ML15153A018

Prairie Island, Units 1 and 2 - License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactors - Response to Request for Additional Information

ML15153A018
Person / Time
Site:	Prairie Island
Issue date:	05/28/2015
From:	Davison K Northern States Power Co, Xcel Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-PI-15-041, TAC ME9734, TAC ME9735
Download: ML15153A018 (98)

Text

Prairie Island Nuclear Generating PlantU XcelEnergy 1717 Wakonade Drive EastWelch, MN 55089May 28, 2015 L-PI-1 5-04110 CFR 50.9010 CFR 50.48(c)U.S. Nuclear Regulatory Commission ATTN: Document Control DeskWashington, DC 20555-0001 Prairie Island Nuclear Generating Plant Units 1 and 2Dockets 50-282 and 50-306Renewed License Nos. DPR-42 and DPR-60License Amendment Request to Adopt NFPA 805 Performance-Based Standard for FireProtection for Light Water Reactors

-Response to Request for Additional Information (TAC Nos. ME9734 and ME9735)

References:

1. NSPM letter, J.P. Sorensen to NRC Document Control Desk, LicenseAmendment Request to Adopt NFPA 805 Performance-Based Standard forFire Protection for Light Water Reactors, L-PI-1 2-089, dated September 28,2012 (ADAMS Accession No. ML12278A405).

2. NSPM letter, S. Sharp to NRC Document Control Desk, Supplement toLicense Amendment Request to Adopt NFPA 805 Performance BasedStandard for Fire Protection for Light Water Reactors, L-PI-1 4-045, datedApril 30, 2014 (ADAMS Nos. ML14125A106 and ML14125A149).

3. NRC email, T. Beltz to S. Chesnutt, Prairie Island Nuclear Generating Plant,Units 1 and 2 -NFPA 805 Requests for Additional Information and ResponseTimeline (TAC Nos. ME9734 and ME9735),

dated March 30, 2015 (ADAMSAccession No. ML15089A157).

In Reference 1, the Northern States Power Company, a Minnesota Corporation (NSPM)doing business as Xcel Energy requested approval from the Nuclear Regulatory Commission (NRC) to transition the fire protection licensing basis for the Prairie IslandNuclear Generating Plant (PINGP) to 10 CFR 50.48(c),

National Fire Protection Association Standard 805 (NFPA 805). Supplemental information was provided inletters dated November 8, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12314A144) and December 18, 2012 (ADAMSAccession No. ML12354A464).

Aooý Document Control DeskPage 2In Reference 2, NSPM submitted a revised Fire Probabilistic Risk Assessment (PRA) ina supplement to the subject License Amendment Request (LAR). In Reference 3, theNRC staff provided requests for additional information (RAIs) regarding this request andalso provided a timeline and due dates for submitting responses within 60, 90, or 120days after an on-site Audit that was conducted March 23-25, 2015.Enclosure 1 provides NSPM's responses to all of the 60-day RAIs (due by May 29,2015) and one 90-day RAI included in Reference 3.Enclosure 2 provides Licensee Identified Changes to the LAR (Reference

2) that are notdirectly associated with RAI responses.

This enclosure includes clarifications andchanges to LAR Attachment S, Modifications and Implementation Items, that are basedon NSPM reviews subsequent to the submittal of Reference

2. These changes will beincluded in a revision to Attachment S to be submitted with the final RAI response letter.Enclosure 3 provides a revision to Approval Request 1 in LAR Attachment L, "NFPA805 Chapter 3 Requirements for Approval,"

in support of the response to FPE RAI 05.This revised Approval Request 1 supersedes the Approval Request 1 submitted inReference 2.Enclosure 4 provides a revision to Request 1 in LAR Attachment T, "Clarifications ofPrior NRC Approvals,"

in support of the response to SSA RAI 02. This revisedRequest 1 supersedes the Request 1 submitted in Reference 2.This letter is submitted in accordance with 10 CFR 50.90. The additional information provided in this letter does not impact the conclusions of the No Significant HazardsEvaluation or Environmental Considerations Evaluation presented in Reference 2.In accordance with 10 CFR 50.91, NSPM is notifying the State of Minnesota of thisadditional information by transmitting a copy of this letter to the designated StateOfficial.

If there are any questions or if additional information is needed, please contact GeneEckholt at 651-267-1742.

Summary of Commitments This letter contains no new commitments and makes no revisions to any existingcommitments.

Document Control DeskPage 3I declare under penalty of perjury that the foregoing is true and correct.Executed on MAY 2 8 2015Kevin DavisonSite Vice President, Prairie Island Nuclear Generating PlantNorthern States Power Company -Minnesota Enclosures (4)cc: Administrator, Region III, USNRCNRR Project Manager, PINGP, USNRCResident Inspector, PINGP, USNRCState of Minnesota L-PI-15-041 NSPMEnclosure 1Response to Requests for Additional Information (RAIs)Regardinq the License Amendment Request toAdopt National Fire Protection Association (NFPA) Standard 805at Prairie Island Nuclear Generating Plant Units 1 and 2Responses to 60-Day RAIsIntroduction This Enclosure provides additional information from the Northern States PowerCompany, a Minnesota corporation (hereafter "NSPM") doing business as Xcel Energy,in support of a License Amendment Request (LAR) to transition the Fire Protection Licensing Basis for the Prairie Island Nuclear Generating Plant (PINGP) to National FireProtection Association Standard 805 (NFPA 805).NSPM submitted an NFPA 805 LAR for PINGP in a letter dated September 28, 2012(ADAMS Accession No. ML12278A405).

Supplemental information was submitted inletters dated November 8, and December 18, 2012 (ADAMS Accession Nos.ML12314A144 and ML12354A464, respectfully).

In a letter dated April 30, 2014, NSPMsubmitted a revised Fire Probabilistic Risk Assessment (PRA) in a supplement to thesubject LAR (ADAMS Accession Nos. ML14125A106 and ML14125A149).

This 2014Supplement included the entire LAR and was designated "Revision 1;" this is theversion referred to in discussions involving "the LAR."After an onsite audit conducted during the week of March 23, 2015, the NRC staffprovided requests for additional information (RAIs) and a response timeline in an emaildated March 30, 2015 (ADAMS Accession No. ML1 5089A1 57). The timeline provideddue dates for 60, 90, and 120 day RAI responses.

This Enclosure provides NSPM's responses to all of the 60-day RAIs which are due byMay 29, 2015, and one 90-day RAI, as follows:* Fire Protection Engineering (FPE) RAI 01, 02, 03*, 04, 05, 06(*note that 03 was originally a 90-day RAI)* Safe Shutdown Analysis (SSA) RAI 01, 02, 03, 04, 05, 06, 08, 09, 10* Fire Modeling (FM) RAI 04, 05, 06* Radioactive Release (RH) RAI 01, 02, 03" Probabilistic Risk Assessment (PRA) RAI 01.a, 01.b, 01.c, 01 .d, 01 .f, 02.b, 02.c,04, 05, 06, 09, 10, 11, 13, 14, 15, 19Each of the RAI questions is quoted in italics and each question is then followed by theNSPM response in normal font. Referenced documents are identified at the end ofeach RAI response.

Page 1 of 81 L-PI-1 5-041 NSPMEnclosure 1Acronyms and Abbreviations ADAMS Agencywide Document Access and Management SystemADT Auxiliary Drain TankAFW Auxiliary Feedwater AFWP Auxiliary Feedwater PumpAHJ Agency Having Jurisdiction ANS American Nuclear SocietyAOP Abnormal Operating Procedure APEO Auxiliary Plant Equipment OperatorAR Action RequestARP Alarm Response Procedure ASD Alternate ShutdownASME American Society of Mechanical Engineers BA Boric AcidBAST Boric Acid Storage TankBTP Branch Technical PositionCAAB Common Area of the Auxiliary BuildingCCDP Conditional Core Damage Probability CDF Core Damage Frequency CE Combustion Engineering CFAST Consolidated Fire and Smoke Transport CFR Code of Federal Regulations CLERP Conditional Large Early Release Probability CLG WTR Cooling WaterC02 Carbon DioxideCR Control RoomCST Condensate Storage TankCVCS Chemical and Volume Control SystemDDFP Diesel-Driven Fire PumpDID Defense in DepthDAW Dry Active WasteDC Direct CurrentEC Engineering ChangeEOP Emergency Operating Procedure EPRI Electrical Power Research Institute EPA Environmental Protection AgencyERFBS Electrical Raceway Fire Barrier SystemES Equipment Selection FA Fire AreaFAQ Frequently Asked QuestionFDS Fire Dynamics Simulator FDT Fire Dynamics ToolFM Fire ModelingF&O Fact & Observation FPA Foote, Pagni, and AlvaresFPE Fire Protection Engineering (or Engineer)

FPEE Fire Protection Engineering Evaluation FPRA Fire Probabilistic Risk Assessment Page 2 of 81 L-PI-15-041 Enclosure 1FQFRANXFREHEAFHEPHEPAHFEHRAHREHRRHSDHUTIFGIGNKSFLARLERFLLRWMAAPMCRMDFPMQHN/ANEINFPANLONPONPSHNRCNSCANSPNSPNSPMNSSSNUREGNUREG/CROCAODCMPAPAUPCsP&IDPIPINGPPNSPORVPRAPSAPWRRARAIRCANSPMFire Quantification Fire Risk Analysis Software ToolFire Risk Evaluation High Energy Arcing FaultHuman Error Probability High Efficiency Particulate AirHuman Failure EventHuman Reliability AnalysisHigher Risk Evolution Heat Release RateHot ShutdownHoldup TankInstrument Failure GuideIgnitionKey Safety FunctionLicense Amendment RequestLarge Early Release Frequency Low Level Radioactive WasteModular Accident Analysis ProgramMain Control RoomMotor-Driven Fire PumpMcCaffrey, Quintiere, and Harkleroad Not Applicable Nuclear Energy Institute National Fire Protection Association Non-Licensed OperatorNon-Power Operations Net Positive Suction HeadNuclear Regulatory Commission Nuclear Safety Capability Assessment Northern States PowerNon-Suppression Probability Northern States Power -Minnesota Nuclear Steam Supply SystemNRC Publication NUREG/Contractor ReportOwner Controlled AreaOffsite Dose Calculation ManualPlant Attendant Physical Analysis UnitPrimary Control StationPiping and Instrumentation DiagramPrairie IslandPrairie Island Nuclear Generating PlantProbability of Non-Suppression Power Operated Relief ValveProbabilistic Risk Assessment Probabilistic Safety Assessment Pressurized Water ReactorRecovery ActionRequest for Additional Information Radiologically Controlled AreaPage 3 of 81 L-PI-15-041 NSPMEnclosure 1RCP Reactor Coolant PumpRCS Reactor Coolant SystemRG Regulatory GuideRHR Residual Heat RemovalRMU Reactor MakeupRO Reactor OperatorRR Radioactive ReleaseRWST Refueling Water Storage TankSAFE GENESIS Database used to develop PINGP analytical modelSCA Single Compartment AnalysisSCBA Self Contained Breathing Apparatus SER Safety Evaluation ReportSF Severity FactorSFP Spent Fuel PoolSFPE Society of Fire Protection Engineers SR Supporting Requirement SSA Safe Shutdown AnalysisSSC Structures,

Systems, and Components TBD Technical Basis DocumentTDAFWP Turbine Driven Auxiliary Feedwater PumpV&V Verification and Validation VCT Volume Control TankVDC Volts Direct CurrentVEWFDS Very Early Warning Fire Detection SystemVFDR Variance From Deterministic Requirements WCAP Westinghouse Commercial Atomic PowerX/Q Chi/Q -Relative Concentration ZOI Zone of Influence Page 4 of 81 L-PI-15-041 NSPMEnclosure 1RAI Responses

-Fire Protection Engineering (FPE)FPE RA101NFPA 805, Section 3.4. 1 (a), requires that a fully-staffed and qualified fire brigade comply withNFPA standards NFPA 600, NFPA 1500, and NFPA 1582, as applicable.

In the Enclosure toyour August 20, 2014, license amendment request (hereafter referred to as the LAR),Attachment A, Section 3.4. 1 (a), it states that the compliance strategy as "Complies via PreviousApproval."

However, the compliance basis does not provide excerpts from the past licenseesubmittal(s) and associated U.S. Nuclear Regulatory Commission (NRC) documentation thatsubstantiates the previous approval.

The NRC endorsed guidance in NEI 04-02, Appendix B,Section B- 1, states that "When claiming previous

approval, excerpts from the NRC documents that provided the formal approval shall be included in documentation, as well as appropriate excerpts from licensee's submittals."

Additionally, Section B. 1 states "for each reference document that is referenced as part of the transition review, provide sufficient documentation toprovide traceability back to the original submittal and approval.

Please provide, as appropriate, information such as revision number, date, and section/page number in order to make the statements as clear as possible to facilitate reviews and long termconfiguration management."

Additionally,

[This RAI includes Subparts a and b, as shown below along with NSPM responses]

NRC Request (FPE RAI 01.a):a. The Technical Specification discussion in the Compliance Basis only addresses minimum staffing and does not address the fire brigade capabilities as addressed in theNFPA standards.

Please provide additional details to substantiate the compliance strategy that the NRChas previously approved the fire brigade for PINGP.NSPM Response (FPE RAI 01.a):a. Correspondence between Northern States Power (NSP) and the NRC that explains theNRC's previous approval of the PINGP fire brigade includes the following:

NSP letter dated January 9, 1979 (Reference F-1): NSP submitted information regarding fire brigade training that included the following excerpts:

PF-44 Fire Brigade Practice SessionsStaff Position:

Practice sessions should be held for fire brigade members on theproper method of fighting various types of fires that could occur in a nuclear powerplant considering such factors as the magnitude of the fire, and the complexity anddifficulty of fire fighting.

These sessions should be designed to provide brigademembers with experience in actual fire extinguishment and the use of emergency breathing apparatus under strenuous working conditions.

The sessions should be inaddition to the scheduled fire brigade training sessions and fire drills and shouldPage 5 of 81 L-PI-15-041 NSPMEnclosure 1include firefighting strategies, (i.e., simple plans showing fire fighting equipment locations, entry and egress points, ventilation, communications and emergency lighting locations and controls).

These practice sessions should be provided atregular intervals, but not exceeding a one year interval for each fire brigade member.Licensee's ResponseThe licensee will respond to this position by 3-2-79. [The response is provided in anexcerpt from the March 9, 1979 letter below (Reference F-2).]PF-6 Fire Brigade Trainingd. Fire Drills should be performed at regular intervals but not to exceed threemonths for each brigade.Licensee's Responsed. With our six (6) fire brigades, this would require a minimum of 24 drills a year.We consider this excessive and are concerned it could lead to perfunctory performance on the part of those participating in the drills as well as anunwarranted state of confusion for others at the facility.

We believe one drill permonth (2 drills per brigade) is the maximum.NSP letter dated March 9, 1979 (Reference F-2): NSP submitted additional information regarding fire brigade training that included the following excerpt:PF-44 Fire Brigade Practice Sessions[See Staff Position in the January 9, 1979 letter above (Reference F-i)]Licensees ResponseWe agree to schedule fire brigade practice sessions as described above at leastannually.

We cannot provide assurance,

however, that 100% of the fire brigademembers will be available to attend each scheduled practice session.

Due tovacation,

sickness, offsite training, and unexpected schedule

changes, some firebrigade members may miss a session.

Because of the amount of preparation andplanning that goes into a practice

session, it is impractical to schedule makeupsessions.

We therefore propose to require, at least 85% of all fire brigade membersto attend each practice session.NSP letter dated May 2, 1979 (Reference F-3): NSP submitted additional information regarding fire brigade staffing that included the following excerpt:J. Nuclear Plant Fire Brigades and Fire Brigade Support TeamsA Fire Brigade of five (5) persons will be on-site at all times. In addition, a FireBrigade Support Team will be on-site at all times to bring the minimum number ofpersons responding to any fire to five. The Fire Brigade Support Team may bedrawn from the site security force. The Support Team assists the Fire Brigade byproviding communications, bringing equipment to the scene, renewing airbreathing

bottles, and providing other support.Page 6 of 81 L-PI-15-041 NSPMEnclosure 1Each Fire Brigade will have an appointed leader. This leader will not be the ShiftSupervisor (the Unit No. 1 Shift Supervisor at Prairie Island).NRC SER dated September

6. 1979 (Reference F-4): The NRC's previous approval ofthe fire brigade for PINGP is provided in a Safety Evaluation Report (SER) datedSeptember 6, 1979. This SER was included in a letter from A. Schwencer (NRC) to L.O.Mayer (NSP) identified as Reference 6.28 in the LAR (Reference F-4).Sections of the SER that apply to the Fire Brigade include the following excerpts:

Section 6.1, Fire Protection Organization, page 6-1:The licensee's fire protection organization contains the organizational responsibilities and lines of communication between the various positions through the use oforganizational charts and functional descriptions of each position's responsibilities.

Upper level offsite management positions are designated which have management responsibility for the formulation, implementation, and assessment of theeffectiveness of the nuclear plant fire protection program.

The results of theseassessments are reported to the upper level management position responsible forfire protection with recommendations for improvements or corrective actions asdeemed necessary.

The organizational responsibilities are delineated through onsite management positions for design, installation,

testing, maintenance, modification, and review offire protection systems and for fire brigade training.

A fire brigade size has been agreed upon. Qualification requirements have beenestablished for fire brigade members, and the position responsible for formulating and implementing the fire protection program.

Satisfactory completion of a physicalexamination including periodic screening for performing strenuous activity is requiredfor all fire brigade members.We find that the fire protection organization conforms to the provisions of Appendix Ato BTP 9.5-1 and is, therefore, acceptable.

Section 6.2, Fire Brigade Training, page 6-2:The fire brigade training program consists of an initial classroom instruction programfollowed by periodic classroom instruction, practice in firefighting and fire drills.Practice sessions are held for fire brigade members on the proper method of fightingvarious types of fires that could occur in a nuclear plant considering such factors asthe magnitude of the fire and the complexity and difficulty of firefighting.

Records oftraining are provided and are available for review including drill critiques.

We find that the fire brigade training conforms to the provisions of Appendix A to BTP9.5-1 and is, therefore, acceptable.

Section 6.5, Firefighting Procedures, page 6-2The licensee has provided an adequate description of its current firefighting procedures, those under development and those planned to be developed in thePage 7 of 81 L-PI-15-041 NSPMEnclosure 1near future. Firefighting procedures/plans are established to cover such items asnotification of a fire, fire emergency procedures, coordination of firefighting activities with offsite fire departments, strategies for fighting fires in all safety-related areas andareas presenting a hazard to safety-related equipment.

Provisions have been madefor including offsite firefighting organizations in fire brigade drills and training asrequired.

We find that the control of firefighting procedures conforms to the provisions ofAppendix A to BTP 9.5-1 and is, therefore, acceptable.

NRC letter dated September 6,1979 (Reference F-4): In addition, the 1979 NRC letter(Reference F-4) included the following statements regarding resolutions of fire brigadeissues.4. Technical Specification 6.1.C.6 requires a fire brigade of three members ratherthan the five members that the staff had requested.

We have completed ourreview of this matter and have concluded that a five man brigade is necessary.

Your staff has agreed to this change and the Technical Specifications have beenchanged in these amendments to read "five" rather than "three" fire brigademembers.5. Figure 6.1-1 does not have the specific delineation of fire protection responsibility and Figure 6.1-2 does not have specific delineation of operator and fireprotection training responsibilities under the Training Supervisor as requested bythe staff. We have completed our review and find that Figure 6.1-1 and Figure6.1-2 should be modified as requested.

Your staff has agreed to these changesand they have been incorporated in these amendments.

The SER concluded that the PINGP Fire Brigade capabilities conform to theprovisions of Appendix A to BTP 9.5.1 and are therefore acceptable.

