade instruction (SOG-16) to address this need, which has been incorporated into fire brigade training. Chemistry/radiation protection (RP) personnel are part of the responding fire brigade team.
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| The licensee stated that current RP procedures and practices adequately describe how to monitor and control liquid and gaseous effluents from the site.
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| This attachment contains Table 3.6-1, "ONS Fire Areas and Their Compliance with the NFPA 805 Radioactive Release Performance Criteria," which summarizes, for each pre-fire plan, (1) the fire areas included in the pre-fire plan, (2) the engineered controls used to minimize radioactive releases generated from the combustion of radioactive materials or from fire suppression activities, and (3) the NRC staff evaluation of the adequacy of these engineered controls and fire brigade training in meeting the NFPA 805 Radioactive Release Performance Criteria.
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| OFFICIAL USE O~JLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 309 Attachment E, Radioactive Release Tables Table 3.6-1, Compliance with NFPA 805 Radioactive Release Performance Criteria Effluent Engineered Controls Fire Area RCA or Pre-Fire Plan Rev NRC Staff Evaluation Fire Zone RCZ Suppression I Combustion Water Smoke TB Turbine Building The NRC staff finds the licensee's statement that 1 Lube Oil Purification Pad 1 No N/A N/A the area has no radioloaical hazards acceotable.
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| The NRC staff finds the licensee's statement that 2 EHC Area 0 No N/A N/A the area has no radioloaical hazards acceotable.
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| The NRC staff finds the licensee's statement that 3 Heater Bay Area 1 No N/A N/A the area has no radioloaical hazards acceotable.
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| The NRC staff finds the licensee's statement that 4 TDEFDW Pump 0 No N/A N/A the area has no radioloaical hazards acceotable.
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| The NRC staff finds the licensee's statement that 5 Condensate Booster Pump 0 No N/A N/A the area has no radioloaical hazards acceotable.
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| The NRC staff finds the licensee's statement that 6 Main Feedwater Pump Area 0 No N/A N/A the area has no radioloaical hazards acceotable.
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| Motor-Driven Emergency Feedwater The NRC staff finds the licensee's statement that 7 0 No N/A N/A (MDEFD\IV) Pump the area has no radiological hazards acceotable.
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| Hotwell Pump and The NRC staff finds the licensee's statement that 8 0 No N/A N/A TB Sump Oil Skimmer the area has no radiological hazards acceotable.
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| The NRC staff finds the licensee's statement that 9 PowdexlLPSW Pump 0 No N/A N/A the area has no radiological hazards acceotable.
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| The NRC staff finds the licensee's statement that 10 Unit 2 Lube Oil Purification 1 No N/A N/A the area has no radiological hazards acceotable.
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| The NRC staff finds the licensee's statement that 11 Unit 2 EHC Area 0 No N/A N/A the area has no radiological hazards acceotable.
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| The NRC staff finds the licensee's statement that 12 Unit 2 Heater Bay 0 No N/A N/A the area has no radiological hazards acceotable.
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| The NRC staff finds the licensee's statement that 13 Unit 2 TDEFDWP 0 No N/A N/A the area has no radiological hazards acceotable.
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| The NRC staff finds the licensee's statement that 14 Unit 2 Condensate Booster Pump 0 No N/A N/A the area has no radiological hazards acceotable.
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| The NRC staff finds the licensee's statement that 15 Unit 2 Main Feedwater Pump Area 1 No N/A N/A the area has no radiological hazards acceptable.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 310 Attachment E, Radioactive Release Tables Table 3.6-1, Compliance with NFPA 805 Radioactive Release Performance Criteria Effluent Engineered Controls Fire Area RCA or Pre-Fire Plan Rev NRC Staff Evaluation Fire Zone RCZ Suppression Combustion Water Smoke The NRC staff finds the licensee's statement that 16 Unit 2 MDEFDWP 0 No N/A N/A the area has no radioloqical hazards acceotable.
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| The NRC staff finds the licensee's statement that 17 Unit 2 HWP,LPSW-B Area 0 No N/A N/A the area has no radioloqical hazards acceotable.
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| The NRC staff finds the licensee's statement that 17A HPSWB Area 0 No N/A N/A the area has no radioloqical hazards acceotable.
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| Unit 2 Powdex, Backup 1A Compressor The NRC staff finds the licensee's statement that 18 0 No N/A N/A and Control Room Chillers the area has no radioloqical hazards acceotable.