NRC Request (FPE RAI 01.b):b. The licensee has stated that the fire brigade has been reviewed against NFPA 600;however there is no statement of compliance with regard to the requirements of NFPA600.Please provide a more detailed description of compliance with NFPA 600.NSPM Response (FPE RAI 01.b):b. NSPM performed a detailed review of NFPA 600, Standard on Industrial Fire Brigades, as documented in Fire Protection Engineering Evaluation FPEE-1 1-031, NFPA 600 -2000 Code Compliance Report, Revision 1, and did not identify any deviations from theCode.Based on this review, the PINGP Fire Brigade complies with NFPA 600, 2000 edition,and therefore, Section 3.4.1 (a) of NFPA 805.Page 8 of 81 L-PI-15-041 NSPMEnclosure 1References F-1 NSP letter, L.O. Mayer to Director of Nuclear Reactor Regulation (NRC), Corrected Pages for NSP Initial Response to NRC Staff Evaluation of Fire Protection Program,dated January 9, 1979.F-2 NSP letter, L.O. Mayer to Director of Nuclear Reactor Regulation (NRC), NRC StaffEvaluation of Fire Protection

Program, dated March 9, 1979.F-3 NSP letter, L.O. Mayer to Director of Nuclear Reactor Regulation (NRC), NuclearPlant Fire Protection Functional Responsibilities, Administrative

Controls, and QualityAssurance, dated May 2, 1979.F-4 Letter from A. Schwencer (NRC) to L.O. Mayer (NSP), Issuance of LicenseAmendment Nos. 39 (DPR-42, Unit 1) and 33 (DPR-60, Unit 2), and Related FireProtection Safety Evaluation Report, dated September 6, 1979.FPE RA102NFPA 805, Section 3.4.1(c),

requires that the fire brigade leader and at least two brigademembers have sufficient training and knowledge of nuclear safety systems to understand theeffects of fire and fire suppressants on nuclear safety performance criteria.

As described in the Compliance Basis for Section 3.4. 1(c) in Attachment A of the LAR, NSPMhas incorporated the NFPA 805 requirement in its procedures;

however, it does not summarize the training and knowledge level required of the non-licensed operators that are assigned to thefire brigade.Please provide additional detail regarding the training that is provided to the fire brigademembers that addresses their ability to assess the effects of fire and fire suppressants onnuclear safety performance criteria.

NSPM Response (FPE RAI 02):NFPA 805, Chapter 3, Section 3.4.1 (c) requires certain fire brigade members to have sufficient training and knowledge of nuclear safety systems to understand the effects of fire and firesuppressants on nuclear safety performance criteria.

The guidance in FAQ 13-0069, Revision 3, states that one acceptable approach to meeting therequirements of NFPA 805 Section 3.4.1 (c) is to provide a Fire Brigade leader that is a licensedreactor operator or has equivalent knowledge of plant systems and at least two brigademembers that have training at the level of Non-Licensed Operator (NLO) or have equivalent knowledge of plant systems.All fire brigade members are trained NLOs. The NLOs are split into two classifications based ontheir level of experience.

The Plant Attendant (PA) reports to the plant and performs the firebrigade and operator/

watchstander duty after receiving initial training.

The Auxiliary PlantEquipment Operator (APEO) is more experienced than the PA because, in addition to the initialPage 9 of 81 L-PI-15-041 NSPMEnclosure 1training, they have also completed their apprenticeship and journeyperson and fire brigadeleader training.

The fire brigade may be made up of both PAs and APEOs but only APEOsperform the duty of the Fire Brigade Leader. All fire brigade members receive training on firecontrol.All NLOs receive training on all systems in their initial qualifications prior to reporting to the plantas a PA. NLOs rotate between watchstanding positions allowing them to gain experience in allof the systems.NLOs are evaluated on each system lesson plan objective during initial and continuing trainingper Systematic Approach to Training processes.

A review of the standard learning objectives insystem training materials was conducted.

The following learning objectives apply to NLOs andReactor Operators (RO)s:* Describe the System Flow Path and Major Components for:o Normal Operations o Abnormal Operations o Emergency Operations

" System integration with respect to other associated systems.o Precautions and limitations pertinent to the system, technical specifications, andUpdated Safety Analysis Report design criteria.

All NLOs and ROs receive training on all plant systems including system integration with respectto other associated

systems, precautions and limitations pertinent to the systems andequipment required for abnormal and emergency operations, providing the members of the firebrigade with sufficient training and knowledge of nuclear safety systems to understand theeffects of fire and fire suppressants on nuclear safety performance criteria.

In addition to thetraining regarding nuclear safety systems, the Fire Brigade Leader has additional experience with all of the systems during their apprenticeship at the various watchstanding positions.

ThePINGP training program for NLOs is consistent with the guidance in FAQ 13-0069, Revision 3,for equivalent knowledge of plant systems."

Based on the above, Fire Brigade training complies with NFPA 805 section 3.4.1 (c).FPE RA103NFPA 805, Section 3. 10.1, requires that if an automatic total flooding and local application gaseous fire suppression system is required to meet the performance or deterministic requirements of Chapter 4, then the system shall be designed and installed in accordance withthe applicable NFPA codes. Attachment A, Section 3.10.1, of the LAR identifies animplementation item for code deviations that require a modification to resolve non-compliances associated with unprotected beam pockets and system supervision, and states that thismodification is identified in Table S-2.Please identify the specific plant modification item to which this applies.NSPM Response (FPE RAI 03):The plant modification described in LAR Attachment A, Section 3.10.1, Gaseous Suppression

Systems, is the Proposed Modification for FA1 8 described in Item # 8 in Table S-2. ThisPage 10 of 81 L-PI-15-041 NSPMEnclosure 1modification will resolve the non-compliance with NFPA 72E (Automatic Fire Detectors) associated with unprotected beam pockets and system supervision.

This modification alsomodifies the Ionization Fire Detection system to provide two zones of coverage in the RelayRoom and P250 Computer Room, and also modifies the C02 fire suppression system toactuate if both Ionization zones detect a fire in lieu of heat detectors.

The NFPA code issue listed in Attachment A, Section 3.10.1 under "Items for Implementation" was identified in the NFPA 72E code compliance evaluation, which is not currently referenced inSection 3.10.1 under "Plant Documentation."

Section 3.10.1 should be revised toreflect evaluation FPEE-1 1-048, Revision 0, Code Compliance Review, NFPA 72E Automatic Fire Detectors, 1974, 1982, 1987, Detector Spacing and Location for Plant Areas/Systems NotAddressed in FPP-5, Revision 2.In addition, the "Compliance Statement" column should also include the compliance basis"Complies with Use of Existing Engineering Equivalency Evaluation."

The "Compliance Basis"should be revised to include applicable sections from FPEE-1 1-048 to demonstrate compliance with the requirements section 3.10.1.FPE RA104NFPA 805, Section 2.4.3.3, states that the use of the Fire Risk Evaluation performance-based approach requires that "The PSA [probabilistic safety assessment]

approach, methods and datashall be acceptable to the AHJ [authority having jurisdiction, which is the NRC]."Attachment S, Table S-2, Plant Modification Item 5, identifies the installation of a Very EarlyWarning Fire Detection System (VEWFDS) to monitor the conditions inside low voltage cabinetslocated in fire compartment FA 18 to reduce the likelihood of fire propagation outside thecabinets.

Please provide a more detailed description of the proposed modification including:

[This RAI includes Subparts a through g, as shown below along with NSPM responses]

NRC Request (FPE RAI 04.a):a. The NFPA code(s) of record, proposed installation configuration (inside cabinets or area-wide, common piping or individual cabinet piping),

and the use of equipment manufacturers recommendations regarding design, installation, and piping.NSPM Response (FPE RAI 04.a):a. The proposed modification identified in Table S-2 Item 5 is to install an air aspirating incipient detection system inside low voltage (less than or equal to 250V) cabinets in theRelay Room, Fire Area 18. This modification is still in the design phase, however, thefollowing is being planned at this time. Common piping will be used for groupings ofcabinets.

The system will be designed to meet the sensitivity criteria in NFPA 76. Thesystem will alarm at the existing fire alarm control panel in the Control Room and will bedesigned, installed, and maintained in accordance with NFPA 72 and manufacturer's recommendations.

The applicable codes at the time of design will be used. ThePage 11 of 81 L-PI-1 5-041 NSPMEnclosure 1guidance in FAQ 08-0046 Closure Memo dated 11/23/2009 (ML093220426),

along withNRC White Paper: Very Early Warning Detection Systems Rev. 1, dated 7/22/2014, willbe followed for system design.NRC Request (FPE RAI 04.b):b. The acceptance

testing, sensitivity and setpoint control(s),

alarm response procedures and training, and routine inspection,

testing, and maintenance that will be implemented to credit the new incipient detection system.NSPM Response (FPE RAI 04.b):b. Factory and Site Acceptance testing will address transport time (maximum of 60seconds from most remote sampling port), and sensitivity thresholds for alert (0.2% perfoot obscuration) and alarm (1.0% per foot obscuration) at each sampling port. AlarmResponse Procedures and associated training will be revised and/or developed todescribe actions to be taken in response to alert and alarm signals.

Inspection, testing,and maintenance procedures will be developed in accordance with equipment manufacturer's recommendations and NFPA 72. Training will be developed with theassistance of manufacturer's recommendations for plant operations, inspection, testingand maintenance groups on the air aspirating incipient detection system.NRC Request (FPE RAI 04.c):c. The configuration and design control process that will control and maintain the setpoints for both alert and alarm functions from the VEWFDS.NSPM Response (FPE RAI 04.c):c. Configuration and design control process that will control and maintain the setpoints forboth alert and alarm functions from the VEWFDS will be in accordance with the sitemodification process and procedures.

NRC Request (FPE RAI 04.d):d. The instructions that will be required of the first responders until the degrading component is repaired, the cabinet is de-energized, or the alarm is satisfactorily reset.NSPM Response (FPE RAI 04.d):d. Alarm Response Procedures for fire alarms and associated training will be revisedand/or developed to describe actions to be taken in response to alert and alarm signals.These actions are still being developed and will be identified as part of the modification process.Page 12 of 81 L-PI-15-041 NSPMEnclosure 1NRC Request (FPE RAI 04.e):e. The credit taken in the Fire PRA [probabilistic risk assessment]

for the VEWFDS inassessing the risk in the fire areas where the system is credited.

NSPM Response (FPE RAI 04.e):e. Credit for incipient fire detection system in the Fire PRA is modeled as a multiplicative factor to the fire ignition frequency.

The credited multiplication factor is 0.02 and isderived using the methods in NUREG/CR-6850 Supplement 1.NRC Request (FPE RAI 04.f):f. Whether this installation and the credit that will be taken will be in accordance with eachof the method elements, limitations and criteria of NUREG/CR-6850, Supplement 1,"Fire Probabilistic Risk Assessment Methods Enhancements,"

Chapter 13, andFrequently Asked Question (FAQ) 08-0046, including the closeout memorandum.

NSPM Response (FPE RAI 04.f):f. Credit was taken in accordance with limitations and criteria of NUREG/CR-6850, Supplement 1, "Fire Probabilistic Risk Assessment Methods Enhancements,"

Chapter13, and Frequently Asked Question (FAQ) 08-0046, including the closeoutmemorandum.

NRC Request (FPE RAI 04.g):g. Justification for any deviations from NUREG/CR-6850, Supplement 1, "Fire Probabilistic Risk Assessment Methods Enhancements,"

Chapter 13, and FAQ 08-0046, including thecloseout memorandum.

NSPM Response (FPE RAI 04.g):g. No deviations from NUREG/CR-6850, Supplement 1, exist within the credit for modelingof the VEWFDS.FPE RAt105NFPA 805, Section 3.5.16, requires "The fire protection water supply system shall be dedicated for fire protection use only." In Attachment L of the LAR, Approval Request 1, the licenseeidentifies the fire protection water supply system at PINGP may periodically be utilized to supplywater for non-fire protection purposes.

The LAR states the fire water system can be aligned forscreen wash system use and "emergency uses, "and as such it does not meet the requirement or allowed exceptions.

Page 13 of 81 L-PI-15-041 NSPMEnclosure 1a. The Approval Request only addresses use of fire water for the screen wash in the "Basis,for the Request, "and does not address the bases for aligning the fire water for "otheremergency uses."Please provide a description of "other emergency uses" that may be in addition to screenwash purposes.

For all non-fire use uses, in accordance with 10 [CFR] 50.48(c)(2)(vii),

include adiscussion on potential impacts to the nuclear safety and radiological releaseperformance criteria.

In addition, describe how safety margin is maintained and discusshow the three elements of defense-in-depth are met.NSPM Response (FPE RAI 05):a. As described in the RAI, Attachment L to the LAR includes a request for approval to usefire protection water for non-fire protection

purposes, including screenwash and "otheremergency uses." NSPM would like to clarify that "other emergency uses" should bereplaced by "spent fuel pool makeup."

Attachment L has been revised to reflect thischange and is included in Enclosure 3 to this letter. The revised Attachment Laddresses potential impacts to the nuclear safety and radiological release performance

criteria, and also includes a discussion regarding safety margin and the three elements(echelons) of defense-in-depth.

FPE RA106NFPA 805, Section 3.6. 1, Standpipe and Hose Stations, requires that "For all power blockbuildings, Class Ill Standpipe and hose stations shall be installed in accordance with NFPA 14,Standard for the Installation of Standpipe, Private Hydrants and Hose Systems."

In Attachment A of the LAR, Section 3.6. 1, Standpipe and Hose Stations, lists NFPA 14 codecompliance evaluations.

The summary statement in the LAR for the code compliance evaluations identifies the need to resolve one unacceptable deviation (NFPA 14-1986) and twooutstanding Action Requests (NFPA 14-1969) resulting from the NFPA 14 reviews.

These openitems listed in the LAR do not appear to have implementation items identified in Attachment S.Please provide more detail with regard to the deviation and two outstanding items, including theidentification of implementation items required for these items, as appropriate or necessary tomeet NFPA 805. If implementation items for these code deviations are deemed unnecessary tomeet NFPA 805, provide additional justification.

NSPM Response (FPE RAI 06):The deviation from NFPA 14-1986 identified in Attachment A of the LAR is that the pipe sizes inthe D5/D6 Standpipe and Hose Station System do not meet the minimum size requirements inTable 2-1.1, based on total pipe length. As described in FPEE 11-050, this deviation will beresolved by performance of a hydraulic calculation.

Performance of this hydraulic calculation isbeing tracked by AR 01470080-02, and its completion will be verified through a newImplementation Item in Table S-3, item #64, as follows:Page 14 of 81 L-PI-15-041 NSPMEnclosure 1Item 64:Update code compliance reviews to document resolution of identified open items.The two outstanding items for NFPA 14-1969 identified in Attachment A of the LAR are asfollows:1) Section 5.10, Hangers.

An additional hanger was determined necessary to preventexcessive movement of a standpipe, as described in FPEE NFPA 14-1969.

This hangeris identified in Engineering Change (EC) 18011, and the installation will be completed aspart of a new Proposed Modification

1. 40 that will be added to Table S-2 of the LAR.2) Section 5.11, Gauges. Fire Protection standpipes in the older portions of the PINGPplant were not installed with pressure gauges at the top of the hose station risers. Thisincludes the Screenhouse, Turbine Building, Auxiliary

Building, and Old Administration Building.

The NFPA 14-1969 code compliance review will be updated to include theresolution of this open item as part of the proposed new Implementation Item #64 inTable S-3, "Update code compliance reviews to document resolution of identified openitems."Page 15 of 81 L-PI-15-041 NSPMEnclosure 1RAI Responses

-Safe Shutdown Analysis (SSA)SSA RA101NFPA 805, Section 2.4.2, Nuclear Safety Capability Assessment, requires licensees to performa nuclear safety capability assessment (NSCA). Regulatory Guide (RG) 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light- Water Nuclear Power Plants, "endorsed the guidance in NEI 00-01, "Guidance for Post-Fire Safe Shutdown Circuit Analysis,"

Chapter 3,as one acceptable approach to perform an NSCA. In Attachment B of the LAR, the alignment basis for NEI 00-0 1, Attribute 3.1.1.4, is "Not in Alignment

[but Prior NRC Approval]."

Thelicensee stated that the Fire PRA employs the use of a hot shutdown (HSD) panel in conjunction with other controls when required to abandon the Control Room (CR) and that the HSD panelsand controls do not meet the definition of a primary control station (PCS) as defined by RG1.205. Additionally, the licensee stated that actions to enable the HSD panel and establish required controls are listed in the PRA Alternative Shutdown Notebook.

[This RAI includes Subparts a through e, as shown below along with NSPM responses]

NRC Request (SSA RAI 01.a):a. RG 1.205 Section C.2.4 describes two cases where operator actions taken outside themain control room may be considered as taking place in the PCS: (a) controls for asystem or component specifically installed to meet the "dedicated shutdown" option inSection IIl.G. 3 of Appendix R and (b) controls for some systems and components thathave been modified to meet the "alternative shutdown" option in Section 1lL G.3.Please provide a discussion of the HSD panel and what attributes of PINGP design andprocedures do not meet the definition of PCS as defined in RG 1.205.NSPM Response (SSA RAI 01.a):a. The PINGP Hot Shutdown Panels were determined to not meet the definition of aPrimary Control Station using the criteria in Regulatory Guide 1.205 and FAQ 07-0030.Regulatory Guide 1.205, Section 2.4 Recovery Actions states: "The staff has identified two cases where operator actions taken outside the main control room may beconsidered as taking place at a primary control station.

These two cases involvededicated shutdown or alternative shutdown

controls, which have been reviewed andapproved by the NRC." FAQ 07-0030 describes that the Primary Control Station shouldhave the requisite system and component

controls, plant parameter indications andcommunications so that the operator can adequately and safely monitor and control theplant using the alternative shutdown equipment.

The PINGP Hot Shutdown Panels have not been reviewed and approved by the NRC.These panels were installed in 1972 before the plant began operation and they weredesigned for General Design Criteria 19, Control Room. Generic Letter 81-12 states thatany modifications that the licensee plans in order to meet the requirements of SectionIII.G.3 of Appendix R must be reviewed and approved by the NRC. PINGP stated in aletter to the NRC dated June 30, 1982 (Reference S-11), that since no modifications wererequired for Alternate

Shutdown, no plans were submitted to the NRC for review.Page 16 of 81 L-PI-15-041 NSPMEnclosure 1The PINGP Hot Shutdown Panel does not meet the definition of a primary control stationbecause:* It has not been reviewed and approved by the NRC, and" The Hot Shutdown Panel contains required indication, but it does not contain therequired system and component

controls, plant parameter indications, andcommunications so that the operator can adequately and safely monitor andcontrol the plant using alternative shutdown equipment.

NRC Request (SSA RAI O1.b):b. As described in FAQ 07-0030, "Establishing Recovery

Actions,

" Revision 5, if a licenseeproposes to make modification(s) to their previously approved

strategy, it may obtainNRC approval of a new PCS strategy.

NRC approval of the new PCS strategy can beobtained by providing the information required in FAQ 07-0030 for either 1) Option 1, todesign and install a primary control station(s) in accordance with the guidance andrequirements of the existing Fire Protection licensing, or 2) Option 2, to develop thedesign and analyze the primary control station(s) using the performance-based approach and provide the necessary evaluation.

If the actions to enable the HSD panel and establish the required controls are credited inthe performance based analysis in the PRA Alternative shutdown notebook as PCSactions, then provide a detailed description of the modification to the dedicated oralternative shutdown strategy sufficient for the staff to verify that the strategy meets theattributes provided in Section C.2.4 [i.e., electrical independence, command and control,instrumentation, actions necessary to enable (if required),

etc.].NSPM Response (SSA RAI 01.b):b. Although the Hot Shutdown Panels in the Auxiliary Feedwater (AFW) Pump rooms arethe primary command and control center after abandoning the control room due to fire(see response to subpart c below for more detail),

the Hot Shutdown Panels are notcredited as the Primary Control Station for the purpose of determining Recovery Actions.All actions taken outside the control room after abandoning due to fire are considered tobe Recovery Actions and the additional risk of those actions is included in the Fire PRAAnalysis.

NRC Request (SSA RAI O1.c):c. For a fire requiring CR abandonment, please describe the actions necessary to enablethe HSD panel and/or PCS to establish required

controls, including required local/remote indications.

Clarify if these actions are identified in LAR Attachment G.NSPM Response (SSA RAI 01.c):c. After the decision to abandon the control room has been made by the Unit 1 ShiftSupervisor, operators make the announcements, don an SCBA if necessary, and travelPage 17 of 81 L-PI-15-041 NSPMEnclosure 1to the Hot Shutdown Panels in the AFW Pump room (per procedure F5 Appendix B,"Control Room Evacuation (Fire)").