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| The NRC staff finds the licensee's statement that 19 Unit 1 Main Feed Water Pump Area 0 No N/A N/A the area has no radioloaical hazards acceotable.
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| The NRC staff finds the licensee's statement that 20 Unit 1 MDEFDWP and Seal Oil Area 0 No N/A N/A the area has no radioloaical hazards acceotable.
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| The NRC staff finds the licensee's statement that 21 Unit 1 HWP,LPSW-A Area 0 No N/A N/A the area has no radioloaical hazards acceotable.
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| The NRC staff finds the licensee's statement that 21A HPSW Pump A 0 No N/A N/A the area has no radioloaical hazards acceotable.
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| The NRC staff finds the licensee's statement that 22 Unit 1 Powdex Area 0 No N/A N/A the area has no radioloqical hazards acceotable.
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| The NRC staff finds the licensee's statement that 22A Lube Oil Storage House 1 No N/A N/A the area has no radioloqical hazards acceotable.
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| The NRC staff finds the licensee's statement that 23 Unit 1 Condensate Booster Pump 0 No N/A N/A the area has no radioloqical hazards acceotable.
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| Unit 1 TDEFDW Pump, The NRC staff finds the licensee's statement that 24 0 No N/A N/A EHC, Turbine and Lube Oil Purification the area has no radioloqical hazards acceotable.
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| The NRC staff finds the licensee's statement that 25 Unit 1 Feed Water Heater Area 0 No N/A N/A the area has no radioloqical hazards acceotable.
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| The NRC staff finds the licensee's statement that 26 Moisture Separators B1 & B2 0 No N/A N/A the area has no radioloaical hazards acceotable.
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| Turbine Oil Tank and The NRC staff finds the licensee's statement that 27 0 No N/A N/A MS Stoo and Control Valves the area has no radioloaical hazards acceotable.
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| The NRC staff finds the licensee's statement that 28 Heater Bay Area 0 No N/A N/A the area has no radioloaical hazards acceotable.
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| The NRC staff finds the licensee's statement that 29 4160 Switchgear 1 No N/A N/A the area has no radiological hazards acceptable.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 311 Attachment E, Radioactive Release Tables Table 3.6-1, Compliance with NFPA 805 Radioactive Release Performance Criteria Effluent Engineered Controls Fire Area RCA or Pre-Fire Plan Rev NRC Staff Evaluation Fire Zone RCZ Suppression Combustion Water Smoke The NRC staff finds the licensee's statement that 30 Unit 2 Moisture Separator B1 & B2 0 No N/A N/A the area has no radioloaical hazards acceptable.
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| Unit 2 MS Stop and Control Valves The NRC staff finds the licensee's statement that 31 0 No N/A N/A and Main Turbine Oil Tank (MTOT) the area has no radioloaical hazards acceptable.
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| The NRC staff finds the licensee's statement that 32 Unit 2 Heater Bay Area 1 No N/A N/A the area has no radioloaical hazards acceptable.
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| The NRC staff finds the licensee's statement that 33 Unit 2 4160 Switchgear 1 No N/A N/A the area has no radioloaical hazards acceptable.
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| The NRC staff finds the licensee's statement that 33A Unit 2 Power Batteries 0 No N/A N/A the area has no radioloaical hazards acceptable.
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| The NRC staff finds the licensee's statement that 34 Unit 1 6900/4160V Switchgear 1 No N/A N/A the area has no radioloaical hazards acceotable.
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| The NRC staff finds the licensee's statement that 34A Unit 1 Power Batteries PA 0 No N/A N/A the area has no radioloaical hazards acceotable.
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| The NRC staff finds the licensee's statement that 35 Unit 1 Heater Bay Area 0 No N/A N/A the area has no radioloaical hazards acceotable.
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| Unit 1 MS Stop Valves, The NRC staff finds the licensee's statement that 36 0 No N/A N/A MSRH's and MTOT the area has no radioloaical hazards acceotable.
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| MCC 2X11, 2X11A, 3X5 The NRC staff finds the licensee's statement that 37 1 No N/A N/A and 3X6 Areas the area has no radioloaical hazards acceotable.
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| The NRC staff finds the licensee's statement that 38 Main Turbine and Offices 1 No N/A N/A the area has no radioloaical hazards acceotable.
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| Heater Bay and The NRC staff finds the licensee's statement that 39 0 No N/A N/A UODer Surae Tanks the area has no radioloaical hazards acceotable.
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| The NRC staff finds the licensee's statement that 39A Turbine Deck and Offices 0 No N/A N/A the area has no radioloaical hazards acceotable.
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| The NRC staff finds the licensee's statement that 40 Unit 2 Main TB Deck 0 No N/A N/A the area has no radioloqical hazards acceotable.
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| Unit 2 Heater Bay and The NRC staff finds the licensee's statement that 41 0 No N/A N/A Upper Surae Tank the area has no radioloaical hazards acceptable.
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| The NRC staff finds the licensee's statement that 42 Unit 1 Main Turbine 0 No N/A N/A the area has no radiological hazards acceptable.
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| The NRC staff finds the licensee's statement that Unit 1 Heater Bay and 43 0 No N/A N/A the area has no radiological hazards acceptable.
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| Upper Surge Tanks OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 312 Attachment E, Radioactive Release Tables Table 3.6-1, Compliance with NFPA 805 Radioactive Release Performance Criteria Effluent Engineered Controls Fire Area RCA or Pre-Fire Plan Rev NRC Staff Evaluation Fire Zone RCZ Suppression Combustion Water Smoke The NRC staff finds the licensee's statement that 44 TB Truck Receiving Bay a No N/A N/A the area has no radiolooical hazards acceptable.
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| The NRC staff finds the licensee's statement that BH12-45 Units 1/2 Block House a No N/A N/A the area has no radioJooical hazards acceptable.
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| The NRC staff finds the licensee's statement that CT4 - 46 CT-4 Block House a No N/A N/A the area has no radiolooical hazards acceptable.
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| The NRC staff finds the licensee's statement that BH3 - 47 Unit 3 Block House a No N/A N/A the area has no radiolooical hazards acceptable.
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| Pre-fire Plans specify Floor Drains ventilation paths routed to Based on the availability of engineered controls for smoke Radwaste and fire brigade training for both fire suppression control, 3A LPI and Processing agent run-off and smoke as described in the LAR, 48 RBS Pumps a Yes system for radiation the NRC staff concludes that, the licensee's indicating monitoring and approach is acceptable to meet the NFPA 805 alarms (RIAs) processing Radioactive Release Performance Criteria.
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| monitor prior to release contamination levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Unit 3 C LPI and Processing for smoke as described in the LAR, the NRC staff concludes 49 B RBS Pumps a Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 50 3C HPI Pump a Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 313 Attachment E, Radioactive Release Tables Table 3.6-1, Compliance with NFPA 805 Radioactive Release Performance Criteria Effluent Engineered Controls Fire Area RCA or Pre-Fire Plan Rev NRC Staff Evaluation Fire Zone RCZ Suppression Combustion Water Smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke HPI, Spent resin, UH AWT Processing for smoke as described in the LAR, the NRC staff concludes 50A 0 Yes and Comp Drain Pumps system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Purifications Demineralizer Room Processing for smoke as described in the LAR, the NRC staff concludes 51 0 Yes and Hatch Area system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Low Pressure Injection Radwaste ventilation paths for both fire suppression agent run-off and smoke Pump's (LPIP's) 2A, 2C Processing for smoke as described in the LAR, the NRC staff concludes 52 1 Yes and Reactor Building Spray system for control, RIAs that, the licensee's approach is acceptable to Pump (RBSP) 2A monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke LPIP's 1B, 2B and Processing for smoke as described in the LAR, the NRC staff concludes 53 0 Yes RBSP 1B & 2B system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans Based on the availability of engineered controls LPIP's 1A, routed to specify for both fire suppression agent run-off and smoke 54 0 Yes 1C & RBSP 1A Radwaste ventilation paths as described in the LAR, the NRC staff concludes Processinq for smoke that, the licensee's approach is acceptable to OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 314 Attachment E, Radioactive Release Tables Table 3.6-1, Compliance with NFPA 805 Radioactive Release Performance Criteria Effluent Engineered Controls Fire Area RCA or Pre-Fire Plan Rev NRC Staff Evaluation Fire Zone RCZ Suppression Combustion Water Smoke system for control, RIAs meet the NFPA 805 Radioactive Release monitoring and monitor Performance Criteria.