The Hot Shutdown panels are credited for processmonitoring indication only. There are no switch manipulations required to activate, orisolate the process monitoring circuits at Hot Shutdown Panels as the processmonitoring indication is always active and isolated from the effects of fire in the ControlRoom, and Relay and Cable Spreading Room via always-active instrumentation isolation devices.

Therefore, no actions are identified in Attachment G.NRC Request (SSA RAI 01.d):d. When CR abandonment is necessary, please describe the command and controlstructure of the operating crew, including locations, communications and coordination required between the decision makers (SROs) and field operators performing recoveryactions.For those actions that are symptom-driven, discuss the instrumentation/indications usedas cues to determine that the action is required.

NSPM Response (SSA RAI O1.d):d. The command and control structure of the operating crew is described in PINGPProcedure F5 Appendix B, Control Room Evacuation (Fire). When the decision is madeto abandon, operators take individual attachments to the procedure and travel to variousparts of the plant to perform local actions to support achieving and maintaining Mode 3Hot Standby.

The last step of their attachment directs them to the AFW Pump room toreport to the Unit 1 Shift Supervisor.

Face-to-face communication is credited forrequired communication.

Procedure F5 Appendix B, states symptoms that may necessitate the need to implement alternate shutdown as evidenced by: "a loss of Control Room control of critical plantfunctions which cannot be adequately addressed by ARP, AOP, IFG or EOP responseactions."

There are no specific symptoms identified that would force abandonment; thedecision is left to the judgment of the Unit 1 Shift Supervisor.

This discretion isappropriate given the risk associated with abandoning all control from the Control Roomto rely solely upon one train of Safe Shutdown Equipment outside of the Control Room.The table directly below lists the symptom-driven actions and their respective cues(indication):

Required Indication Action Driven by indication 1 LI-433C Charging pump speed is adjusted using level2LI-433C indication, obtained from 1 LI-433C and 2LI-433Cfor the Unit 1 and Unit 2 pressurizers respectively, to maintain pressurizer level as necessary.

Page 18 of 81 L-PI-15-041 Enclosure 1NSPMRequired Indication Action Driven by indication 1LI-487A TDAFW pump discharge flow is adjusted by2LI-487A manually throttling MV-32238 (11 AFW TO 11 SGFl-18032 MV) for Unit 1 or MV-32246 (22 AFW TO 21 SGF1-18035 MV) for Unit 2, >200 gpm as indicated by Fl-18032 and Fl-1 8035 for the 11 and 22 TDAFWpumps respectively, to maintain S/G level,obtained from 1 LI-487A and 2LI-487A for 11 and21 S/Gs respectively, as necessary.

55001 IF 55001 indicates a loss of cooling waterpressure with D1 running, then stop #1Emergency diesel generator.

11330 (11 CLG WTR Strainer Inlet PI) Check the differential pressure across the 11, 12,11395 (12 CLG WTR Strainer Inlet PI) 21, and 22 cooling water strainers via their11397 (21 CLG WTR Strainer Inlet PI) respective inlet and outlet pressure indicators.

IF11625 (22 CLG WTR Strainer Inlet PI) the DP across the affected strainer is greater than11370 (11 CLG WTR Strainer Outlet PI) 4 psid, and it is not backwashing, THEN attempt to11396 (12 CLG WTR Strainer Outlet PI) manually operate the strainer motor per F5 App. B11523 (21 CLG WTR Strainer Outlet PI) Attachment C step BB.11626 (22 CLG WTR Strainer Outlet PI)PI-11054 Switchover to CIg Wtr from CST for AFWPPI-1 1081 suctions will be necessary when AFWP suctionpressure approaches 3 psi. IF it becomesnecessary to switchover the AFWP suction to CIgWtr, THEN perform the following:

1. OPEN the following MCC Breakers:

a. 1 Al -A2, 11 TD AFW PMP SUCT CLG WTRSPLY MV-32025b. 2A2-A3, CLG WTR TO 22 TD AFW PMP SUCTMV-320302. Locally OPEN the following MVs:a. MV-32025, 11 TD AFW PMP SUCT CL SPLYMVb. MV-32030, 22 TD AFW PMP SUCT CL SPLYMVOther actions not listed which are driven by local indication include:

manually re-positioning valves due to local valve indication, locally starting/stopping motors due tolocal indications, and locally re-positioning breakers due to local breaker indication.

NRC Request (SSA RAI 01.e.i):e. In the alignment basis for Attribute 3.1.1.4 of NEI 00-0 1, the licensee stated that prior toabandoning the CR, action is taken on the control board to close the power-operated relief valves (PORVs),

and additional actions are taken at a switch panel (to be installed as modification item 27 in Attachment S of the LAR) to isolate excess letdown, headvents, pressurizer vents, the pressurizer PORV, and pressurizer heaters.Page 19 of 81 L-PI-15-041 NSPMEnclosure 1(i) Please clarify if the actions at the new switch panel are required to be successful inorder to meet the nuclear safety performance criteria of Section 1.5. 1 (b) associated with RCS inventory and pressure control.

If so, describe the methodology forensuring that manual actuation of these switches is successful and discuss how themanual actuation aligns with Attribute

3. 1. 1.10 of NEI 00-01.NSPM Response (SSA RAI 01.e.i):(i) NSPM would like to clarify that the action described in the Alignment Basis forAttribute 3.1.1.4 in Attachment B is to close the PORV block valves (MV-32195, MV-32196, MV-32197, and MV-32198) on the main control board. The action to isolatethe PORV control valves (CV-31231, CV-31232, CV-31233, and CV-31234) at thenew isolation panel installed as part of Modification Item #27 in Attachment S, TableS-2, is not credited in the Nuclear Safety Performance Criteria; this modification is forrisk reduction purposes only. The action to de-energize the PORV control valvesoutside the control room (at DC switch panels in the Battery Rooms) is the creditedaction to assure that the PORVs are closed and remain closed, thus meeting therequirements of NSPC 1.5.1 (b).The new isolation switches are being installed to reduce Fire PRA risk only, and are*not credited by the NSCA; and therefore, the provisions of Attribute 3.1.1.10 of NEI00-01 do not apply. PINGP is transitioning the exemption that allows closure of thePORV block valves from the control room prior to abandonment and de-energizing the PORV control valves outside the control room at the DC switch panels to ensurethe PORV relief path is isolated (described in LAR Attachment K, "Appendix RExemption, Control Room, Use of repair to remove fuses (III.G.1 criteria),

Fire Area13;" and as clarified in Attachment T of the PINGP LAR.NRC Request (SSA RAI O1.e.ii):

(ii) Please provide a description of the switch design and location.

Include in thediscussion how the switch provides the desired electrical isolation forthe subject components.

NSPM Response (SSA RAI O1.e.ii):

(ii) The design function of the PORV switch is to preclude or isolate (mitigate) hot shortsfrom causing spurious opening of the PORVs. This will be accomplished by breakingthe circuit power connection of the solenoid valve that is not affected by a fire in thearea. The design will be such that hot shorts at this switch in the control room wouldnot be able to cause spurious operation of the components.

This modification is forrisk reduction only, and is not credited in the NSCA. This modification, Table S-2,item #27, is still in the design phase.

References:

S-1 NSP Letter, D. Musolf to Director, Office of Nuclear Reactor Regulation, "Fire Protection Safe Shutdown Analysis and Compliance with Section III.G of 10 CFR Part 50,Appendix R, Including Requests for Relief,"

dated June 30, 1982 (NRC PDR No.8207060240 820630).Page 20 of 81 L-PI-15-041 NSPMEnclosure 1SSA RA102The Executive Summary in the LAR states that the PINGP transition process was performed inaccordance with RG 1.205, Revision

1. Regulatory Position 2.3.2, Previously NRC-Approved Alternatives to NFPA 805, Section 4.2.3, Deterministic Requirements, in RG 1.205, Revision 1,states that in accordance with NFPA 805, Section 2.2.7, licensees may use existing exemptions

... to demonstrate compliance with the specific deterministic fire protection design requirements in Chapter 4 of NFPA 805, provided the NRC staff determines that the licensee has acceptably addressed the continued validity of any exemption

... in effect at the time of the NFPA 805license amendment application and that the exemption

... does not involve a recovery action asdefined in NFPA 805, Section 1.6.52, that is used to demonstrate the availability of a successpath for the nuclear safety performance criteria.

In Attachment K of the LAR, the licensee stated that the previously approved exemption to allowmanual removal of fuses from the PORV control circuit in the event of a fire in the Control Room(CR) (Fire Area 13) will be transitioned in NFPA 805 as clarified in LAR Attachment T, to allowopening of disconnect switches installed subsequent to the exemption

approval, in lieu of pullingfuses. Clarify which action, or both, is credited for disabling the PORV control circuits.

Please describe the procedural steps and the feasibility analysis performed for these actions.NSPM Response (SSA RAI 02):To clarify which action PINGP is crediting to disable PORV control circuits, NSPM is revisingAttachment T to the LAR, which clarified a previously approved exemption.

The requested clarification will allow for the opening of disconnect switches in lieu of the original exemption action of pulling fuses as a follow-on action taken outside of the Control Room, to prevent fire-induced spurious opening of the PORVs. The manual operation to open disconnect

switches, demonstrated by PINGP to be feasible and reliable, is simpler than pulling fuses and isequivalent in intent and function.

A revision to Attachment T to describe this change is includedin Enclosure 4 to this letter.Table S-2 Modification Item #27, to install isolation switches for the PORVs, etc. is credited inthe Fire PRA for risk reduction purposes only, and as such, will not eliminate the need to takethe follow-on action of opening the disconnect switches for fire scenarios involving a ControlRoom evacuation.

The procedural steps for assuring PORV isolation during a Control Room fire are as follows:PINGP Procedure F5 Appendix B, Control Room Evacuation (Fire)* Attachment C Ul RO Actions:

Step F. CLOSE pressurizer PORV block valves.(Performed in the control room after reactor trip, turbine trip, RCP trip, MFW trip, andMSIV closure)* Attachment D U2 RO Actions:

Step F. CLOSE pressurizer PORV block valves.(Performed in the control room after reactor trip, turbine trip, RCP trip, MFW trip, andMSIV closure)* Attachment B U2 Shift Supervisor Actions:o Step A. Don SCBAo Step B. Proceed to 11 Battery Room (Unit 1, Train A)Page 21 of 81 L-PI-15-041 NSPMEnclosure 1o Step C. Turn off switch PNL 11-8 which powers PNL 171 which powers CV-31232Ul PZR PORV Ao Step D. Proceed to 12 Battery Room (Unit 1, Train B)o Step E. Turn off switch PNL 12-8 which power PNL 181 which powers CV-31231 UlPZR PORV Bo Step F. Proceed to 22 Battery Room (Unit 2, Train B)o Step G. Turn off switch PNL 22-10 which powers PNL 281 which powers CV-31233U2 PZR PORV Bo Step H. Proceed to 21 Battery Room (Unit 2, Train A)o Step I. Turn off switch PNL 21-10 which powers PNL 271 which powers CV-31234U2 PZR PORV AThe feasibility of these actions has been evaluated by existing PINGP Calculation, GEN-PI-055, "Manual Action Feasibility."

Appendix D to this calculation provides a timeline to perform theseactions as described in PINGP procedure F5 Appendix B, "Control Room Evacuation (Fire)."These actions (from Attachments B, C, and D, as listed above) have been walked down bymultiple operating crews to validate the timing, and are included in regular training:

" The Attachment B, U2 Shift Supervisor steps A-I are completed in 9 minutes.* The Attachment C, Ul RO action to close Unit 1 PORV block valves is completed in 1minute." The Attachment D, U2 RO action to close Unit 2 PORV block valves is completed in 1minute.LAR Attachment S, Table S-3 Item 53 was inadvertently deleted in the 2014 LAR Supplement; this item will be re-instated and revised as follows, to update GEN-PI-055 to reflect PINGP'stransition to NFPA 805, and to document how the feasibility criteria of FAQ 07-0030 is met:Item 53:Update GEN-PI-055, "1 OCFR50 Appendix R Manual Action Feasibility Study," to reflectPINGP's transition to NFPA 805, including addition of new recovery

actions, actions tomaintain safe and stable conditions, and to document how the criteria, as defined byFAQ 07-0030, are met.SSA RAI 03NFPA 805, Section 2.4.2, Nuclear Safety Capability

Analysis, requires licensees to perform anuclear safety capability analysis (NSCA). In RG 1.205, the NRC staff states that oneacceptable approach to perform the NSCA is to follow the guidance in NEI 00-01, Chapter 3.Attribute 3.2.1.5 of NEI 00-01 states that instrument circuits (e.g., resistance temperature detectors, thermocouples, pressure transmitters, and flow transmitters) should be assumed tofail upscale,

midscale, or downscale as a result of fire damage, whichever is worse, and that aninstrument performing a control function is assumed to provide an undesired signal to thecontrol circuit.

In Attachment B of the LAR, Section 3.2.1.5, the licensee stated that PINGPassumes that instrumentation circuits fail in their worst-case positions when damaged by the fireunless an analysis was performed to show that the failure mode is incredible.

Page 22 of 81 L-PI-15-041 NSPMEnclosure 1Please describe the analysis method used to determine the failure mode is incredible andprovide an example of the types of instruments that would be analyzed using this method.NSPM Response (SSA RAI 03):In general, PINGP assumed that instrumentation circuits involving shielded twisted pair cableswill fail in the worst-case mode (upscale, downscale or midrange);

however, in a very few cases,additional analysis was performed in the PINGP Non-Power/NSCA Operations Review forNFPA 805, EC 20612, to determine that the worst-case failure mode is incredible, and toidentify the true fire-induced failure mode. A description of the methodology and evaluation

criteria, along with an example circuit, is provided below.A fire that directly affects cabling in instrumentation circuits can produce open circuits, conductor-to-conductor bolted or resistive hot shorts (inter-cable or intra-cable),

and shorts toground. As stated in the RAI, these fire-induced faults may drive the process variable upscale,downscale, or midscale.

To determine that the worst-case failure mode is incredible, instrument loops were evaluated and determined to meet all of the following criteria:

1. Fire-induced faults that cause an open circuit on the target conductors cannot fail thecredited function

/ component in an undesired position/state.

This criteria is the fundamental building block (others to follow) for the concept that a loss of signal to the credited function

/component shall not be capable of causing an undesired consequence.

2. The circuit must be capable of being positively isolated from any signals (interlocks, processvariables, etc.), that could cause an undesired output or positioning.

For emphasis, manualisolation of the circuit is required so that the circuit may be placed in its lowest output(equivalent to fail-safe) state. The circuit CANNOT require manual adjustment beyond itslowest output state, and cannot rely upon, nor can it be affected by any automatic functions, interlocks, etc. Additionally, by requiring that the circuit be in its lowest state, leakagecurrent away from the circuit and wire-to-wire resistive shorting is assured to not be ofconcern.

This criteria is in keeping with the guidelines provided in NUREG/CR-7150, Volume 1, Section 5.0.3. Open circuits, conductor-to-conductor bolted or resistive hot shorts (inter-cable or intra-cable), and shorts to ground must be demonstrated to have no effect on the credited portionof the circuit, or be demonstrated to not be credible (e.g., due to the physical design,credited state of the circuit, etc.).a. Example:

In the partial schematic directly below, an open circuit on the conductors between the hand controller and the E/P (clouded conductors),

cannot cause the valveto fail in the undesired position.

Page 23 of 81 L-PI-1 5-041Enclosure 1NSPM4. The device performing the isolation and controlled level output (e.g., hand controller),

cannotbe located in the area where the fire is occurring.

This device must be capable of positively isolating any process variable inputs and interlocks that may inadvertently cause anundesired output or positioning.

5. Components exposed to the fire (e.g., cable, valve, E/P, positioner),

cannot be credited toprovide outputs or positions that differ from the fail-safe (low/no-signal) output or position.

Additionally, the fire-exposed components must not contain electronic systems that aresusceptible to producing sustained output spiking, or are capable of independently generating an output signal that may be out of proportion to the input signal.-a. Example:

An E/P (I/P) that operates on the principle that a magnetic coil positions amechanical beam that regulates a mechanical air relay (force-balance),

would beacceptable as the coil is not capable of generating a sustained signal greater than it'sinput. An E/P that operates on the principle that an analog or digital input signal iso converted to an output signal via a microcontroller in the E/P, would not be anacceptable configuration as the output is capable of being driven independently of theinput.b. Additionally, the component must not have a fail / hold last state function (sometimes referred to as a lock-up function).

6. For the fire area of concern (location of the fire), any given fault(s) cannot cause an isolation of the loop regulating device (e.g., transmitter, controller),

such that the loop power supplycan apply full current to the credited portion of the circuit.

This criterion is fully dependent upon the specific design of the loop circuit.a. Example:

As can be seen from the drawing directly below, the power supply to thecredited controlled output is integral to the hand controller itself (all other portions of thePage 24 of 81 L-PI-15-041 Enclosure 1NSPMcircuit will be isolated in manual mode); therefore, when the hand controller is taken tomanual, cable faults between the hand controller and the E/P cannot isolate the handcontroller (regulating portion of the credited circuit) from the power supply.ransmitter RA9IC t4 IMTRUewrI' I ~ I~lýW ý'4.7c-)~ ~ ~ + (4 p .. , --.4- -: -(!4L- i1%.+4-602. 15(.4 19 L ý)TE RM. CABINFL 1238i g mum~ AZ~tLI I-"r]'----I.4:1)LC-S117. The portion of the circuit exposed to the fire cannot contain more than a single twisted pair.This criteria prevents intra-cable shorts from occurring that could increase the signalindependent of the hand controller from wire to wire shorts and internal re-referencing of theshield.8. The credited cable shield must be of a braided type (no foil). This criteria, coupled with thecriteria as described below, prevents inter-cable faults from falsely increasing the signal onthe circuit.a. Example:

The cable shown in the drawingtwisted pair cable with a braided shield.directly below is a typical single pair, shieldedB1W PART A,-8437-H-OO2 PIONEER SERVICE 5' ENG. CO.I Lt.E ;-I A -.-.,, 2K ;

",G.- c..,V'A*liv- es,,:L7.)

vy- .llo4 .

U; 13 .ý .l/u- 7-- 9 / ,d/Na,,v 7;9/L-,f ,zfdu*_cc.-

9. The shield must be continuous and must be tied to ground. This criteria, coupled with thecriteria described above, prevents inter-cable faults from falsely increasing the signal on thePage 25 of 81 L-PI-15-041 NSPMEnclosure 1credited portion of the control circuit.

Open circuiting of the cable due to cleaving (e.g.falling debris) may cause a loss of ground on the shield conductor;

however, this would alsocause a loss of reference of the internal conductors and thus, re-referencing of a cleaved(ungrounded shield) would not be able to cause an undesired effect to the circuit.Additionally, the circumferential design of a braided shield conductor aids in its ability toremain in continuity in the case of a partial cleave (i.e., where only one internal conductor iscleaved).

a. Example:

The block diagram directly below demonstrates a typical shield termination scheme for the plant. Shields are continuous and tied to ground.PB-10©I 7--G0)61Conclusion:

The as-built circuit must meet all of the above criteria for a fault (fault consequence) on aninstrument loop to be considered "incredible."

This method is in keeping with the guidanceprovided in NUREG/CR-7150, Section 5.0.LAR Attachment S, Table S-3 Item 51 will be updated to include addition of this specificmethodology to Engineering Manual 3.4.3, "Safe Shutdown Circuit Analysis."

Item 51 will bechanged as follows:Item 51:Revise EM 3.4.3, "Safe Shutdown Circuit Analysis" to incorporate applicable details ofvendor document EPM-DP-EP-004 as well as the detailed methodologqy for analyzing shielded twisted pair instrumentation and controls circuits as referenced in EC 20612,"PINGP Non-Power/

NSCA Operations Review for NFPA 805."Page 26 of 81 L-PI-15-041 NSPMEnclosure 1SSA RAI 04NFPA 805, Section 1.3.1, Nuclear Safety Goal, states: "The nuclear safety goal is to providereasonable assurance that a fire during any operational mode and plant configuration will notprevent the plant from achieving and maintaining the fuel in a safe and stable condition."

NFPA 805, Section 1.5. 1 (d), Vital Auxiliaries, states: "Vital auxiliaries shall be capable ofproviding the necessary auxiliary support equipment and systems to assure that the systemsrequired under (a), (b), (c) and (e) are capable of performing their required nuclear safetyfunction."

In Section 4.2.1.2 of the LAR, the licensee stated that CR temperature will remain belowequipment limits for up to 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> with actions taken only within the CR itself. A portable fanmay be used to maintain temperatures below equipment limits indefinitely.

If required, theportable fan will be powered by a designated welding receptacle or a 480-VA C [volts alternating current]

portable generator located outside the building.

Please provide the following additional information:

[This RAI includes Subparts a through f, as shown below along with NSPM responses]

NRC Request (SSA RAI 04.a):a. Describe the steps taken to maintain CR temperature below equipment limits for 36hours.NSPM Response (SSA RAI 04.a):a. The steps to maintain CR temperature below equipment limits for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> are definedin C37.9 AOP1, "Loss of Control Room Cooling,"

and include opening doors / vent pathsfrom the Control Room. The evaluation identified in LAR Section 4.2.1.2 (EC 23925)validated that the required actions specified in C37.9 AOP1 (opening doors / vent paths)will maintain control room temperature below the critical temperature limit for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.NRC Request (SSA RAI 04.b):b. Clarify if the use of the portable fan and the associated welding receptacle or portablegenerator is credited for achieving and maintaining safe and stable conditions.

Ifso, provide the justification for excluding this recovery action in Attachment G of theLAR.NSPM Response (SSA RAI 04.b):b. As stated in LAR Section 4.2.1.2, "A PINGP thermal-hydraulic analysis was performed for a mission time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to assure that safe and stable conditions can be achievedwithin that time period."

The portable fan is used at approximately 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, which isafter safe and stable conditions have been achieved (within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />), and therefore, theaction to power the portable fan is not considered to be a recovery action.Page 27 of 81 L-PI-15-041 NSPMEnclosure 1NRC Request (SSA RAI 04.c):c. Provide additional details on the storage and usage locations of the 480-VAC portablegenerator and its potential impact, if any, on the NSCA structures,

systems, andcomponents (SSCs) that are in the vicinity of these locations.

NSPM Response (SSA RAI 04.c):c. The portable generator will be stored and operated outside the Power Block as definedin Attachment I to the LAR, and therefore, it will not be operated in the immediate vicinityof NSCA SSCs. Details regarding connection of the portable generator to the portablefan will be determined as part of Implementation Item #63 (Table S-3), as revised in theresponse to SSA RAI 04.d.NRC Request (SSA RAI 04.d):d. Describe the type and quantity of fuel associated with the portable generator and theavailability and location(s) of sufficient fuel sources to support maintaining safe andstable conditions for the time period required.

NSPM Response (SSA RAI 04.d):d. The portable generator will be fueled with diesel fuel. Attachment S, Table S-3, Item #63will be revised to have procedural controls for the operation, maintenance,

storage, andrefueling of the portable generator to support maintaining safe and stable conditions.

The revision will be as follows:Item 63:Provide procedural guidance to connect a diesel powered portable generator locatedoutside the power block to power a temporary fan for the Main Control Room to maintainsafe and stable conditions.

Additionally, procedural guidance shall be provided for theoperation, maintenance, stora-ge, and refueling of the portable generator, and for trainingand drills.NRC Request (SSA RAI 04.e):e. Justification that refueling the portable generator does not present a fire exposurehazard to NSCA SSCs.NSPM Response (SSA RAI 04.e):e. The portable generator will be located and operated outside the Power Block; therefore re-fueling will be performed

outside, and not in the vicinity of NSCA SSCs.NRC Request (SSA RAI 04.f):f. A summary of the procedure guidance for the use of the portable generator to power theportable fan, and how this action aligns with each of the feasibility criteria ofFAQ 07-0030 (i.e., training, procedures, drills, etc.).Page 28 of 81 L-PI-15-041 NSPMEnclosure 1NSPM Response (SSA RAI 04.f):f. Procedural guidance for use of the portable generator to power the portable fan is stillunder development and will be finalized per the modification process.

At this time,guidance is planned to include the following (note that if power is available, the portablefan will be powered from a welding receptacle):

1. Place a power cart* on the turbine deck (if the portable generator will beneeded).2. Place the fan to take suction from the Control Room and direct it into theRecords Room.3. Plug the fan into the power cart (or the welding receptacle, as noted above).4. Move the portable generator to a suitable service location.

5. Connect the power cart to the portable diesel generator.

6. Start and operate the portable generator, including refueling instructions.

7. Start and operate the portable fan and power cart.* Note that the power cart has not been designed as of this response;

however, it isexpected that the power cart will contain the properly sized fused disconnect, adequately sized cabling and connections to connect the portable fan to thegenerator.

Procedural guidance will also be developed for training and drills.These actions, taken to maintain safe and stable conditions, will be evaluated forfeasibility against the criteria of FAQ 07-0030 as described by the reinstated and revisedLAR Attachment S, Table S-3 Item #53, as described in the response to SSA RAI 02.SSA RAI 05NFPA 805, Section 1.3. 1, Nuclear Safety Goal, states: "The nuclear safety goal is to providereasonable assurance that a fire during any operational mode and plant configuration will notprevent the plant from achieving and maintaining the fuel in a safe and stable condition."

NFPA 805, Section 1.4.1, Nuclear Safety Objectives, states: "In the event of a fire during anyoperational mode and plant configuration, the plant shall be as follows:(1) Reactivity Control.

Capable of rapidly achieving and maintaining subcritical conditions (2) Fuel Cooling.

Capable of achieving and maintaining decay heat removal and inventory control functions (3) Fission Product Boundary.

Capable of preventing fuel clad damage so that the primarycontainment boundary is not challenged."

In Section 4.3.2 and Attachment D of the LAR, the licensee provided the results of theevaluation process for Non-Power Operations (NPO) analysis.

Please provide additional details as follows:Page 29 of 81 L-PI-15-041 Enclosure 1NSPM[This RAI includes Subparts a, b, and c, as shown below along with NSPM responses]

NRC Request (SSA RAI 05.a):a. During NPO modes, spurious actuation of valves can have a significant impact on theability to maintain decay heat removal and inventory controlProvide a description of any actions being credited to minimize the impact of fire-induced spurious actuations on power operated valves (e.g., air-operated valves and motor-operated valves) during NPO (e.g., pre-fire rack-out, actuation of/or pinning of valves,and isolation of air supplies).

NSPM Response (SSA RAI 05.a):a. During NPO modes, there are four RHR suction valves in each unit where spuriousactuations could impact decay heat removal, as shown in the diagrams below. A pre-fireaction is taken prior to entering higher risk evolutions to tag open the breakersassociated with two of the four RHR suction valves, per procedures 10C1.3-M5 and2C1.3-M5, "Unit 1 (2) Shutdown to Mode 5", for each unit, MV-32164 (Unit 1), MV-32231(Unit 1), MV-32192 (Unit 2), and MV-32233 (Unit 2) during shutdown

cooling, mode 5, tominimize the impact of fire-induced spurious actuations of RHR suction valves from theRCS Hot Leg. During higher risk evolutions PINGP will use administrative controls listedon page D-10 of the LAR, to minimize the risk of potential fire damage that could impactoperation of the following RHR suction valves: MV-32165, MV-32230, MV-32193, andMV-32232.

Unit 1 RHR Suction ValvesPage 30 of 81 L-PI-1 5-041Enclosure 1NSPMUnit 2 RHR Suction ValvesNRC Request (SSA RAI 05.b):b. During normal outage evolutions, certain credited NPO equipment will have to beremoved from service.Describe the types of compensatory actions that will be used during such equipment down-time and how are they determined to be adequate.

NSPM Response (SSA RAI 05.b):b. Updates to procedure 5AWI 15.6.1, 'Shutdown Safety Assessment' will ensure that NPOcredited equipment is not removed from service during High Risk Evolutions (HRE)without adequate compensatory measures.

Plant procedures will provide guidelines andidentify compensatory actions that can be taken when fire safe shutdown components are out of service.

Hot Work and Safe Shutdown Safety Assessment procedures will berevised during NFPA 805 implementation (Table S-3, Items 37 and 40), and thefollowing types of compensatory actions will be added / retained for fire risk mitigation:

Hot Work Restrictions, Transient Combustible

Controls, Access Limitations, Automatic Detection and Suppression

Systems, Fire Watch Patrols, etc.Additional Key Safety Function (KSF) pinch points introduced by removal of creditedequipment from service will be identified through administrative procedures, shutdownrisk management, and work control.

Also in the unlikely event that such equipment isdeliberately removed from service coincident with a planned or emergent HRE, thePage 31 of 81 L-PI-15-041 NSPMEnclosure 1plant's Fire Protection Team will consider appropriate contingency measures to reducefire risk at the additional locations.

NRC Request (SSA RAI 05.c):c. In Attachment D of the LAR, the licensee states that operator actions taken to mitigatethe loss of a Key Safety function (KSF) are credited in the NPO analysis contained withinPINGP Engineering Evaluation EC-20612, "Non-Power/NSCA Operations Review forNFPA 805."Describe the operator actions credited to maintain KSF and the feasibility analysisperformed for these actions.NSPM Response (SSA RAI 05.c):c. Types of actions credited to maintain KSF include the following:

de-energizing poweroperated valves to preclude or mitigate spurious operation, adding procedural actions toopen a redundant flow path, re-powering an Instrument Bus from the alternate panel,and closing manual valves to isolate flow diversion paths.The feasibility analysis contained in GEN-PI-055 will be revised as shown in thereinstated and revised S-3 table item #53, as described in the response to SSA RAI 02.SSA RA106NFPA 805, Section 1.3.1, Nuclear Safety Goal, states: "The nuclear safety goal is to providereasonable assurance that a fire during any operational mode and plant configuration will notprevent the plant from achieving and maintaining the fuel in a safe and stable condition."

In Section 4.2.1.2 of the LAR, the licensee stated that the determination of the final state of thesafe and stable conditions will be based upon the extent of the fire damage, the inventory remaining in the Refueling Water Storage Tank (RWST), the ability to provide makeup to theRWST, and the ability to re-establish inventory in the Condensate Storage Tank (CST)or realignment of Auxiliary Feedwater (AFW) to its alternate source (cooling water system).Please provide the additional information to support the review of the NFPA 805 licensing basesfor maintaining the fuel safe and stable:[This RAI includes Subparts a and b, as shown below along with NSPM responses]

NRC Request (SSA RAI 06.a):a. The licensee stated that the PINGP thermal-hydraulic analysis was performed for amission time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to assure that safe and [stable]

conditions can be achievedwithin that time period.Page 32 of 81 L-PI-15-041 NSPMEnclosure 1Provide a qualitative risk assessment for extending the mission time beyond 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />sand implementing operator actions to establish makeup to the RWST, the CST and/orrealignment of AFW to its alternate source.NSPM Response (SSA RAI 06.a):a. The qualitative risk assessment for actions to maintain safe and stable conditions is asfollows:The actions to refill the RWST, CST, and re-align AFW to the alternate cooling watersource have a low risk associated with them. It is assumed that every fire causes a tripof both Unit 1 and 2, and a loss of Main Feedwater.

It is likely that many fires would notcause a reactor trip and loss of Main Feedwater, which reduces the likelihood that theseactions would need to be performed.

The actions to makeup to the RWST, the CST,and realign AFW suction are well described in the existing plant procedures; there aremultiple sources available; and the operators are familiar with these actions.

After 24hours, additional personnel will be available from the Emergency Plan, as well asadditional resources (e.g., power, vehicles, equipment, etc.) to support performing theseactions.

Even if the CST were to be depleted, the AFW pumps can be re-aligned to takesuction from the Mississippi River, providing a virtually unlimited supply of water. TheAFW re-alignment can be performed from the control room, or locally by opening thecooling water supply valve without having to stop the AFW pump. There is CST levelindication and indication for AFW suction pressure to provide diverse indication.

Because the Reactor Coolant Pump Seals have been replaced with seals that are notinitially subject to excessive

leakage, there is substantial time available to make up tothe RWST. The single and multiple spurious combinations that could cause RWST orCST drain down events have been addressed by the thermal-hydraulic analysisdescribed in EC 20736, "Reactivity Control,"

and EC 20738, "Decay Heat Removal"(References 6.56 and 6.57 in the LAR). Therefore the dependency on the RWST tomake-up to the RCS is minimized.

There are redundant RWST level indicators and lowlevel annunciators in the control room.Beyond time frames of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, existing plant procedures specific to the situation canbe implemented to establish system alignments and to use equipment in ways differing from normal plant practice or training.

Such procedures and equipment address such awide spectrum of possibilities that it is not considered useful to develop all possiblerecovery contingencies.

Additional guidance for options beyond those described in theplant AOPs and EOPs is provided by the PINGP document ASB, "Alternative Sources ofPower, Water and Air Book," which may be used in formulating strategies in an eventthat has placed the plant in a condition that goes beyond the normal and emergency useof operating procedures and equipment.

Therefore the risk associated with these actions is low and PINGP can maintain safeand stable conditions indefinitely.

NRC Request (SSA RAI 06.b):b. The licensee also stated that operator actions are performed to align makeup/alternate water sources to the RWST beyond 38 hours4.398148e-4 days <br />0.0106 hours <br />6.283069e-5 weeks <br />1.4459e-5 months <br />, and to the CST or AFW pumps beyond 20hours.Page 33 of 81 L-PI-15-041 NSPMEnclosure 1Provide the details of the makeup sources and the process for aligning these sources,including a discussion of the feasibility analysis performed for these actions.NSPM Response (SSA RAI 06.b):b. It is noted that the preferred method to maintain the plant(s) in a shutdown condition withthe required Keff level will be to use the Boric Acid Storage Tanks (BASTs) with theReactor Makeup (RMU) Tanks to blend to the Volume Control Tank (VCT).Methods and Sources by which Replenishment of the RWST May beAccomplished:

The RWST can be replenished from several different sources if and when required:

1. The RWST can be refilled from the RMU tanks and BASTs using the BA Blenderper procedure 1C12.5, "Unit 1 Boron Concentration Control,"

or procedure 2C12.5, "Unit 2 Boron Concentration Control"2. Water can be transferred from the other unit's RWST per procedure C1 6, "SpentFuel Cooling System"3. Water can be transferred from the CVCS Holdup tank(s) per procedure C12.8,"CVCS Holdup, Monitor, and Concentrates Holdup Tanks"4. Water can be transferred from the CVCS Monitor tank(s) per procedure C12.8,"CVCS Holdup, Monitor, and Concentrates Holdup Tanks"Methods and Sources by which Replenishment of the CST May be Accomplished:

The CST can be replenished from several different sources by unit, unit-to-unit, and byvarious methods if and when required using procedure 1C28.1 AOP2, "Loss ofCondensate Supply to Aux Feedwater Pump Suction" or 2C28.1 AOP2, "Loss ofCondensate Supply to Aux Feedwater Pump Suction" (note that the PINGP Unit 1 andUnit 2 CSTs are normally cross-tied and therefore, their collective volumes will beavailable to either unit without any additional action):1. Water can be transferred from the Water Treatment System2. Water can be transferred from the Condenser

a. Using the Condensate Pumpb. Using the Condenser Spray Pump3. Water can be transferred from the ADT Monitor Tank (Note that this option isonly available when the ADT Monitor Tank has sufficient level to support thisactivity)

Methods to cross-tie and align cooling water to the AFW System:The AFW system can be cross-connected between units or aligned to the cooling watersystem if and when required:

1. The AFW Pump(s) can be aligned to take suction from the Cooling Water Systemper procedures 1 C28.1 AOP2, "Loss of Condensate Supply to Aux Feedwater Pump Suction" or 2C28.1 AOP2, "Loss of Condensate Supply to Aux Feedwater Pump Suction"2. The AFW systems can be cross-connected per procedures 1C28.1, "Auxiliary Feedwater System Unit 1" or 2C28.1, "Auxiliary Feedwater System Unit 2"Page 34 of 81 L-PI-15-041 NSPMEnclosure 1Existing Plant Procedures Credited for Accomplishing Replenishment of theRWST and CST:Details of the lineups and how the lineups are accomplished are described in thefollowing procedures (procedures have been placed on the Portal):1. C16, "Spent Fuel Cooling System"2. C12.8, "CVCS Holdup, Monitor, and Concentrates Holdup Tanks"3. 1C12.5, "Unit 1 Boron Concentration Control"4. 2C12.5, "Unit 2 Boron Concentration Control"5. 1C28.1 AOP2, "Loss of Condensate Supply to Aux Feedwater Pump Suction"6. 2C28.1 AOP2, "Loss of Condensate Supply to Aux Feedwater Pump Suction"7. 1C28.1, "Auxiliary Feedwater System Unit 1"8. 2C28.1, "Auxiliary Feedwater System Unit 2"A formal feasibility analysis of the methods to replenish the RWST and CST (actions tomaintain safe and stable conditions) is scheduled to be completed during theimplementation phase; however, the methods for RWST and CST replenishment, ascredited by the NFPA 805 analysis, are defined by existing operations procedures, forwhich Operations has received training.

By allowing the use of many different methodsand / or sources to accomplish replenishment, including replenishing from the unaffected unit and aligning to cooling water, there is a high confidence that a method will beavailable for use when and if required.

Additionally, PINGP will reinstate and revise LARAttachment S, Table S-3 Item # 53 as described in the response to SSA RAI 02.SSA RAI 07- Clarification of VFDRs for FA 13 and 18Response to SSA RAI 07 is being provided in separate correspondence by June 26, 2015.SSA RA108NFPA 805, Section 4.2.2, requires that "For each fire area either a deterministic or performance-based approach shall be selected in accordance with Figure 4.2.2. The performance-based approach shall be permitted to utilize deterministic methods for simplifying assumptions withinthe fire area."In Attachment C of the LAR, Table B-3, the licensee indicated that the regulatory basis for FireArea 71 is the deterministic approach as described in NFPA 805, Section 4.2.3. However, inthe table titled "Required Fire Protection Systems and Features",

an Electrical Raceway FireBarrier System (ERFBS) was identified as being required for risk. This implies that the ERFBSwas evaluated using the performance-based approach using the Fire Risk Evaluation approachin accordance with NFPA 805, Section 4.2.4.2.

If the Fire Area 71 regulatory basis is thedeterministic

approach, to maintain compliance with NFPA 805, Section 4.2.3, the ERFBSshould meet the requirements of NFPA 805, Section 3.11.5, and be identified as being requiredfor separation.

Page 35 of 81 L-PI-15-041 NSPMEnclosure 1Please clarify if the ERFBS is credited to protect a nuclear safety performance

function, and ifFire Area 71 was evaluated using deterministic or performance-based approach.

NSPM Response (SSA RAI 08):The ERFBS is credited to protect one train of Pressurizer Level Indication (1 L-433) and meetsthe requirements of NFPA 805 Section 3.11.5. Fire Area 71 meets Deterministic Requirements.

The Fire PRA performed detailed analysis of Fire Area 71, but detailed PRA analysis of the areadoes not necessarily mean that the area needs to be Performance Based (Section 4.2.4.2).

LAR Attachment C should be revised as follows:REQUIRED FIRE PROTECTION SYSTEMS AND FEATURESFire Category ID Type Required?

NotesArea S L E R D71 Feature See ERFBS Y N N Y N Cable 2CF-74 has 3MNote N Interam wrapSSA RA109NFPA 805, Section 2.4.3.3, states that when performing Fire Risk Evaluations, "The PSAapproach,

methods, and data shall be acceptable to the AHJ" (which is the NRC).NFPA 805, Section 3.11.2, Fire Barriers, states that "Fire barriers required by Chapter 4 shallinclude a specific fire-resistance rating. Fire barriers shall be designed and installed to meet thespecific resistance rating using assemblies qualified by fire tests. The qualification fire tests shallbe in accordance with NFPA 251, Standard Methods of Tests of Fire Endurance of BuildingConstruction and Materials, or ASTM E 119, Standard Test Methods for Fire Tests of BuildingConstruction and Materials."

In Attachment C of the LAR, the licensee stated that radiant energy shields are credited for riskin Fire Area 1 to protect raceway ICV- T421 and in Fire Area 32 to protect raceway 1SG-LB22.