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| processing contamination prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 55 HPIP's 1A & 1B 1 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 55A HPIP's 1C & 2C 1 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation for both fire suppression agent run-off and smoke Processing paths, RIAs for as described in the LAR, the NRC staff concludes 56 HPIP's 2A & 2B 1 Yes system for smoke control that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Purification & Deborating Processing for smoke as described in the LAR, the NRC staff concludes 57 0 Yes Demineralizer system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 315 Attachment E, Radioactive Release Tables Table 3.6-1, Compliance with NFPA 805 Radioactive Release Performance Criteria Effluent Engineered Controls Fire Area RCA or Pre-Fire Plan Rev NRC Staff Evaluation Fire Zone RCZ Suppression Combustion Water Smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke 1st Floor Hallwayffank Room Processing for smoke as described in the LAR, the NRC staff concludes 58 0 Yes and 2nd Floor Tank Rooms system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke LPI Cooler, Processing for smoke as described in the LAR, the NRC staff concludes 59 0 Yes HPI Seal Return Coolers system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 60 LPI Hatch Area 0 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 61 HPI Hatch area 0 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans Based on the availability of engineered controls routed to specify for both fire suppression agent run-off and smoke 62 Waste Disposal Control Room 0 Yes Radwaste ventilation paths as described in the LAR, the NRC staff concludes ProcessinQ for smoke that, the licensee's approach is acceptable to OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 316 Attachment E, Radioactive Release Tables Table 3.6-1, Compliance with NFPA 805 Radioactive Release Performance Criteria Effluent Engineered Controls Fire Area RCA or Pre-Fire Plan Rev NRC Staff Evaluation Fire Zone RCZ Suppression Combustion Water Smoke system for control, RIAs meet the NFPA 805 Radioactive Release monitoring and monitor Performance Criteria.
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| processing contamination prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 63 Letdown Storage Tank 0 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Auxiliary Service Water Processing for smoke as described in the LAR, the NRC staff concludes 64 0 Yes and Switchgear system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
| |
| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Unit 2 Decay Heat Cooler, Processing for smoke as described in the LAR, the NRC staff concludes 65 0 Yes Seal Supply Filter and CRD system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 66 HPI Hatch area 0 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 317 Attachment E, Radioactive Release Tables Table 3.6-1, Compliance with NFPA 805 Radioactive Release Performance Criteria Effluent Engineered Controls Fire Area RCA or Pre-Fire Plan Rev NRC Staff Evaluation Fire Zone RCZ Combustion Suppression Water Smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Units 1&2 Hatch Area and Processing for smoke as described in the LAR, the NRC staff concludes 67 & 70 0 Hot Machine Shop Tunnel system for control, RlAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
| |
| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 68 & 72 Units 1 & 2 HPI Hatch area 0 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
| |
| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 69 Units 1 & 2 Waste Control Panel 1 system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
| |
| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Unit 2 Letdown Storage Tank Processing for smoke as described in the LAR, the NRC staff concludes 71 0 Yes and Letdown Filter system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
| |
| prior to release levels of smoke Floor Drains Pre-fire Plans Based on the availability of engineered controls Unit 1 Letdown Storage Tank routed to specify for both fire suppression agent run-off and smoke 73 0 and Filter Room Radwaste ventilation paths as described in the LAR, the NRC staff concludes ProcessinQ for smoke that, the licensee's approach is acceptable to OFFICIAL USE ONLY SECURITY RELATED INFORMATION
| |
| | |
| OFFICIAL USE ONLY SECURITY RELI\TED INFORMATION 318 Attachment E, Radioactive Release Tables Table 3.6-1, Compliance with NFPA 805 Radioactive Release Performance Criteria Effluent Engineered Controls Fire Area RCA or Pre-Fire Plan Rev NRC Staff Evaluation Fire Zone RCZ Combustion Suppression Water Smoke system for control, RIAs meet the NFPA 805 Radioactive Release monitoring and monitor Performance Criteria.
| |
| processing contamination prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Unit 1 Post-Accident Liquid Sample Processing for smoke as described in the LAR, the NRC staff concludes 74 0 Yes (PALS) Room system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
| |
| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Unit 1 LPI Cooler, Processing for smoke as described in the LAR, the NRC staff concludes 75 Pipe Chase, 0 Yes system for control, RIAs that, the licensee's approach is acceptable to CRD Filter Room monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
| |
| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke IA & 1B BHUT, CBAST Processing for smoke as described in the LAR, the NRC staff concludes 76 0 Yes and RC Bleed Transfer Pump system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
| |
| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 77 Second Floor Hallway 0 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 319 Attachment E, Radioactive Release Tables Table 3.6-1, Compliance with NFPA 805 Radioactive Release Performance Criteria Effluent Engineered Controls Fire Area RCA or Pre-Fire Plan Rev NRC Staff Evaluation Fire Zone RCZ Suppression Combustion Water Smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 78 Spent Fuel Cooler Room a Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 79 Component Cooler Room a Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Waste Gas Compressor Processing for smoke as described in the LAR, the NRC staff concludes 80 and Tank Area a Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 81 Units 1 & 2 Second Floor Hallway a Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans Based on the availability of engineered controls Units 1 & 3 Spent Fuel Coolant routed to specify for both fire suppression agent run-off and smoke 82 Pump and Cooler a Yes Radwaste ventilation paths as described in the LAR, the NRC staff concludes Processinq for smoke that, the licensee's approach is acceptable to OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 320 Attachment E, Radioactive Release Tables Table 3.6-1, Compliance with NFPA 805 Radioactive Release Performance Criteria Effluent Engineered Controls Fire Area RCA or Pre*Fire Plan Rev NRC Staff Evaluation Fire Zone RCZ Suppression Combustion Water Smoke system for control, RIAs meet the NFPA 805 Radioactive Release monitoring and monitor Performance Criteria.
| |
| processing contamination prior to release levels of smoke Floor Drains Pre*fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Units 1 &2 Component Processing for smoke as described in the LAR, the NRC staff concludes 83 0 Yes Cooling Pump system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
| |
| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Units 1 & 2 Waste Gas Processing for smoke as described in the LAR, the NRC staff concludes 84 0 Yes Compressor and Tank system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
| |
| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 85 Unit 1 Second Floor Hallway 0 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
| |
| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke 3rd Floor Hallway, Change Rooms, Processing for smoke as described in the LAR, the NRC staff concludes 86 0 Yes Hatch and Lab Areas system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OF=F=ICIAL USE ONLY SECURITY RELATED INF=ORMATION 321 Attachment E, Radioactive Release Tables Table 3.6-1, Compliance with NFPA 805 Radioactive Release Performance Criteria Effluent Engineered Controls Fire Area RCA or Pre-Fire Plan Rev NRC Staff Evaluation Fire Zone RCZ Suppression Combustion Water Smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 87 Cask Decon Room 0 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
| |
| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 88 Spent Fuel Receiving Bay 0 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
| |
| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 89 Equipment Room 2 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
| |
| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Unit 2 Hallway, Change Room, Processing for smoke as described in the LAR, the NRC staff concludes 90 0 Yes Laundry RM, and RP Lab system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
| |
| prior to release levels of smoke Floor Drains Pre-fire Plans Based on the availability of engineered controls routed to specify for both fire suppression agent run-off and smoke 91 Unit 2 Cask Decon Room 1 Yes Radwaste ventilation paths as described in the LAR, the NRC staff concludes Processinq for smoke that, the licensee's approach is acceptable to OF=F=ICIAL USE ONLY SECURITY RELATED INF=ORMATION
| |
| | |
| OFFICIAL USE O~JLY SECURITY RELATED INFORMATIO~J 322 Attachment E, Radioactive Release Tables Table 3.6-1, Compliance with NFPA 805 Radioactive Release Performance Criteria Effluent Engineered Controls Fire Area RCA or Pre-Fire Plan Rev NRC Staff Evaluation Fire Zone RCZ Suppression Combustion Water Smoke system for control, RIAs meet the NFPA 805 Radioactive Release monitoring and monitor Performance Criteria.
| |
| processing contamination prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 92 Unit 2 Equipment Room 1 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
| |
| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Units 1 & 2 Fuel Loading Area Processing for smoke as described in the LAR, the NRC staff concludes 93 0 Yes and Spent Fuel system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
| |
| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Unit 1 Hatch, Change Room Processing for smoke as described in the LAR, the NRC staff concludes 94 0 Yes and Tool Storage system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
| |
| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 95 Unit 1 Equipment Room 1 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 323 Attachment E, Radioactive Release Tables Table 3.6-1, Compliance with NFPA 805 Radioactive Release Performance Criteria Effluent Engineered Controls Fire Area RCA or Pre-Fire Plan Rev NRC Staff Evaluation Fire Zone RCZ Suppression Combustion Water Smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 96 Hot Machine Shop 1 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
| |
| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 97 Unit 1 Cask Decon Room 1 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 98 West Penetration Room 0 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 99 East Penetration Room 0 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans Based on the availability of engineered controls routed to specify for both fire suppression agent run-off and smoke 100 Control Battery Room 0 Yes Radwaste ventilation paths as described in the LAR, the NRC staff concludes ProcessinQ for smoke that, the licensee's approach is acceptable to OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 324 Attachment E, Radioactive Release Tables Table 3.6-1, Compliance with NFPA 805 Radioactive Release Performance Criteria Effluent Engineered Controls Fire Area RCA or Pre-Fire Plan Rev NRC Staff Evaluation Fire Zone RCZ Suppression Combustion Water Smoke system for control, RIAs meet the NFPA 805 Radioactive Release monitoring and monitor Performance Criteria.