Please provide specific details of the nuclear safety functions that credit these radiant energyshields and discuss the extent of how the radiant energy shields are credited in the fire riskevaluations.

In your discussion include how the fire resistance rating claimed in the riskanalysis has been established through fire testing.NSPM Response (SSA RAI 09):No nuclear safety function credit is taken for radiant energy shields and they are not credited infire risk evaluations.

The Fire PRA did not credit the radiant energy shield to protect cables incable trays 1 CV-T421 in FA 1 or 1 SG-LB22 in FA 32 from fire damage. FRE-FA01 and FRE-FA32 will be revised to indicate that the Radiant Energy Shield is not credited for RiskReduction in FA 1 and FA 32. LAR Attachments A, K, and C should be changed as follows:Attachment A, Section 3.11.2 should be revised to delete the discussion regarding FPEE-11-020, for credit is not taken for radiant energy shields.Page 36 of 81 L-PI-15-041 NSPMEnclosure 1Attachment K (page K-18) should be revised to delete the statement that "The existing radiantenergy shield in Unit 1 has been determined to be acceptable in accordance with the NFPA 805performance based approach."

Attachment C should be revised to indicate that the Radiant Energy Shield is not credited forRisk Reduction in FA 1 and FA 32 as follows:REQUIRED FIRE PROTECTION SYSTEMS AND FEATURESFire Category ID Type Required?

NotesArea S L E R D4 Fature -GV- RE_& N N N Y N Racew'ay GCV T42 isT424t wrapped with a radiantREQUIRED FIRE PROTECTION SYSTEMS AND FEATURESFire Category ID Type Required?

NotesArea S L E R ID32 Featue RE-& N N N Y N Cab-le SG I B22has a radiant energyshield (PIES),____ _ _ ___ ___ _____ ________

____ ____ ____ ____ ____ Marinitc boar-dSSA RAI 10NFPA 805, Section 2.4.2.4, requires "An engineering analysis shall be performed in accordance with the requirements of Section [2.4] for each fire area to determine the effects of fire or firesuppression activities on the ability to achieve the nuclear safety performance criteria ofSection 1.5." RG 1.205, Revision 1, endorsed NEI 04-02, Revision 2, as one acceptable approach to performing and documenting the engineering analyses required to transition to arisk-informed, performance-based fire protection program in accordance with 10 CFR 50.48(c)and NFPA 805. On a fire area basis, NEI 04-02 requires that the licensee document how thenuclear safety performance criteria are met. The guidance in NEI 04-02 recommends that thisinformation be presented in Table B-3, Fire Area Transition.

In the LAR, Section 4.2.4,Overview of the Evaluation

Process, Step 5 -Disposition, the licensee states that the finaldisposition of VFDRs should be documented in Attachment C (NEI [04]-02 Table B-3).[This RAI includes Subparts a and b, as shown below along with NSPM responses]

NRC Request (SSA RAI 10.a):a. Attachment S of the LAR, Table S-2, Modification Items #34 and #35 involve protecting cables from fire damage in Fire Areas 32 and 58 to ensure electrical power availability tosupport the nuclear safety performance criteria (NSPC). However, Attachment C of theLAR does not describe the need for these modifications in the fire area assessment forFire Areas 32 and 58.Please discuss how these modifications support the appropriate NSPC and explain whythese modifications are not identified in LAR Attachment C for Fire Area 32 and 58.Page 37 of 81 L-PI-1 5-041 NSPMEnclosure 1NSPM Response (SSA RAI 10.a):a. Modification

1. 6 will re-route 1C-333 out of FA 32 and FA 58 so the 1 RY offsite powersupply will be available in FA 32 and FA 58 allowing the vital auxiliaries NSPC to be met.Modification

1. 34 will protect cables 15402-G, 15402-K, and 1CA-1 140 that could bypassthe sync check switch or relay for the #1 emergency diesel generator output breakerfrom risk significant fire initiators allowing the vital auxiliaries NSPC to be met.Modification

1. 35 will protect cable 1 C-332 from fire damage in fire area 32 and 58 toensure BUS 16 can be re-powered from the 1 RY transformer allowing the vitalauxiliaries NSPC to be met.LAR Attachment C, should be revised to include modification item #s 34 and 35 as wellas modification

1. 6 for fire areas 32 and 58. The disposition of VFDR-058-1-11 in LARAttachment C should be revised to read as follows:* No recovery actions are credited.

, Modifications identified in Table S-2, Item #s 6, 34, and 35 will ensure BUS-15 orBUS-1 6 will remain powered from either the 1 RY or their respective dieselgenerator source." This VFDR has been evaluated and it was determined that the risk, safety marginand defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4with plant modifications credited.

The disposition of VFDR-032-1-02 in LAR Attachment C should be revised to add thefollowing:

0 Modification identified in Table S-2, Item #34 will protect cables, from firedamage in Fire Area 32, associated with the D1 emergency diesel generator's output breaker so that the 1 RY source will remain available to BUS-1 6.* Modification identified in Table S-2, Item #35 will protect cable 1C-332 from firedamage in Fire Area 32 so that the 1 RY source will remain available to BUS-1 6.LAR Attachment S, Table S-2, Item #34 will be revised to include cables 15402-G,15402-K, and 1CA-1 140 in the Proposed Modification Column as follows:Item Rank Unit Problem Statement Proposed In Comp Risk InformedModification FPRA Measure Characterization 34 Medium 1,2 A fire in FA 13, 18, 32 or 58 could damage Protect cables Yes No This will reduce riskcables that control the Emergency Diesel 15402-G, 15402- by makingGenerator output breaker and bypass the K, and 1 CA-1 140 modifications tosync check switch or relay which could to prevent reduce the numbercause a spurious closure of the breaker bypassing the of fire scenarios which could cause a lock-out of the 4 kV sync check switch that could causesafeguards Bus which powers one train of or relay from risk fire damage to asafeguards equipment.

The loss of the Bus is significant fire 4kV safeguards risk significant for some fire scenarios, initiators, bus.Page 38 of 81 L-PI-15-041 NSPMEnclosure 1NRC Request (SSA RAI 1O.b)b. In Table S-2 of Attachment S of the LAR, the licensee describes several modifications that specify "protecting cables or circuits" (e.g., modification

1. 6, #14, etc.).Please describe the protection schemes that may be used for 'protecting" cables.NSPM Response (SSA RAI 10.b):b. Cables will be protected in accordance with the requirements specified in NFPA 8054.2.3.2, 4.2.3.3, or 4.2.3.4 as applicable or, will be re-routed outside the Zone ofInfluence (ZOI) of the Fire Initiator(s) of concern to achieve compliance with NFPA 805Section 4.2.4.2 Performance-Based Approach.

Page 39 of 81 L-PI-15-041 NSPMEnclosure 1Response to RAIs -Fire ModelingFM RAI 01- Fire Modeling Tools and MethodsResponse to FM RAI 01will be provided in separate correspondence by June 26, 2015.FM RAI 02- Fire Modeling Damage CriteriaResponse to FM RAI 02 will be provided in separate correspondence by June 26, 2015.FM RAI 03 -Fire Modeling V&VResponse to FM RAI 03 will be provided in separate correspondence by June 26, 2015.FM RAI 04NFPA 805, Section 2.7.3.3, states that acceptable engineering methods and numerical modelsshall only be used for applications to the extent these methods have been subject to verification and validation.

These engineering methods shall only be applied within the scope, limitations, and assumptions prescribed for that method.In Section 4.7.3 of the LAR, the licensee states that engineering methods and numerical modelsused in support of compliance with 10 CFR 50.48(c) are used and were used appropriately asrequired by Section 2.7.3.3 of NFPA 805.Regarding the limitations of use:[This RAI includes Subparts a, b and c, as shown below along with NSPM responses]

NRC Request (FM RAI 04.a):a. The NRC staff notes that algebraic models cannot be used outside the range ofconditions covered by the experiments on which the model is based. NUREG-1805 includes a section on assumptions and limitations that provides guidance to the user interms of proper and improper application for each FDT.Please explain how it was ensured that algebraic models were not used outside theirlimits of applicability as described in NUREG- 1805.NSPM Response (FM RAI 04.a):a. The application of the algebraic fire models in support of the PINGP Fire PRA wassubjected to the Fire Modeling verification and validation process documented inNUREG-1 824 Volume 1 and NUREG-1934.

This verification and validation processindicates that the fire modeling results generated by the algebraic models were:Page 40 of 81 L-PI-15-041 NSPMEnclosure 1" Used within their limits of applicability as described in NUREG-1 824 or" In cases where the model is outside the applicability limits, sensitivity analyseshave been developed to justify the use of the fire modeling results.Consistent with the verification and validation process described in NUREG-1824 andNUREG-1934, the applicable dimensionless parameters have been determined for thealgebraic fire models and compared with the applicability limits. This process in partensures that the fire modeling analysis has been conducted with approved models, andthe models have either been used within their validated applicability range or theirresults have been determined to be conservative based on appropriate sensitivity analyses.

As indicated in the text of the RAI, the algebraic models described in NUREG-1 805 arequalified by a set of assumptions and limitations that should be considered whenapplying them. In the specific case of the PINGP Fire PRA, the four algebraic modelsused the most are:" Heskestad's fire plume temperature correlation,

• Point source flame radiation model,* Method of Foote, Pagni, and Alvares (FPA) room temperature correlation, and" Method of McCaffrey, Quintiere, and Harkleroad (MQH) room temperature correlation.

The assumptions and limitations associated with these models as described in NUREG-1805 were explicitly considered by the fire modeling analysts when developing theanalysis.

Specific considerations include the following:

" Heskestad's Plume Temperature Correlation:

The assumptions and limitations are described in section 9.5 of NUREG-1805.

The PINGP analysis is consistent with these assumptions because:o It is assumed that energy is released at one point in space (i.e., the pointwhere the fire is located),

o The plume temperature is used to determine fire hazards in the early stagesof the fire (the generation of the initial zone of influence and/or thedetermination of the critical heat release rate for generating damage to thefirst target) before the hot gas layer may have an effect, ando It is only used for diffusion flame applications for ignition sources such aselectrical cabinets as opposed to jet or pre-mixed flames.* Point Source Flame Radiation Model: The assumptions and limitations aredescribed in Section 5.5 of NUREG-1805.

The assumption and limitations indicate that the model offers conservative estimates of the fire hazard (i.e.,thermal radiation).

These are general statements that are addressed specifically in the PINGP analysis through the validation process based on the scenarioconfiguration and input parameters.

An additional assumption indicates that thebase of the fire is circular.

Consistent with this analysis, the base of the fire inthe PINGP analysis is assumed to be circular with an equivalent area of theignition source as applicable.

" MQH and FPA Room Temperature Correlations:

The assumptions andlimitations are listed in Section 2.10 of NUREG-1805.

Page 41 of 81 L-PI-15-041 NSPMEnclosure 1o The assumption associated with the application in conventional sizecompartments and its corresponding shapes are addressed through thevalidation process in the evaluation of the compartment aspect ratio.o The correlation is also used for steady state and transient heat release rateprofiles consistent with the listed assumptions.

o The resulting hot gas layer temperatures are treated as "global" indications ofroom temperatures.

The results are not interpreted as localized to a specificregion of the compartment.

o The correlations are used for determining if the hot gas layer temperatures reach target damage temperatures, which is a temperature range within themodel capabilities and the range of validation in NUREG-1824.

o Wall and corner fire location factors are utilized when applicable whendetermining the hot gas layer temperatures.

o The FPA and MQH models do not require a single heat loss fraction value.Therefore, this parameter is not used.NRC Request FM RAI 04.b):b. It is stated on page J-8 in Table J- 1 of Attachment J of the LAR that the V& V of A/pert'sceiling jet temperature correlation is documented in NUREG-1824.

It is also stated thatthe V& V demonstrates that the ceiling jet correlation is implemented correctly and in allcases provides conservative bounding estimates.

Please provide technical justification for the second statement, because the fact that amodel receives V& V does not prevent its use outside the model's limitations.

NSPM Response (FM RAI 04.b):b. The ceiling jet correlation is not used in the PINGP Fire PRA. Consequently, noverification and validation is necessary for this correlation and it does not need to belisted in Attachment J to the LAR. Additionally, the values for the horizontal component of the zone of influence (ZOI) are at least bounded by the point source radiation model(which is a validated model in NUREG-1824) per guidance in Chapter 8 and Appendix Fof NUREG/CR-6850.

NRC Request (FM RAI 04.c):c. It is stated on page J-9 in Table J- 1 of Attachment J of the LAR that "[FDS] is used withinthe limits of its range of applicability as documented in FPRA-PI-MCR."

It is furtherstated that "For relevant scenarios where the input parameters are outside of the limits,control room abandonment conditions are still predicted."

The two statements appear tobe contradictory.

Please provide technical justification for the application of FDS with input parameters outside the allowable range, and for using the corresponding calculated abandonment times in the Fire PRA.Page 42 of 81 L-PI-15-041 NSPMEnclosure 1NSPM Response (FM RAI 04c):c. The statement on page J-9 in Table J-1 of Attachment J of the LAR that "[FDS] is usedwithin the limits of its range of applicability as documented in FPRA-PI-MCR" isincorrect.

This response will provide the justification for the model inputs.To demonstrate that the FDS model is used within the limits of its range of applicability, the non-dimensional parameters associated with the fire strength, the compartment

geometry, the ventilation, and the flame heights were compared against the applicable ranges provided in Table 2-5 of NUREG-1824, Volume 1. The non-dimensional parameters were calculated for the complete range of fire size bins, from the smallestfires (bin 1) to the largest fires (bin 15). The results, listed in FPRA-PI-MCR, Appendix I,are described below:The fire Froude Number falls within the NUREG-1824 Validation

& Verification (V&V) range for most of the electrical cabinet fires within the corresponding probability distribution for peak heat release rates and only fire size bins 14 and15 for the transient fires.o Electrical Cabinet Fires: The fire size bins that fall outside of the validation range are bins 1 and 12-15. The higher heat release rate bins predictabandonment during the growth period of the ignition source and the heatrelease rate at the abandonment time produces a Froude Number within thevalidation range. The low heat release bin falls below the valid rangebecause the modeling parameters select a fixed fuel area that artificially decreases the Froude Number. In this configuration, the fire plume willentrain more air, resulting in a more rapid smoke layer descent, and acorresponding faster reduction in visibility.

Low Fire Froude Number casesare conservative due to the rapid reduction in visibility when compared to anequivalent case that falls within the validation space.o Transient Fires: Only fire size bins 14 and 15 are within the validation range.The low heat release bins fall below the valid range because the modelingparameter selected uses a fixed fuel area that artificially decreases theFroude Number. In this configuration, the fire plume will entrain more air,resulting in a more rapid smoke layer descent, and a corresponding shorterabandonment time. Low Fire Froude Number cases are conservative due tothe rapid reduction in visibility when compared to an equivalent case that fallswithin the validation space.Also, the selected soot yield of the fire is conservatively chosen to produce moresoot than reported for similar cable materials in the SFPE Handbook andNUREG/CR-7010.

This conservative parameter selection is chosen toconservatively bias the model prediction of abandonment time sufficiently tocounteract the out of range Froude Number.The flame height ratio falls within the NUREG-1824 V&V range for most of thescenarios.

o Electrical Cabinet Fires: Bin 1 falls below the valid range, while bins 10through 15 are above the valid range.o Transient Fires: Bins 1 through 6 fall below the valid range, and no cases areabove the valid range.Page 43 of 81 L-PI-15-041 NSPMEnclosure 1The flame height validation parameter is used primarily to evaluate the exposureof targets within the fire plume [NUREG-1824].

Control room abandonment evaluations assume inherently that occupants will be able to perform their dutiesoutside of the fire plume zone of influence.

Therefore, the flame height ratio isonly important insofar as it may impact the hot gas layer temperature andvisibility.

Fires in which the flame impinges upon the ceiling will produce higherhot gas layer temperatures, and therefore force abandonment sooner than anequivalent case that falls within the validation range. Furthermore, fires in whichthe room is substantially taller than the flame height will allow greaterentrainment into the plume, resulting in more rapid descent of the smoke layer.Also, the selected soot yield of the fire is conservatively chosen to produce moresoot than reported for similar cable materials in the SFPE Handbook andNUREG/CR-7010.

This conservative parameter selection is chosen toconservatively bias the model prediction of abandonment time sufficiently tocounteract the out of range flame height ratio.The equivalence ratio falls within the NUREG-1 824 V&V range for most of thescenarios.

o Electrical Cabinet Fires: Bin 1 falls below the valid range and no cases areabove the valid range.o Transient Fires: Bins 1 through 4 fall below the valid range and no cases areabove the valid range.This parameter does not negatively impact the analysis because the parameter issimply showing that there is sufficient oxygen available for the fire to burn withinthe MCR volume.* The enclosure ratio falls within the NUREG-1824 V&V range for most of the.scenarios.

o The enclosure ratios are within the NUREG-1 824 V&V range for the electrical cabinets outside of the horseshoe.

The enclosure ratio for the electrical cabinets inside of the horseshoe exceeds the NUREG-1 824 V&V range. Theelectrical cabinet dimensions that are outside of the valid range are due to theextreme elevation of the fuel source, relative to the ceiling height. Asdiscussed for the Flame Height Ratio parameter above, the effect of anelevated fuel source is considered to produce consistent or conservative results when compared to an equivalent case within the valid range.o The enclosure ratios are within the NUREG-1 824 V&V range for all transient fires.In addition to the conservative fire modeling parameter selections described above, the probability of abandonment calculation identifies the mostconservative time to abandonment for all fires in the FDS simulations.

Theshortest time to abandonment is used for all fire scenarios.

This is a boundingconservative treatment of the time to abandonment and it will bound the riskassociated with any of the above stated V&V parameters that were identified asout of range.Page 44 of 81 L-PI-15-041 NSPMEnclosure 1FM RAI 05NFPA 805, Section 2.7.3.4 states that personnel who use and apply engineering analysis andnumerical methods shall be competent in that field and experienced in the application of thesemethods as they relate to nuclear power plants, nuclear power plant fire protection, and powerplant operations.

In Section 4.5.1.2 of the LAR, the licensee states that fire modeling was performed as part ofthe Fire PRA development (NFPA 805, Section 4.2.4.2).

The NRC staff notes that this requiresthat qualified fire modeling and PRA personnel work together.

Furthermore, in Section 4.7.3 ofthe LAR, the licensee states that "For personnel performing fire modeling or Fire PRAdevelopment and evaluation, NSPM will develop and maintain qualification requirements forindividuals assigned various tasks. Position Specific Guides will be developed to identify anddocument required training and mentoring to ensure individuals are appropriately qualified perthe requirements of NFPA 805 Section 2.7.3.4 to perform assigned work."Regarding qualifications of users of engineering analyses and numerical models (i.e., firemodeling techniques):

[This RAI includes Subparts a, b and c, as shown below along with NSPM responses]

NRC Request (FM RAI 05.a):a. Please describe the requirements to qualify personnel for performing fire modelingcalculations in the NFPA 805 transition.

NSPM Response (FM RAI 05.a):a. Fire Modeling performed to support the development of the NFPA 805 LAR wascompleted by qualified contractors.

The vendor provided credentials for the fire modelinganalysts.

Credentials were reviewed to ensure analysts were appropriately qualified perthe requirements of NFPA 805, Section 2.7.3.4 which included education in fireprotection engineering and fire modeling, and extensive experience performing firemodeling studies.

NSPM reviewed the vendor's credentials of the analysts performing the fire modeling tasks and ensured that each task was performed by analysts withappropriate training in the fire modeling area being performed.

During the transition process post NFPA 805 LAR submittal, NSPM will continue to require credentials toensure analysts are knowledgeable in fire modeling techniques, including interpreting and maintaining fire modeling software.

NRC Request (FM RAI 05.b):b. Please describe the process for ensuring that fire modeling personnel have theappropriate qualifications, not only before the transition but also during and following thetransition.

Page 45 of 81 L-PI-15-041 NSPMEnclosure 1NSPM Response (FM RAI 05.b):b. Fire Modeling performed for supporting the development of the NFPA 805 LAR wascompleted by qualified contractors.

The vendor provided the fire modelers' credentials.

Credentials included education in fire protection engineering and fire modeling, andextensive experience performing fire modeling studies.