| |
| processing contamination prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 101 Cable Room and Elevator lobby 1 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 102 Unit 2 West Penetration Room 0 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 103 Unit 2 East Penetration Room 0 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 104 Unit 2 Control Battery Room 1 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 325 Attachment E, Radioactive Release Tables Table 3.6-1, Compliance with NFPA 805 Radioactive Release Performance Criteria Effluent Engineered Controls Fire Area RCA or Pre-Fire Plan Rev NRC Staff Evaluation Fire Zone RCZ Suppression Combustion Water Smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 105 Unit 2 Cable Room 1 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 106 Unit 1 Cable Room 0 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 107 Unit 1 West Penetration Room 0 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 108 Unit 1 East Penetration Room 0 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans Based on the availability of engineered controls routed to specify for both fire suppression agent run-off and smoke 109 Unit 1 Control Battery Room 0 Yes Radwaste ventilation paths as described in the LAR, the NRC staff concludes ProcessinQ for smoke that, the licensee's approach is acceptable to OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 326 Attachment E, Radioactive Release Tables Table 3.6-1, Compliance with NFPA 805 Radioactive Release Performance Criteria Effluent Engineered Controls Fire Area RCAor Pre-Fire Plan Rev NRC Staff Evaluation Fire Zone RCZ Suppression Combustion Water Smoke system for control, RIAs meet the NFPA 805 Radioactive Release monitoring and monitor Performance Criteria.
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| processing contamination prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Unit 1 AHU, Storage, Processing for smoke as described in the LAR, the NRC staff concludes 109A 0 Yes and Control Room system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke The NRC staff finds the licensee's statement that 110 Units 1 & 2 Control Room 0 No N/A N/A the area has no radiological hazards acceptable Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Unit 2 AHU Room, Storage, Processing for smoke as described in the LAR, the NRC staff concludes 111 0 Yes and Control Room Lobby system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke The NRC staff finds the licensee's statement that 112 Unit 3 Control Room 0 No N/A N/A the area has no radioloQical hazards acceptable Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 113 AHU, Control Room Entrance Lobby 1 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans Based on the availability of engineered controls 114 Purge Inlet Room 0 Yes routed to specify for both fire suppression agent run-off and smoke Radwaste ventilation paths as described in the LAR, the NRC staff concludes OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| | |
| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 327 Attachment E, Radioactive Release Tables Table 3.6-1, Compliance with NFPA 805 Radioactive Release Performance Criteria Effluent Engineered Controls Fire Area RCA or Pre-Fire Plan Rev NRC Staff Evaluation Fire Zone RCZ Suppression Combustion Water Smoke Processing for smoke that, the licensee's approach is acceptable to system for control, RIAs meet the NFPA 805 Radioactive Release monitoring and monitor Performance Criteria.
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| processing contamination prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 115 Purge Exhaust Room 0 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 116 AHU and SFP Change Rooms 0 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 117 Unit 2 Purge Inlet Room 0 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans Based on the availability of engineered controls routed to specify for both fire suppression agent run-off and smoke Radwaste ventilation paths as described in the LAR, the NRC staff concludes 118 Unit 2 Purge Exhaust Room 0 Yes Processing for smoke that, the licensee's approach is acceptable to system for control, RIAs meet the NFPA 805 Radioactive Release monitoring and monitor Performance Criteria.
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| processino contamination OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 328 Attachment E, Radioactive Release Tables Table 3.6-1, Compliance with NFPA 805 Radioactive Release Performance Criteria Effluent Engineered Controls Fire Area RCA or Pre-Fire Plan Rev NRC Staff Evaluation Fire Zone RCZ Suppression Combustion Water Smoke prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 119 Units 1 & 2 Air Handling Room 1 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 120 Unit 1 Purge Inlet Room 0 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes 121 Unit 1 Purge Inlet Room 0 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke Reactor Buildinqs Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes RB1 122 Unit 1 Reactor Bldg 0 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| prior to release levels of smoke RB2123 Unit 2 Reactor Buildinq 0 Yes Floor Drains Pre-fire Plans Based on the availability of engineered controls OFFICIAL USE ONLY SECURITY RELATED INFORM,A,TION
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| OFFICIAL USE ONLY SECURITY RELATED INFORM;\TION 329 Attachment E, Radioactive Release Tables Table 3.6-1, Compliance with NFPA 805 Radioactive Release Performance Criteria Effluent Engineered Controls Fire Area RCA or Pre-Fire Plan Rev NRC Staff Evaluation Fire Zone RCZ Suppression Combustion Water Smoke routed to specify for both fire suppression agent run-off and smoke Radwaste ventilation paths as described in the LAR, the NRC staff concludes Processing for smoke that, the licensee's approach is acceptable to system for control, RIAs meet the NFPA 805 Radioactive Release monitoring and monitor Performance Criteria.
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| processing contamination Drior to release levels of smoke Floor Drains Pre-fire Plans routed to specify Based on the availability of engineered controls Radwaste ventilation paths for both fire suppression agent run-off and smoke Processing for smoke as described in the LAR, the NRC staff concludes RB3124 Unit 3 Reactor Building 0 Yes system for control, RIAs that, the licensee's approach is acceptable to monitoring and monitor meet the NFPA 805 Radioactive Release processing contamination Performance Criteria.
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| Drior to release levels of smoke Essential Vacuum Siphon (ESV) The NRC staff finds the licensee's statement that ESV 1 No N/A N/A Buildinq the area has no radioloqical hazards acceptable The NRC staff finds the licensee's statement that KEO Keowee Hydro Station 1 No N/A N/A the area has no radioloqical hazards acceptable The NRC staff finds the licensee's statement that SSF Standby Shutdown Facility 1 No N/A N/A the area has no radioloqical hazards acceptable.
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| 230 KV Switchyard and The NRC staff finds the licensee's statement that SYD 1 No N/A N/A Relav House the area has no radioloqical hazards acceptable.
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| Area is regulated under a different NRC license Building Independent Spent Fuel Storage 1 Yes N/A N/A and is therefore not subject to the requirements 8027 Installation (ISFSI) Facility of 10 CFR 50.48(c).
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| Radiation Protection Drainage paths Based on the availability of engineered controls personnel go to Chemical for fire suppression agent run-off and fire brigade responding with Building Warehouse 5Z Treatment monitoring of smoke as described in the LAR, the 1 Yes Fire Brigade 8055 Old Warehouse 7 Pond #3 for NRC staff concludes that, the licensee's Monitor smoke monitoring and approach is acceptable to meet the NFPA 805 for release Radioactive Release Performance Criteria.
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| contamination OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 330 Attachment E, Radioactive Release Tables Table 3.6-1, Compliance with NFPA 805 Radioactive Release Performance Criteria Effluent Engineered Controls Fire Area RCA or Pre-Fire Plan Rev NRC Staff Evaluation Fire Zone RCZ Suppression Combustion Water Smoke Radiation Drainage paths Protection Based on the availability of engineered controls go to Chemical personnel for fire suppression agent run-off and fire brigade Building Treatment responding with monitoring of smoke as described in the LAR, the RCP Motor Refurbishment 1 Yes 8087 Pond #3 for Fire Brigade NRC staff concludes that, the licensee's monitoring and Monitor smoke approach is acceptable to meet the NFPA 805 release for Radioactive Release Performance Criteria.
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| contamination.
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| Radiation Drainage paths Protection Based on the availability of engineered controls go to Chemical personnel for fire suppression agent run-off and fire brigade Building Treatment responding with monitoring of smoke as described in the LAR, the Radwaste Facility 1 Yes 8089 Pond #3 for Fire Brigade NRC staff concludes that, the licensee's monitoring and Monitor smoke approach is acceptable to meet the NFPA 805 release for Radioactive Release Performance Criteria.
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| contamination.