NSPM reviewed the vendor'scredentials of the analysts performing the fire modeling tasks and ensured that each taskwas performed by analysts with appropriate training in the fire modeling area beingperformed.

During the transition process post NFPA 805 LAR submittal, NSPM willcontinue to require credentials to ensure analysts are knowledgeable in fire modelingtechniques, including interpreting and maintaining fire modeling software.

Once the NFPA 805 transition process is completed, fire modeling calculations will beperformed by a Fire Protection or PRA Engineer who meets the qualification requirements of Section 2.7.3.4 of NFPA 805. This will be ensured through qualification requirements and training that will be developed as described in Table S-3,Implementation Item #26. An analyst will be required to complete this qualification before modeling in support of the Fire PRA or a qualified person will need to review andsign off the prepared material before its use within the Fire PRA. Qualifications aretracked through NSPM's training program and are procedurally required to be checkedprior to completing the task that requires fire modeling.

The NSPM fire modelingqualification requires an analyst to be qualified in PRA before that analyst can modelfires for their input into the Fire PRA.NRC Request (FM RAI 05.c):c. When fire modeling is performed in support of the Fire PRA, please describe how propercommunication between the fire modeling and Fire PRA personnel is ensured.NSPM Response (FM RAI 05.c):c. During the development phase of the Fire PRA, Fire Protection Engineers (FPE) whoconducted the fire modeling and the PRA engineers maintained frequentcommunications.

Specifically, the fire modeling personnel populated the databases orspreadsheets in which all the relevant fire modeling inputs are maintained.

The firescenario frequencies generated by these tools are electronically transmitted to the PRAengineers who perform the risk quantification.

Both the FPEs and the PRA engineers participated in the cutset review meetings during the development of the Fire PRA.The NSPM qualification for Fire Modeling precludes the use of Fire Modeling in the FirePRA unless it is reviewed by a qualified Fire PRA Engineer.

This ensurescommunication between Fire Modeling personnel and Fire PRA personnel.

Page 46 of 81 L-PI-15-041 NSPMEnclosure 1FM RAI 06NFPA 805, Section 2.73.5 states that an uncertainty analysis shall be performed to providereasonable assurance that the performance criteria have been met.In Section 4.7.3 of the LAR, the licensee states that "Uncertainty analyses were performed asrequired by 2.7.3.5 of NFPA 805 and the results were considered in the context of theapplication.

This is of particular interest in fire modeling and Fire PRA development."

Regarding the uncertainty analysis for fire modeling:

[This RAI includes Subparts a and b, as shown below along with NSPM responses]

NRC Request (FM RAI 06.a):a. Please describe how the uncertainty associated with the fire model input parameters was accounted for in the fire modeling analyses.

NSPM Response (FM RAI 06.a):a. The uncertainty associated with the fire model input parameters was accounted for byusing conservative input parameters and varying input parameters in sensitivity cases,as described below:* The input parameter that has the most impact on results is the heat release rate.The uncertainty is accounted for by using bounding heat release rates. Heatrelease rates are selected to be the screening values (98th percentiles) of thedistributions recommended in NUREG/CR-6850.

For the case of cable fires assecondary combustibles, the guidance in NUREG/CR-7010 associated withmodeling cable fires (i.e., the heat release rate per unit area recommended forthe FLASH-CAT model) was utilized.

" The analysis conservatively assigns the lowest radiant heat flux damagethreshold and damage temperature threshold, 6 kW/m2 and 205 0C, suggested inAppendix H of NUREG/CR-6850.

The use of these damage criteria bounds thefire impacts expected for raceways containing a mixture of cables with varyinginsulation types.* Sensitivity cases are performed when normalized parameters are outside of thevalidation range. Sensitivity evaluations were run for room aspect ratio, ventsizes, and fire height versus ceiling height.NRC Request (FM RAI 06.b):b. Please describe how the "model" and "completeness" uncertainty was accounted for inthe fire modeling analyses.

NSPM Response (FM RAI 06.b):b. The fire modeling verification and validation analysis was completed as part of thedevelopment of the Fire PRA. This analysis covers all the fire modeling, including CFAST and hand calculations, used in the PINGP Fire PRA. The verification andPage 47 of 81 L-PI-15-041 NSPMEnclosure 1validation analysis determines whether models are used within their validated range. Ifthe models are found to be used outside of the range, then the input parameters arevaried in a conservative direction (i.e., more challenging fire conditions) and the revisedmodel results are used as input to the Fire PRA; i.e., the PINGP fire modeling analysisresults are applied considering the uncertainty associated with the model validation range. It should be noted that for fire models used outside the range of applicability, sensitivity cases are run, as suggested in NUREG-1934.

The uncertainty sources and treatments associated with fire scenario development anddetailed fire modeling for single compartment fire scenarios in the PINGP Fire PRAinclude uncertainties related to the selection of transient zones, fire location, fire growthand propagation, activation and function of the detection and suppression system, theselection of damage criterion, conduit routing, selection of fire models, and the inputs tothe chosen fire models. A summary of the uncertainty sources and treatments associated with the fire scenario development and detailed fire modeling for singlecompartment fire scenarios in the PINGP Fire PRA is provided next:1. Uncertainty Associated with the Scenario Development Processa. Selection of Transient Zones: Transient zones are postulated so that most of theopen floor area is captured in the analysis.

These are areas where a transient firecould be postulated.

There is however uncertainty in the selected sizes andboundaries of the transient zones. As a conservative

practice, and whenpractically

possible, the transient zones have been selected large enough tocapture target damage beyond a typical zone of influence of an ignition source.In addition, a target overlap between transient zones has been incorporated inthe analysis to account for targets near the boundaries.

That is, the targetslocated nearby the boundaries between transient zones have been mapped to alltransient zones adjacent to that boundary.

When applicable, fire propagation between transient zones is included in the scenario progression.

b. Fire Location (All fire zones except the Main Control Room): The location of thefire is unavoidably a source of uncertainty as all fires in the Fire PRA areassumed to occur at a specific point within the fire zone. This fire locationimpacts the heating process of nearby targets.

This source of uncertainty isaddressed in the Fire PRA in a consistent approach.

That is, the guidance forassigning fire location has been consistently applied throughout the analysis.

Theguidance is based on a conservative practice of assigning fixed and transient firelocations consistently for all scenarios.

For fixed sources, the fires have beenpostulated at the elevation of the ignition source. In the specific case of electrical panels, the fires are postulated 1' below the top of the cabinet as clarified inNUREG/CR-6850 Supplement 1 (Chapter 12). In the case of general transients or transient fires due to hotwork, the base of the fire has been postulated 2' fromthe floor. Since the transient fires are due to items brought temporarily into thefire zone, there is uncertainty associated with the fire elevation.

The 2' elevation is a necessary practical assumption to account for combustibles that may not belocated at floor level and representative of typical equipment carts, for example,that are brought into fire zones. Oil spill fires are postulated at floor level.c. Fire Growth and Propagation:

The uncertainties associated with heat releaserates, fire growth profiles, fire propagation and determination of time to targetdamage are documented in a number of industry documents and are notpractical to list in this report. For example, NUREG-1805, NUREG-1824, andNUREG/CR-6850 list and in some cases quantify these uncertainties.

For thePage 48 of 81 L-PI-15-041 NSPMEnclosure 1PINGP Fire PRA, generic industry guidance on fire growth rates has beenutilized.

For fire propagation, the Zone of Influence used to map targets to eachignition source conservatively bounds the damage produced by a fire reachingthe 98th percentile HRR for that source. As such, damage to cable trays andconduits as the result of a fire propagating through secondary combustibles isaccounted for.2. Uncertainties Associated with Detection and Suppression

a. Activation:

The activation times for detection and suppression systems is asource of uncertainty in the Fire PRA primarily because of the complexity of theconfigurations encountered in the scenarios postulated in the different zones inthe power plant. Examples of these complex configurations include deviceslocated in heavily obstructed

ceilings, sprinklers within cable trays, etc. As aconservative

practice, detection and suppression systems are modeled such thatapplicable targets are damaged prior to activation.

b. Suppression:

The ability to control or completely suppress the fire as a functionof time using the different suppression means available in a given scenario is asource of uncertainty.

In the Fire PRA, detection and suppression is treated withan event tree approach where both outcomes, successful and failed suppression activities, are modeled.

In addition, suppression times are conservatively assessed based on selected input parameters to the fire models and the use ofvalidated models.3. Damage Criterion The damage criterion is a source of uncertainty that is considered in the analysis.

The generic guidance for damage criteria suggests point estimates for damagethresholds, which are used in this study. These generic point values are based onexperimental observations yielding a range of values associated with cable damage.Such uncertainty is captured in a point value that for the most part is expected to bebounding (see Appendix H of NUREG/CR-6850 and Appendix A of NUREG-1805).

For the PINGP Fire PRA, damage thresholds associated with Thermoplastic cablesare assumed throughout the analysis.

4. Conduit RoutingConduit routing is a source of uncertainty in the Fire PRA. The conduit locations arepartially available on drawings and are also partially labeled in the field.Consequently, routing of conduits has been evaluated on an individual basis basedon conduit location and risk significance.

Conduits which could not be located in theplant were failed for every applicable scenario.

5. Fire Modeling Selection For the PINGP Fire PRA, CFAST and hand calculations have been utilized.

Thesemodels have been subjected to verification and validation studies for selectedscenarios as described in NUREG-1824, where model uncertainty ranges have beendocumented.

6. Fire Modeling InputsThe uncertainty intervals associated with the input parameters include:a. Heat Release RatesPage 49 of 81 L-PI-15-041 NSPMEnclosure 1i. Fixed Ignition Sources:

The gamma distributions of the heat release ratevalues for most fixed ignition source types are given in Appendix E ofNUREG/CR-6850.

ii. Oil Fires: In the case of large oil fires, a two-step approach was used: Step 1)Large oil fires were assumed to cause a hot gas layer without a SeverityFactor (SF) = 1.0. Step 2). If this approach was too conservative, then theamount of oil contained within the pump was determined and the approachdocumented.

The uncertainty associated with the heat release rate intensity and duration of oil fires is mostly governed by the postulated size of the spilland the amount of oil. The FPRA assumes bounding heat release rates for oilspills to account for the uncertainty associated with the amount of oil in theignition source (i.e., pumps). At the same time, the frequency of oil fires iscalculated using two probabilistic parameters:

1) The probability that theignition source fire is associated with oil (split fraction from Table 6-1 ofNUREG/CR-6850 and 2) The Severity factor for small and large oil firesdescribed in Appendix E of NUREG/CR-6850.

These inputs are treated asuncertain parameters in the risk equation.

b. Cable fire modeling parameters:

i. Tray width is assumed to be 18" for all heat release rate contribution calculations.

ii. Tray heat release rate per unit width is taken from Table R-1 of NUREG/CR-6850 in comparison to the value provided in Section 9.2.2 of NUREG/CR-7010.iii. The heat of combustion per unit weight of cable insulation was selected to be16 MJ/kg for cables in general as recommended in Section 9.2.2 ofNUREG/CR-7010.

iv. The number of cables per tray was assumed such that the calculated heatrelease rate would bound the typical arrangement found at PINGP.c. Other Fire Modeling Inputs: The other fire modeling inputs include compartment geometry and ventilation characteristics.

The compartment geometry andventilation characteristics are obtained from plant drawings.

In summary, the fire modeling analysis in the Fire PRA has been conducted consistent with the industry guidance and practices including fire modeling verification andvalidation.

Consistent with the guidance, uncertainties associated with the fire modelingparameters are reflected in the risk quantification as follows:" Severity factor, which is calculated using a critical heat release rate value asdescribed earlier in this response.

In addition to the conservative determination of critical heat release rate values, the uncertainty associated with the severityfactor is explicitly modeled in the uncertainty task of the Fire PRA." Non suppression probabilities, which are calculated using the time to damageresulting from fire modeling analysis and the conditional core damageprobability/conditional large early release probability.

In addition to theconservative determination of non-suppression probability values, the uncertainty associated with the severity factor is explicitly modeled in the uncertainty task ofthe Fire PRA.Page 50 of 81 L-PI-15-041 NSPMEnclosure 1Conditional core damage probability/conditional large early release probability, which is calculated based on the targets associated with each fire scenario.

Asdiscussed earlier in this response, mapping of targets to the different firescenarios follows a conservative process to ensure that the resulting probabilities are bounding.

Completeness associated with fire models is addressed in the PINGP Fire PRA withinthe overall quantification

process, as the PRA is an integrated analysis.

Fire Modelingprovides inputs to a broad comprehensive Fire PRA which includes modeling ofelectrical

systems, operator

actions, and the plant systems and components needed toshutdown the plant. One of the first steps in the fire modeling process is to identify thefire scenarios that will be analyzed.

In some situations, the scenario analysis invokesfire modeling capabilities that are not currently available, generating the completeness uncertainty situation described in the question.

When the fire modeling does not providea full answer or an answer with sufficient resolution, the scenario definition and targetmapping within the Fire PRA conservatively compensates for the lack of information.

The Fire PRA allows the analyst to conservatively compensate for the lack of firemodeling capabilities outside the fire modeling analysis so that the scenario is properlymodeled in the Fire PRA. Some examples are listed below:* The determination of time to automatic suppression.

Available detection modelsare not fully applicable to many of the postulated scenarios; therefore, as part ofthe scenario definition, targets are failed intentionally before the automatic suppression is credited.

" Oil spill fires are difficult to analyze; therefore, full fire zones or transient zoneshave been assumed failed due to oil fires.* Both zones of multi-compartment combinations are failed conservatively whenfire modeling propagation calculations from one compartment to another are notconducted.

" Full main control board panels are failed due to.the lack of analytical firemodeling

methods, with appropriate verification and validation

studies, to predictflame propagation within a panel.The four examples above illustrate how the completeness uncertainty associated withfire modeling calculations is addressed "outside of the fire modeling" by conservatively failing targets in the fire scenarios so that the risk contribution is bounding.

Page 51 of 81 L-PI-15-041 NSPMEnclosure 1RAI Responses

-Radioactive ReleaseRadioactive Release RAI 01NFPA 805, Section 1.5.2 states that "Radiation release to any unrestricted area due to the directeffects of fire suppression activities (but not involving fuel damage) shall be as low asreasonably achievable and shall not exceed applicable 10 CFR, Part 20, limits."Attachment E of the LAR does not identify use of outside yard areas to store radioactive materials.

Please discuss if there are any outside yard areas where radioactive materials are stored (e.g.in sealand type containers.)

Since outside yard areas are open to the atmosphere, if radioactive material is stored outside, provide an analysis for a fire occurring in outside yard areas thatdemonstrates that the gaseous and liquid effluent releases will result in doses that are less thanthe 10 CFR 20 annual dose limits to a member of the public.NSPM Response (RR RAI 01):Although not identified in Attachment E of the LAR, radioactive materials are stored in anoutside yard area and NSPM has performed an analysis that demonstrates that the gaseousand liquid effluent releases due to a fire and firefighting activities will result in doses that are lessthan the 10 CFR 20 annual dose limits to a member of the public.There is a fenced off area west of the switchyard that contains thirteen (13) 20 foot Sealandcontainers and eight (8) other Low Level Radioactive Waste (LLRW) containers.

There is a totalactivity of approximately 20 milli-curies total in all of the containers.

(This is a typical inventory and activity level of material stored in this area.)In addition, the spare Reactor Coolant Pump (RCP) motor that was previously located on theTurbine Operating Deck in Fire Area 8 has been relocated to the newly constructed Distribution Center Warehouse in the Owner Controlled Area (OCA). The spare RCP motor activity is5.7E-1 milli-curies.

Technical Basis Document

1. 12-002, Dose Due to a DAW Trailer Fire (Reference R-1), is acalculation used to verify that offsite dose and site boundary liquid radionuclide concentrations from effluents due to a trailer fire at the plant site are below Technical Specification limits. ThisDAW trailer fire analysis also bounds releases due to a fire in the Distribution CenterWarehouse.

The methodology and results of the calculation are summarized below.Methodoloqgy The calculation is based on a fire scenario involving a 20 foot Sealand van of dry active waste(DAW). The total activity in the Sealand van of DAW is approximately 200 milli-curies, based onthe highest activity DAW shipment in the prior five (5) years. This bounds the total activity of 20milli-curies in the fenced off area west of the switchyard and the total activity of 5.7E-1 milli-curies for the spare RCP motor in the Distribution Center Warehouse.

Water used in firefighting activities and hose stream application will transport the total activityinto the plant discharge canal and then into the Mississippi River. A member of the public isassumed to drink two liters at a point downstream of the plant discharge.

The analysis assumesPage 52 of 81 L-PI-15-041 NSPMEnclosure 1that activity from the fire is diluted by the discharge canal flow rate and the Mississippi River lowflow rate; no credit is taken for dilution from water used in firefighting or hose stream activities.

Even though there is no water intake for 300 miles downriver, dilution by the Mississippi flow atthe discharge point is a valid assumption and drinking the water just after this dilution occurs isconservative.

Gaseous effluent from this fire includes the total trailer radioactivity that becomes airborne oversix hours. This airborne cloud is diluted using the X/Q for the Design Basis Accident exclusion area boundary and a member of the public is exposed to the cloud for six hours using thestandard man breathing rate.For both the liquid and gaseous releases, the dose conversion factors were the most limitingfrom US EPA Federal Guidance Report No. 11, Limiting Values of Radionuclide Intake and AirConcentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion.

ResultsThe total activity analyzed for the fire scenario involving a Sealand van of DAW is approximately 200 milli-curies.

The total activity is 20 milli-curies in the fenced off area west of the switchyard inclusive of all Sealand and other LLRW containers.

As such, the DAW trailer fire addressed inthe Technical Basis Document bounds any single or combination of Sealand containers andboxes in the area. The total analyzed activity in the Sealand van of DAW also bounds the totalactivity of 5.7E-1 milli-curies for the spare RCP motor in the Distribution Center Warehouse.

The PINGP Offsite Dose Calculation Manual (ODCM, Reference R-2) adopted the limits of 10CFR 50 Appendix I, Numerical Guides for Design Objectives and Limiting Conditions forOperation to Meet the Criterion "As Low As Reasonably Achievable" for Radioactive Material inLight-Water-Cooled Nuclear Power Reactor Effluents.

As such, the ODCM and 10 CFR 50Appendix I limits are the same. The ODCM and 10 CFR 50 Appendix I limits are morerestrictive than the 10 CFR 20 limits.The liquid release dose of 0.0065 mrem whole body does not exceed the ODCM and 10 CFR50 Appendix I limit of 3 mrem/quarter (two unit site).The gaseous release dose of 0.45 mrem whole body and 5.7 mrem lung dose (maximum organ)does not exceed the ODCM and 10 CFR 50 Appendix I limit of 15 mrem/quarter (two unit site).Based on the results of the Technical Basis Document calculation, the DAW trailer fire scenariodoes not exceed ODCM, 10 CFR 50 Appendix I, and 10 CFR 20 offsite dose limits.

References:

R-1 PINGP Technical Basis Document 12-002, Dose Due to a DAW Trailer Fire,Revision 0, November 23, 2012.R-2 PINGP Offsite Dose Calculation Manual (ODCM).Page 53 of 81 L-PI-1 5-041 NSPMEnclosure 1Radioactive Release RAI 02NFPA 805, Section 1.5.2 states that "Radiation release to any unrestricted area due to the directeffects of fire suppression activities (but not involving fuel damage) shall be as low asreasonably achievable and shall not exceed applicable 10 CFR, Part 20, limits."Attachment E of the LAR identified Fire Area 4, the Fuel Handling Area, and Fire Area 61A, theAuxiliary Building Hatch Area, as not having ventilation where the potential transfer ofcontaminated smoke to the exterior can occur. Attachment E indicates that revised firestrategies will incorporate mitigating actions to monitor and filter potentially contaminated smokebased on radiological conditions identified during the conduct of firefighting activities.

Please provide information on how this will be accomplished.

NSPM Response (RR RAI 02):For fire areas without ventilation controls such as Fire Areas 4 and 61A, NSPM will implement actions to monitor and filter potentially contaminated smoke, as stated in Attachment E of theLAR, Based on radiological conditions, these actions may include opening doors to adjacentareas with filtered ventilation (see LAR Table S-3 Item 13), or use of portable HEPA ventilation units, as will be addressed in new Table S-3 Item 65, as described below. These actions arefurther described below.Fire Areas With No Ventilation ControlsFire Area 4 runs from the 695 ft elevation to the roof of the 755 ft elevation.