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| Radiation Drainage paths Protection Based on the availability of engineered controls go to Chemical personnel for fire suppression agent run-off and fire brigade BUilding Treatment responding with monitoring of smoke as described in the LAR, the Scaffold Storage 1 Yes 8091 Pond #3 for Fire Brigade NRC staff concludes that, the licensee's monitoring and Monitor smoke approach is acceptable to meet the NFPA 805 release for Radioactive Release Performance Criteria.
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| contamination.
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| Radiation Protection personnel Drainage paths responding with Based on the availability of engineered controls go to Chemical Fire Brigade for fire suppression agent run-off and fire brigade Building Warehouse 3 Zone 2 Treatment Monitor smoke monitoring of smoke as described in the LAR, the 2 Yes 8093 Old Warehouse Pond #3 for for NRC staff concludes that, the licensee's monitoring and contamination. approach is acceptable to meet the NFPA 805 release Radioactive Release Performance Criteria.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION 331 Attachment E, Radioactive Release Tables Table 3.6-1, Compliance with NFPA 805 Radioactive Release Performance Criteria Effluent Engineered Controls Fire Area RCA or NRC Staff Evaluation Pre-Fire Plan Rev Fire Zone RCZ Suppression Combustion Water Smoke Radiation Drainage paths Protection Based on the availability of engineered controls go to Chemical personnel for fire suppression agent run-off and fire brigade Building Warehouse 3C Treatment responding with monitoring of smoke as described in the LAR, the 1 Yes 8096 Old Warehouse 6 Pond #3 for Fire Brigade NRC staff concludes that, the licensee's monitoring and Monitor smoke approach is acceptable to meet the NFPA 805 release for Radioactive Release Performance Criteria.
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| contamination.
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| N/A - Not Applicable Principal Contributors: Paul Lain, NRR/DRA Stephen Dinsmore, NRR/DRA Date: December 29, 2010 OFFICIAL USE ONLY SECURITY RELATED INFORMATION
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION T. Gillespie -2 Pursuant to Section 2.390 of Title 10 of the Code of Federal Regulations (10 CFR), by letter dated December 6, 2010, the NRC sent the licensee the draft Safety Evaluation approving the proposed amendments for an opportunity for the licensee to comment on any proprietary or security-related aspects of the draft Safety Evaluation. By letter dated December 22,2010, the licensee provided comments. The NRC reviewed and accepted all comments made by the licensee. Pursuant to 10 CFR 2.390 the NRC has redacted information as identified by blank space enclosed within double brackets as shown here [[ ]].
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| In addition, the December 6, 2010, letter also requested the licensee to provide comments on factual errors or clarity concerns contained in the draft Safety Evaluation. By letter dated December 22, 2010, the licensee provided comments. The NRC has considered each comment and changed the Safety Evaluation as appropriate.
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| A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
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| If you have any questions, please call me at 301-415-1345.
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| Sincerely, IRA!
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| John Stang, Senior Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-269, 50-270, and 50-287
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| ==Enclosures:==
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| : 1. Amendment No. 371 to DPR-38
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| : 2. Amendment No. 373 to DPR-47
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| : 3. Amendment No. 372 to DPR-55
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| : 4. Safety Evaluation contains offioial use only seourity related information cc: Distribution via Listserv DISTRIBUTION:
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| PUBLIC LPLlI-1 RtF RidsAcrsAcnw_MailCtr Resource RidsNrrDirsltsb Resource RidsNrrDorlDpr Resource RidsNrrDorlLpl2-1 Resource RidsNrrApla Resource RidsNrrAfpb Resource RidsNrrLAMOBrien Resource (hard copy)
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| RidsOgcRp Resource RidsRgn2MailCenter Resource RidsNrrPMOconee Resource (hard copy)
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| PLain, NRR SDinsmore, NRR ADAM 5 Accession No. ML1o3630612 *BS'i E-mal./ Date d OFFICE NRRlLPL2-1/PM NRRlLPL2-1/LA NRR/APLAlBC NRRlAFPB/BC OGC NRRlLPL2-1/BC NRRlLPL2-1/PM NAME JSlang MO'Brien DHarrison* AKlein* DRolh GKulesa JSlang DATE 12/29/10 12/29/10 12/29/10 12/28/10 12/23110 12/29/10 12/29/10 OFFICIAL RECORD COpy Enclosure 4 transmitted herewith contains security related infurmation. ','''hen separated from Enclosure 4, this document is decontrolled.
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| OFFICIAL USE ONLY SECURITY RELATED INFORMATION}}
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