Fire Area 4 is opento Fire Area 61A on the 755 ft elevation by the open space over the top of Fire Area 61, theSpent Fuel Pool Area, which is a part-height enclosure on the 755 ft elevation.

During the development of Attachment E for the LAR supplement, a modification wasimplemented that abandoned the fan and failed closed the damper for the Laundry Fan ExhaustFilter in Fire Area 61, the Auxiliary Building Anti-C Clothing Area on the 735 ft elevation.

Themodification removed the ability to locally operate the Laundry Room Filters and Exhaust Fansto prevent gaseous effluent from escaping Fire Area 61. Fire Area 61 is not provided with anyother form of ventilation controls and, as such, is addressed in this response as a non-ventilated area along with Fire Area 4 and Fire Area 61A.Fire Area 4 and Fire Area 61 A are part of the Common Area of the Auxiliary Building (CAAB)that is not provided with ventilation controls.

Fire Area 61 is located directly below Fire Area61A and an open stairway connects the two fire areas. Fire Area 61 is also part of the CAABthat is not provided with ventilation controls.

Credit is taken for the large volume of the CAAB tocontain radioactive gaseous effluent from the CAAB. Mitigating actions will be based onradiological conditions as monitored by radiation protection and communicated to the firebrigade leader during fire events.Miti-Qatinci ActionsMitigating actions could include opening the doors to Fire Area 62, Spent Fuel Pool Area, todirect potentially contaminated smoke from a fire in a CAAB fire area to the Spent Fuel Poolnormal ventilation system. The Spent Fuel Pool Normal Ventilation System filters air throughroughing and HEPA filters to the spent fuel pool normal ventilation exhaust stack prior toPage 54 of 81 L-PI-15-041 NSPMEnclosure 1release.

The revised fire strategies and associated training materials described inImplementation Items 6 through 13 in LAR Table S-3 will provide procedural guidance forresponse to such events.Mitigating actions could also include the use of portable HEPA filters for use in any fire areahaving the potential for generation of radiological gaseous effluents, not just in the identified areas without ventilation controls.

Portable HEPA filters will be strategically located in theRadiologically Controlled Area (RCA) and will be available for use based on radiological conditions as monitored by radiation protection personnel and communicated to the fire brigadeleader during fire events. Table S-3, Item 65, will be added to provide these portable HEPAfilters as follows:Item 65:Provide portable HEPA filters strategically located in the Radiologically Controlled Area(RCA) that will be available for use based on radiological conditions as monitored byradiation protection personnel and as communicated to the fire brigade leader during fireevents.Radioactive Release RAI 03NFPA 805, Section 1.5.2 states that "Radiation release to any unrestricted area due to the directeffects of fire suppression activities (but not involving fuel damage) shall be as low asreasonably achievable and shall not exceed applicable 10 CFR, Part 20, limits."Attachment E of the LAR identified Fire Area 93, the Low Level Rad Waste Area, as not havingventilation where the potential transfer of contaminated smoke to the exterior can occur.Attachment E states that a combination of containerization and administrative controls will limitthe amount of exposed contaminated combustible materials, and that revised fire strategies willincorporate mitigating actions to filter potentially contaminated smoke.Please describe what administrative controls (e.g. limits on activity, etc.) and types of containers will be used in this area that will be used to meet the applicable 10 CFR 20 requirements?

Furthermore, Table S-3, Implementation Item 15, provides a container with booms, portablefiltered ventilation and other appropriate equipment to contain effluent releases in the Low LevelRad Waste Area.Please clarify if this container

[will] be staged in this building to be used in case of a fire?NSPM Response (RR RAI 03):Background Fire Area 93, the Low Level Rad Waste Area, is used for storage of plant equipment, radioactive trash, spent resin, and spent filters.

The old upper reactor internals from units 1 and 2 arestored in this building.

There is no processing of waste in the building.

The main open area ofFire Area 93 does not contain externally contaminated containers.

The old upper reactorinternals are stored behind a part-height concrete shield wall along the south wall of the buildingand are enclosed in a carbon steel container.

The main open area does not present radiological Page 55 of 81 L-PI-15-041 NSPMEnclosure 1release concerns.

The truck loading area in the southeast corner of the building is used tostage potentially contaminated waste and personal protective clothing (radioactive trash) prior toloading the material into a Sealand van. The staging area is designated by ropes and is smallerin size than a 20 foot Sealand van. Administrative controls and containerization described below apply only to this location in the truck loading area.Administrative Controls and Types of Containers Administrative controls and containerization will be applied to limit the exposed storage ofpotentially contaminated waste and personal protective clothing resulting primarily duringoutages.

A portion of the truck loading area is typically roped off and used to store plastic bagsof exposed potentially contaminated waste and personal protective clothing prior to loading ofthe material into a Sealand van. Administrative controls identified in Table S-3 Item 16 will beimplemented to minimize the amount of such exposed materials by requiring that the plasticbags be placed in metal containers prior to being loaded instead of being left exposed in thetruck loading area. The size of the metal containers will be dependent on the amount ofexpected materials, taking into account the need for accessing the containers for loading of thematerials.

Technical Basis Document

1. 12-002 Rev. 0, Dose Due to a DAW Trailer Fire, dated 11/23/2012 is used to verify that offsite dose and site boundary liquid radionuclide concentrations fromeffluents due to a trailer fire at the plant site are below Technical Specification limits. Thecalculation is based on a fire scenario involving a 20 foot Sealand van of dry active waste(DAW) based on the highest activity DAW shipment in the prior five (5) years. Based on theresults of the Technical Basis Document calculation, the DAW trailer fire scenario does notexceed ODCM offsite dose limits. Refer to the response to Radiological Release RAI 01 fordetails associated with the Technical Basis Document calculation.

The DAW trailer fire scenario in the Technical Basis Document bounds the maximum size ofmetal containers that will be used to contain outage-related plastic bags of potentially contaminated waste and personal protective clothing.

The DAW trailer fire scenario alsobounds the maximum activity of DAW in the metal containers that will be used to containoutage-related plastic bags of potentially contaminated waste and personal protective clothing.

The container described in Table S-3, Implementation Item 15, with booms, portable filteredventilation and other appropriate equipment to contain effluent releases in the Low Level RadWaste Area, will be stored in a location adjacent to, but not within, the Low Level Rad WasteArea. The specific location will be determined as part of implementation, and will be in thenormal travel path to the Low Level Rad Waste Area.Page 56 of 81 L-PI-15-041 NSPMEnclosure 1RAI Responses

-Probabilistic Risk Assessment (PRA)PRA RAI 01 -Fire Event Facts and Observations Section 2.4.3.3 of NFPA 805 states that the probabilistic safety assessment (PSA) (PSA is alsoreferred to as PRA) approach,

methods, and data shall be acceptable to the authority havingjurisdiction (AHJ), which is the NRC. The RG 1.205 identifies NUREG/CR-6850 asdocumenting a methodology for conducting a fire PRA (FPRA) and endorses, with exceptions and clarifications, NEI 04-02, Revision 2, as providing methods acceptable to the NRC staff foradopting a fire protection program consistent with NFPA 805. The RG 1.200 describes a peerreview process utilizing an associated ASME/ANS standard (currently ASME/ANS-RA-Sa-2009) as one acceptable approach for determining the technical adequacy of the PRA onceacceptable consensus approaches or models have been established for evaluations that couldinfluence the regulatory decision.

The primary result of a peer review are the facts andobservations (F&Os) recorded by the peer review and the subsequent resolution of these F&Os.Please clarify the following dispositions to fire F&Os and Supporting Requirement (SR)assessments identified in Attachment V of the LAR that have the potential to impact the FPRAresults and do not appear to be fully resolved:

[This RAI includes Subparts a through h, as shown below along with NSPM responses]

NRC Request (PRA RAI 01.a):a. ES-Cl-01:

This F&O cites incomplete treatment of instrumentation needed to supportfire response operator actions;

however, Appendix D of FPRA-PI-ES, Revision 1,appears to only address instrumentation required for credited internal events actions.Please describe how fire-induced instrument failure is addressed by the FPRA humanreliability analysis (HRA) for both internal events and fire response operator actions.Include a description of how instrumentation that is relied on for credited operatoractions was identified and verified as available to a level of detail commensurate with therisk importance and quantification of human error probabilities (HEPs).NSPM Response (PRA RAI 01.a):a. Fire-induced instrument failures have been addressed by the Fire PRA human reliability analysis.

The quantification process fails the instrument cables that are mapped to thedifferent fire scenarios within the plant. Consequently, the instruments are failed in themodel, which influences the quantification of CCDP and CLERP values. Key elementsof the quantification process are as follows:* If the instrument indications necessary to provide the cue for a given HFE arefailed, the fault tree logic causes the associated human action to beunsuccessful.

• In some situations where only partial instrumentation indications are available, but still sufficient to provide a cue for needed operator action, a specific HFE wasdeveloped that accounts for the adverse effect of degraded instrumentation in thehuman error probability (HEP) (as opposed to the lower HEP associated with fullinstrumentation).

Page 57 of 81 L-PI-15-041 NSPMEnclosure 1" For some HFEs, the relied upon instrumentation is sufficiently redundant anddiverse to ensure that a cue will be available for operator action, thus only theHFE is modeled in the fault tree." For some HFEs, the action is directed procedurally such that instrumentation-supplied cues are not required for success of the action. In this situation, noinstrumentation is modeled in the FPRA, only the HFE is.The table in Appendix D to the Equipment Selection

notebook, FPRA-PI-ES, wasupdated in Revision 2 to include all instrumentation required for Human Failure Eventscredited in the Fire PRA.NRC Request (PRA RAI O1.b):b. CS-A 10-01: The disposition to this F&O states that cables are routed by fire area in thecable database, and according to FPRA-PI-SCA, Revision 1, this database lacks uniqueconduit identifiers.

As a result, please clarify the FPRA's treatment of conduits with unknown routing andinclude justification of the process used to map such conduits to fire compartments.

NSPM Response (PRA RAI 01.b):b. The PINGP Fire PRA uses an export from SAFE GENESIS (the Fire PRA cabledatabase) for cable routing to map targets.

The SAFE GENESIS model relates cable toequipment, cable to raceway, and raceway to location for analysis purposes.

The exportfrom the cable routing database does not include unique conduit identifiers for each firearea. For example, "400-CND" identifies only that a 4 inch conduit is routed through afire compartment.

To facilitate conduit mapping to fire scenarios, the generic cablerouting information from SAFE GENESIS is manipulated to map specific cable/conduit combinations to specific fire areas. The cable routing sequence for each cable is studiedto evaluate the location of the cable through different raceways (cable trays andconduits).

The conduit location in the routing is established by comparing the location ofthe known cable trays in the route sequence to the known fire area locations of the cableitself and determining the conduit to fire area relationship based on that routing.

Theconduit generic identification is concatenated with the cable name.For example, from the "cable to raceway" export from SAFE GENESIS, cable 111C-4has five sequence points on its route:Page 58 of 81 L-PI-15-041 Enclosure 1NSPMTable 1: SAFE GENESIS CableRoute for 111 C-4CABLE RACEWAYSEQ ID111C-4 400-CND 1111C-4 1SG-LA29 2111C-4 1SR-LA2 3111C-4 1SM-LA4 4111C-4 2SM-LA27 5The first route point in the cable sequence is conduit 400-CND (First SEQ-ID in Table Q-1 above). The last four sequence route points are routed through cable trays. From the"cable to fire area" export from SAFE GENESIS, cable 111C-4 is located in 5 fire areas:Table 2: 111C-4 to Fire AreaCable IDFire Area111 C-4 29111 C-4 32111 C-4 37111 C-4 80111C-4 81From the "raceway to fire area" export from SAFE GENESIS, the four cable trays haveknown locations in fire areas 32, 80, and 81. Therefore, conduit 400-CND-1 11C-4(Concatenation of Conduit ID and Cable ID) is mapped to the "remaining" fire areas, 29and 37. Further, the conduit is also conservatively mapped to fire area 32 such that thepath from the conduit to the Cable Tray in 1 SG-LA29 (Route Point #2) is captured; it isassumed that the cable could run through conduit in that fire area before it reaches cabletray 1 SG-LA29.

Similarly, if a conduit falls in the middle of a route sequence, the conduitis mapped to the fire area of the cable tray in sequence before and after the conduit.The approach described above provides a high level of confidence that conduitscontaining Fire PRA target cables are mapped to the appropriate fire compartments.

Forcases where additional refinement was needed for detailed fire modeling, additional walkdowns were performed to identify conduits and cable trays within the zone ofinfluence.

NRC Request (PRA RAI 01.c):c. PRM-A 1-02: The disposition to this F&O indicates that due to the presence of pipingwith soldered joints, the instrument air system is only credited for a limited number of firescenarios within the Relay Room (Fire Area 18).Page 59 of 81 L-PI-15-041 NSPMEnclosure 1Please justify the criteria (e.g., damage threshold, system response, etc.) used todetermine those fire scenarios that do not lead to failure of the instrument air system.NSPM Response (PRA RAI 01.c):c. The instrument air system is not currently credited within the Fire PRA, and all scenarios result in the failure of the instrument air system.NRC Request (PRA RAI 01.d):d. FSS-B2-01:

The licensee's analysis (Section 6.0 of FPRA-PI-MCR) indicates that thereare "a large number of cable raceways, particularly cable trays, "located under the raisedfloor within the main control room (MCR); however, it appears that these raceways areexcluded from the MCR scenario development, both as ignition sources (i.e., self-ignited cable fires) and potential targets of other ignition sources (i.e., transient fires, transient fires due to welding and cutting, and cable fires due to welding and cutting).

Please justify this exclusion.

NSPM Response (PRA RAI O1.d):d. There are five types of ignition sources that could be postulated as fire scenario initiators under the raised floor of the Main Control Room. These ignition sources are: 1) self-ignited cable fires, 2) junction boxes, 3) transient fires, 4) transient fires due to hotwork,and 5) cable fires due to hotwork.The identification of cable type and linear feet was done on a cable-by-cable basis, withthe total linear feet of each cable type also identified.

The thermoset cable length isremoved since it cannot self-ignite.

Subsequently, a series of filters were applied to thethermoplastic and cables with unknown cable jacket or insulation material in order toassess the potential for self-induced cable fires at PINGP. Cables were filtered foridentifying cables in conduit, low voltage (under 125VDC) cables, and cables that are inthe database as placeholder duplicates for future engineering changes.

The results ofthe applied filters indicate that the entire quantity of thermoplastic cable and cable withunknown jacket or insulation material are associated with cables in conduits, low voltagecables, or cables in the database as placeholder duplicates for future engineering changes.The above described methodology identified the following:

Total Linear Feet: 3,825,854 ftThermoset Linear Feet: 86%Thermoplastic Linear Feet: 13% (*See Note 1)Unknown Linear Feet: 1% (*See Note 1)*Note 1: Cables with thermoplastic jacket or insulation material and the cables withunknown insulation material are associated cables in conduits, low voltage cables, orcables outside of Fire PRA compartments.

NUREG-1 805 section 7.3 states "It is common practice to consider only self-induced cable fires to occur in power cable trays since they carry enough electrical energy forPage 60 of 81 L-PI-15-041 NSPMEnclosure 1ignition.

Control and instrumentation cables typically do not carry enough electrical energy for self-ignition."

Self-Ignited Cable Fires: Based on these results, it is concluded that that self-ignited cable fires do not need to be included in the PINGP Fire PRA model, including the areaunder the raised floor, by documenting that the thermoplastic cabling at PINGP is eitherin conduit or low voltage (low energy) cabling.

All of the cables beneath the flooring inthe control room were found to not be subject to self-ignition.

Cable Fires due to Welding/Cutting

& Junction Boxes: The risk contribution of cablefires due to hotwork and junction box fires in the main control room was assessedfollowing the guidance recommended in Fire PRA FAQ 13-0005 and Fire PRA FAQ 13-0006 respectively.

The analysis suggests that the risk contribution is relatively low (i.e.,less than 1 E-8) for each of the ignition sources.Transient Fires & Transient Fires due to Welding/Cutting:

Transient fires (including transient fires due to hotwork) that could occur in locations near cabinets or the maincontrol board are postulated in the Main Control Room as follows.

For each panelscenario identified, a transient fire ignition frequency and a transient fire due to hotworkcontribution are added. That is, the total scenario ignition frequency for a main controlboard panel or electrical cabinet scenario includes a contribution based on multiplying the geometric floor area factor times the total transient frequency in the control room.The postulated transient fires are considered to have the same impacts as the fixed-ignition-source-scenarios originating in the panel or cabinet.

These transient firescenarios incorporate all of the cable failures associated with the adjacent panels. Forexample, using the methodology from NUREG/CR-6850, transient scenarios involving a317 kW fire postulated in the open floor area around each main control board panel orelectrical cabinet includes damage to cabinets within 4 feet of the edge of the panel.Thus, all cables that are associated with cabinets in this geometric area will be includedin a transient scenario.

Exposed cabling, not associated with the cabinets in thisgeometric area that are located under the MCR floor, are excluded from these transient scenarios based on the following justification:

the low likelihood of hotwork beingconducted beneath the floor while the plant is in operation and the low likelihood of firepropagation due to quick suppression activities in the control room.Finally, the fire modeling results for the Main Control Room abandonment analysissuggests that electrical cabinets alone are enough to generate abandonment conditions.

Since the Fire PRA cables are mapped to the applicable panels as targets, thequantified CCDPs and CLERPs include the impact of cables that may be routed throughthe under floor. Furthermore, given the relative quick suppression activities in thecontrol room (as suggested by a manual suppression curve with a constant of 0.33), theanalysis assumes that transient and cabinet fires will, on average, be controlled beforepropagating to the underfloor.

NRC Request (PRA RAI 01.e):e. FSS-B2-01:

Evaluation of MCB FiresResponse to subpart (e) of this RAI will be submitted in separate correspondence by June26,2015.Page 61 of 81 L-PI-15-041 NSPMEnclosure 1NRC Request (PRA RAI 01.f):f. FSS-C5-01:

The disposition to this F&O states that self-ignited cable fires are screenedfrom consideration for all locations on the basis that all cables are either qualified orrouted through conduit, as concluded in Engineering Change (EC) 20695. However, thelicensee's analysis indicates that not all cable trays identified as targets are listed in EC20695 (Section 3.1 of FPRA-PI-RRA);

there are significant amounts of thermoplastic cabling located within cable trays in Fire Area 18, a risk significant area based on riskresults in Attachment W of the LAR (Section 5.1.3 of FPRA-PI-RRA);

and that a differing conclusion of EC 20695 may exist (Section F3 of FPRA-PI-SCA).

Given these discrepancies, please provide justification for the exclusion of self-ignited cable fires from the FPRA. If such fires are excluded on the basis of cable voltage,provide a technical basis for doing so. Alternatively, provide updated risk results as partof the aggregate change-in-risk analysis requested in PRA RAI 03, treating such firesconsistent with accepted guidance (e.g., FAQ 13-0005).

NSPM Response (PRA RAI 01.f):f. Self-ignited cable fires are screened for all locations in the plant based on the evaluation documented in the Engineering Change Package 20695 (EC 20695). The Engineering Change Package 20695 concludes the following:

The identification of cable type and linear feet was done on a cable-by-cable basis, withthe total linear feet of each cable type also identified.

The thermoset cable length isremoved since it cannot self-ignite.

Subsequently, a series of filters were applied to thethermoplastic and cables with unknown cable jacket or insulation material in order toassess the potential for self-induced cable fires at PINGP. Cables were filtered foridentifying cables in conduit, low voltage (under 125VDC) cables, and cables that in thedatabase as placeholder duplicates for future engineering changes.

The results of theapplied filters indicate that the entire quantity of thermoplastic cable and cable withunknown jacket or insulation material are associated with cables in conduits, low voltagecables, or cables in the database as placeholder duplicates for future engineering changes.The above described methodology identified the following:

Total Linear Feet: 3,825,854 ftThermoset Linear Feet: 86%Thermoplastic Linear Feet: 13% (*See Note 1)Unknown Linear Feet: 1% (*See Note 1)*Note 1: Cables with thermoplastic jacket or insulation material and the cables withunknown insulation material are associated cables in conduits, low voltage cables, orcables outside of Fire PRA compartments.

NUREG-1805 section 7.3 states "It is common practice to consider only self-induced cable fires to occur in power cable trays since they carry enough electrical energy forignition.

Control and instrumentation cables typically do not carry enough electrical energy for self-ignition."

Page 62 of 81 L-PI-15-041 NSPMEnclosure 1Based on these results, it is concluded that that self-induced cable fires do not need tobe included in the PINGP Fire PRA model by documenting that the thermoplastic cablingat PINGP is either in conduit or low voltage (low energy) cabling.NRC Request (PRA RAI 01.g):g. FSS-D7-02:

Total failure probability of credited detection and suppression systemsResponse to subpart (g) of this RAI will be submitted in separate correspondence by June26, 2015.NRC Request (PRA RAI 01.h):h. IGN-A 1-01: Fire ignition frequencies Response to subpart (h) of this RAI will be submitted in separate correspondence by July24, 2015, pending resolution of outstanding RAIs.PRA RAI 02 -Internal Event F&OsSection 2.4.3.3 of NFPA 805 states that the PRA approach,

methods, and data shall beacceptable to the NRC. The RG 1.205 identifies NUREG/CR-6850 as documenting amethodology for conducting a FPRA and endorses, with exceptions and clarifications, NEI 04-02,Revision 2, as providing methods acceptable to the staff for adopting a fire protection programconsistent with NFPA 805. The RG 1.200 describes a peer review process utilizing anassociated ASME/ANS standard (currently ASME/ANS-RA-Sa-2009) as one acceptable approach for determining the technical adequacy of the PRA once acceptable consensus approaches or models have been established.

The primary results of a peer review are theF&Os recorded by the peer review and the subsequent resolution of these F&Os.Please provide clarification to the following dispositions to Internal Events F&Os and SRassessments identified in Attachment U of the LAR that have the potential to impact the FPRAresults and do not appear to be fully resolved:

[This RAI includes Subparts a, b, and c, as shown below along with NSPM responses]

NRC Request (PRA RAI 02.a):a. QU-C2: Joint human error probabilities (HEPs)Response to subpart (a) of this RAI will be submitted in separate correspondence by June26, 2015.NRC Request (PRA RAI 02.b):b. SY-A8: The disposition to this F&O states that in evaluating the extent of theinconsistency highlighted by the peer review, similar modeling inconsistencies related toPage 63 of 81 L-PI-15-041 NSPMEnclosure 1instrumentation and control components were identified yet, 'The current model isconsidered conservative and adequate for the risk-informed NFPA -805 application".

Please summarize how this conclusion was reached.NSPM Response (PRA RAI 02.b):b. The referenced finding (SY-A8) found that a more consistent treatment of the component boundaries was needed. Although this Finding remains open, the current modeling isconsidered conservative.

This conservatism is due to inclusion of individual basicevents within the Fire PRA model that are already accounted for within the overallcomponent boundary (and are therefore included in the associated failure rate) of alarger component.

This results in a slight increase in the failure rates associated withthese components, which is deemed conservative.

This double counting originates fromthe Internal Events model upon which the Fire PRA is based. This conservative modeling is reflected in both compliant and variant cases of the model.NRC Request (PRA RAI 02.c):c. SY-B14: The disposition to this F&O does not address the issue associated with loss ofpump net positive suction head (NPSH) identified by the peer review.Please clarify whether and how the PRA assesses the impact of not crediting containment fan cooler units or containment spray on NPSH for recirculation.

NSPM Response (PRA RAI 02.c):c. The PRA model considers potential NPSH issues related to RHR pump operability during events in which the containment sump is used as the supply source to theReactor Coolant System (RCS). Based on pump NPSH requirements and NPSH testingand analysis, it has been concluded that:* Pump operability would not be impacted by debris if containment systemsoperate as designed (sump water is passed through filter/strainer prior to pumpsuction).

• Pump operability would also not be adversely impacted by the risingtemperatures in the containment sump and does not rely on containment overpressure for NPSH requirements.

Based on the above assumptions and supporting thermodynamic calculations, it Wasconcluded that there is no need to model Containment Fan Coil Units and/orContainment Spray operation/failure in the accident sequence evaluations.

PRA RAI 03 -Integrated AnalysisResponse to PRA RAI 03 will be submitted in separate correspondence by July 24, 2015,pending resolution of outstanding RAIs.Page 64 of 81 L-PI-15-041 NSPMEnclosure 1PRA RAI 04- Transient Fire Placement at Pinch PointsSection 2.4.3.3 of NFPA 805 states that the PRA approach,

methods, and data shall beacceptable to the NRC. The RG 1.205 identifies NUREG/CR-6850 as documenting amethodology for conducting a FPRA and endorses, with exceptions and clarifications, NEI 04-02, Revision 2, as providing methods acceptable to the staff for adopting a fireprotection program consistent with NFPA 805. Methods that have not been determined to beacceptable by the NRC staff, or acceptable methods that appear to have been applieddifferently than described, require additional justification to allow the NRC staff to complete itsreview of the proposed method.The NRC staff could not identify in the LAR a description of how "pinch points" for transient fireswere treated in the FPRA. Per NUREG/CR-6850 Section 11.5.1.6, transient fires should at aminimum be placed in locations within the plant PAUs where CCDPs are highest for that PAU,i.e., at "pinch points."

Pinch points include locations of redundant trains or the vicinity of otherpotentially risk-relevant equipment.

Cable congestion is typical for areas like the CableSpreading Room (CSR), and so placement of transient fire at pinch points in those locations isimportant.

Hot work should be assumed to occur in locations where hot work is possible, evenif improbable, keeping in mind the same philosophy.

[This RAI includes Subparts a and b, as shown below along with NSPM responses]

NRC Request (PRA RAI 04.a):a. Please clarify how '`pinch points" were identified and modeled for general transient firesand transient fires due to hot work.NSPM Response (PRA RAI 04.a):a. The ignition frequencies of transient fires, transient fires due to hotwork, and cable firesdue to hotwork are not apportioned by counting ignition

sources, as is the case for fixedsources such as electrical

cabinets, pumps, etc. Instead, the frequency of individual firescenarios postulated in a given compartment is apportioned using the following factorswhen applicable:

(1) the floor area ratio (also referred as the geometrical weighting factor);

(2) the maintenance,

storage, and occupancy influence factors as defined inChapter 6 of NUREG/CR-6850; and (3) cable load weighting factors as recommended inChapter 6 of NUREG/CR-6850.

As a practical and convenient

approach, the contribution from these ignition sources isrigorously assigned to the same transient fire locations (referred to as transient zones inthe PINGP Fire PRA) so that no location in the plant is left without a contribution fromthese ignition sources.

These transient zones are defined so that the entire open floorarea of the compartment is accounted for. No portion of a floor area has been excluded.

Thus, the contribution from these three ignition sources has been included in every openfloor area within every compartment in the scope of the Fire PRA. These fires arepostulated so that they propagate through the cable trays in the transient zone anddamage conduits that are also located in the transient zone. This approach ensures thatno pinch points in the compartments are missed.Page 65 of 81 L-PI-15-041 NSPMEnclosure 1Trays and conduits in the transition of transient zones have been mapped to theadjacent transient zones to ensure targets are accounted for in adjacent fire scenarios.

NRC Request (PRA RAI 04.b):b. Please describe how general transient fires and transient fires due to hot work aredistributed within the PAUs at Prairie Island. In particular, identify the criteria used todetermine where such ignition sources are placed within the PA Us.NSPM Response (PRA RAI 04.b):b. Transient fires and transient hot work fires are included in all compartments quantified inthe fire PRA. For compartments where no detailed fire modeling is performed, the totaltransient fire frequency (e.g., transient fire frequency plus the transient hot work firefrequency) is included in the total fire compartment frequency and all fire PRA targets inthe compartment are damaged.

For fire compartments where detailed fire scenarios have been defined, transient fires are postulated in smaller areas designated astransient zones that together cover the total floor area in each fire compartment (i.e.,there are no open floor areas where transient fire scenarios are excluded).

It should benoted that there is no physical overlap between the transient zone floor areas howeverthe targets are extended beyond the specific zones to incorporate spread of the fire.PRA RAI 05- Cable Fires Caused by Welding and CuttingSection 2.4.3.3 of NFPA 805 states that the PRA approach,

methods, and data shall beacceptable to the NRC. The RG 1.205 identifies NUREG/CR-6850 as documenting amethodology for conducting a FPRA and endorses, with exceptions and clarifications, NEI 04-02, Revision 2, as providing methods acceptable to the NRC staff for adopting a fire protection program consistent with NFPA 805. In a letter dated July 12, 2006, to NEI (ADAMS Accession No. ML061660105),

the NRC established the ongoing FAQ process where official agencypositions regarding acceptable methods can be documented until they can be included inrevisions to RG 1.205 or NEI 04-02.Appendix H of the LAR does not indicate that FAQ 13-0005, "Cable Fires Special Cases: Self-Ignited and Caused by Welding and Cutting,

" dated June 26, 2013, was used in preparation ofthe FPRA.Please explain whether the treatment of cable fires caused by welding and cutting is consistent with FAQ 13-0005, and if not, provide justification.

If justification cannot be provided, thenprovide treatment of such fires consistent with NRC guidance in the integrated analysis providedin response to PRA RAI 03.NSPM Response (PRA RAI 05):The Prairie Island Fire PRA follows the guidance of Fire PRA FAQ 13-0005 for assessing cablefires caused by welding and cutting.

No exceptions from the FAQ were used. Fire PRA FAQ13-0005 documents a method for quantifying the risk contribution of cable fires due to hot work.The FAQ suggests that cable fires due to hot work will be limited to one cable tray. Based onPage 66 of 81 L-PI-15-041 NSPMEnclosure 1the assumption that fires will be limited to one cable tray, the first two screening stepsdocumented in Fire PRA FAQ 13-0005 suggest to calculate the CCDP for each of the cabletrays in the fire compartment.

The maximum CCDP per fire compartment is selected andmultiplied by the cable fire due to hot work frequency assigned to the fire compartment.

Thatmultiplication results in a CDF associated with cable fires due to hot work for each firecompartment.

The first two steps in the FAQ described above were performed for the firecompartment receiving detailed fire modeling analysis.

For the remaining fire compartments (i.e., "those fire compartments treated as full compartment burn"), the full frequency of cablefires due to welding and cutting corresponding to the specific fire compartment is assigned.

PRA RAI 06- Junction BoxesSection 2.4.3.3 of NFPA 805 states that the PRA approach,

methods, and data shall beacceptable to the NRC. The RG 1.205 identifies NUREG/CR-6850 as documenting amethodology for conducting a FPRA and endorses, with exceptions and clarifications, NEI 04-02, Revision 2, as providing methods acceptable to the NRC staff for adopting a fire protection program consistent with NFPA 805. In letter to NEI dated July 12, 2006 (ADAMS Accession No.ML061660105),

the NRC established the ongoing FAQ process where official agency positions regarding acceptable methods can be documented until they can be included in revisions toRG 1.205 or NEI 04-02.Appendix H of the LAR does not indicate that FAQ 13-0006, "Modeling Junction Box Scenarios in a FPRA," dated May 6, 2013, was used in preparation of the FPRA.Please explain whether the treatment of junction box fires is consistent with FAQ 13-0006, and ifnot, provide justification.

If justification cannot be provided, then provide treatment of junctionbox fires consistent with NRC guidance in the integrated analysis provided in response to PRARAI 03.NSPM Response (PRA RAI 06):Fire PRA FAQ 13-0006 was used for the PINGP Fire PRA. No exceptions from the FAQ wereused. Fire PRA FAQ 13-0006 guidance documents a method for quantifying the riskcontribution of junction boxes. The FAQ suggests that junction box fires will be limited to onejunction box. The screening process documented in the FAQ is similar to that described forcable fires due to hot work. Option 2 of the FAQ was used for apportioning junction boxfrequency.

Since the ignition frequency of cable fires due to hot work is higher than the junctionbox frequency, and the count of raceways is assumed to be similar to the count of junctionboxes the risk contributions for cable fires due to hot work provide a bounding estimate of whatthe contribution for junction boxes can be. Therefore, these are relatively low contributions compared to the plant CDF.PRA RAI 07 -Sensitive Electronics Response to PRA RAI 06 will be provided in separate correspondence by June 26, 2015.Page 67 of 81 L-PI-15-041 NSPMEnclosure 1PRA RAI 08 -Conditional Probabilities of Spurious Operations Response to PRA RAI 08 will be provided in separate correspondence by June 26, 2015.PRA RAI 09- Counting and Treatment of Bin 15 Electrical CabinetsSection 2.4.3.3 of NFPA 805 states that the PRA approach,

methods, and data shall beacceptable to the NRC. The RG 1.205 identifies NUREG/CR-6850 as documenting amethodology for conducting a FPRA and endorses, with exceptions and clarifications, NEI 04-02, Revision 2, as providing methods acceptable to the NRC staff for adopting a fire protection program consistent with NFPA 805. Methods that have not been determined to be acceptable by the NRC staff, or acceptable methods that appear to have been applied differently thandescribed, require additional justification to allow the NRC staff to complete its review of theproposed method.The NRC staff could not identify in the LAR or the licensee's analysis how the licensee countedand treated Bin 15 Electrical Cabinets.

In light of this, address the following:

[This RAI includes Subparts a through d, as shown below along with NSPM responses]

NRC Request (PRA RAI 09.a):a. Per Section 6.5.6 of NUREG/CR-6850, fires originating from within "well-sealed electrical cabinets that have robustly secured doors (and/or access panels) and that house onlycircuits below 440 V" do not meet the definition of potentially challenging fires and,therefore, should be excluded from the counting process for Bin 15. By counting thesecabinets as ignition sources within Bin 15, the frequencies applied to other cabinets maybe inappropriately reduced.Please clarify that this guidance is being applied.

If not, then address the impact as partof the integrated analysis performed in response to PRA RAI 03.NSPM Response (PRA RAI 09.a):a. Cabinets were counted in Bin 15 following the guidance of NUREG/CR-6850.

Additional supporting guidance from Supplement 1 of NUREG/CR-6850 was used oncharacterizing and counting this bin, including dividing Bin 15 into two bins, Bin 15.1 -Non-High Energy Arcing Fault (HEAF) and Bin 15.2 -HEAF. Consistent with thisguidance, well-sealed electrical cabinets that have robustly secured doors (and/oraccess panels) and that house only circuits below 440V were not counted as electrical cabinets (i.e., Bin 15).NRC Request (PRA RAI 09.b):b. Please clarify if the criteria used to evaluate whether electrical cabinets below 440V are"well sealed" are consistent with guidance in Chapter 8 of Supplement 1 of NUREG/CR-Page 68 of 81 L-PI-15-041 NSPMEnclosure 16850. If not, then address the impact as part of the integrated analysis performed inresponse to PRA RAI 03.NSPM Response (PRA RAI 09.b):b. The counting guidance provided in Chapter 8 of Supplement 1 of NUREG/CR-6850 isfollowed whencounting "well sealed" cabinets below 440V.NRC Request (PRA RAI 09.c):c. All cabinets having circuits of 440V or greater should be counted for purposes of Bin 15frequency apportionment based on the guidance in Section 6.5.6 of NUREG/CR-6850.

Please clarify that this guidance is being applied.

If not, then address the impact as partof the integrated analysis performed in response to PRA RAI 03.NSPM Response (PRA RAI 09.c):c. Identified cabinets having circuits above 440V were counted in Bin 15 following theguidance of NUREG/CR-6850.

Additional supporting guidance from Supplement 1 ofNUREG/CR-6850 was used on characterizing and counting this bin, including dividingBin 15 into two bins, Bin 15.1 -Non-HEAF and Bin 15.2 -HEAF.NRC Request (PRA RAI 09.d):d. For those cabinets that house circuits of 440V or greater, propagation of fire outside theignition source should be evaluated based on guidance in Chapter 6 of NUREG/CR-6850, which states that "an arcing fault could compromise panel integrity (an arcing faultcould burn through the panel sides, but this should not be confused with the high energyarcing fault type fires)."Please describe how fire propagation outside of well-sealed cabinets greater than 440 Vis evaluated.

If propagation is not evaluated, then address the impact as part of theintegrated analysis performed in response to PRA RAI 03.NSPM Response (PRA RAI 09.d):d. Well-Sealed cabinets were not considered for cabinets greater than 440V in the IgnitionFrequency binning process; as such fires in these cabinets are evaluated as if thecabinet was vented (i.e., fire propagation was modeled outside of the cabinets).

Firepropagation outside of electrical cabinets greater than 440V is evaluated as outlined inChapter 6 of NUREG/CR-6850.

PRA RAI 10 -High Energy Arcing Faults (HEAF)Section 2.4.3.3 of NFPA 805 states that the PRA approach,

methods, and data shall beacceptable to the NRC. The RG 1.205 identifies NUREG/CR-6850 as documenting amethodology for conducting a FPRA and endorses, with exceptions and clarifications, NEI 04-Page 69 of 81 L-PI-15-041 NSPMEnclosure 102, Revision 2, as providing methods acceptable to the NRC staff for adopting a fire protection program consistent with NFPA 805. Methods that have not been determined to be acceptable by the NRC staff, or acceptable methods that appear to have been applied differently thandescribed, require additional justification to allow the NRC staff to complete its review of theproposed method.The NRC staff could not identify in the LAR or licensee's analysis a description of how HEAFswere modeled.

Per Appendix P of NUREG/CR-6850, HEAF events and other types of fireshave different non-suppression probability (NSP) curves. In addition, the NRC staff'sinterpretation of the NUREG/CR-6850 guidance is that the growth of a fire subsequent to aHEAF event, unlike other types of fires, instantaneously starts at a non-zero HRR because ofthe intensity of the initial heat release from the HEAF.Please confirm that HEAF events have been modelled using the acceptable HEAF evaluation methods.

If alternative methods have been used, provide a justification of the FPRA's treatment of HEAF events and the ensuing fire that includes a discussion of conservatisms and non-conservatisms relative to the accepted methods and assesses the associated impacts on thefire total and delta risk results.

Alternatively, replace the current approach with an acceptable approach in the integrated analysis performed in response to PRA RAI 03. Note that theresponse should address the treatment of all HEAF scenarios, including in the hot gas layer andmulti-compartment analyses.

NSPM Response (PRA RAI 10):The PINGP Fire PRA models high energy arcing faults using acceptable evaluation methods.The Fire PRA HEAF analysis follows the guidance described in Appendix M of NUREG/CR-6850. This approach suggests that high energy arcing faults are events with two distinctphases: an initial energetic phase (the arcing explosion),

and an ensuing fire phase. Based onthis approach, the peak heat release rate value is assigned to the switchgear or load centerwhere the high energy arcing fault is postulated.

It is assumed that the peak heat release rate isachieved at the time of the explosion (i.e., no fire growth phase is credited).

Fire propagation through secondary combustibles above the switchgear or load center is modeled following theguidance described in Appendix M of NUREG/CR-6850.

The high energy arcing fault fire scenarios in the Fire PRA model are quantified using thesuppression curve listed in Table 14-2 of Supplement 1 to NUREG/CR-6850 for high energyarcing fault scenarios.

This suppression curve for high energy arcing faults is used fordetermining the non-suppression probabilities associated with the different damage statesassociated with each fire scenario.

The Non-Suppression Probability (NSP) credit taken forHEAF scenarios is 0.011.PRA RAI 11 -Time to Delayed Detection Section 2.4.3.3 of NFPA 805 states that the PRA approach,

methods, and data shall beacceptable to the NRC. Section 2.4.4.1 of NFPA 805 further states that the change in publichealth risk arising from transition from the current fire protection program to an NFPA 805 basedprogram, and all future plant changes to the program, shall be acceptable to the NRC. TheRG 1.174 provides quantitative guidelines on CDF, LERF, and identifies acceptable changes tothese frequencies that result from proposed changes to the plant's licensing basis andPage 70 of 81 L-PI-15-041 NSPMEnclosure 1describes a general framework to determine the acceptability of risk-informed changes.

TheNRC staff's review of the information in the LAR has identified additional information that isrequired to fully characterize the risk estimates.

The licensee's analysis (Section 3.0 of FPRA-PI-SCA) appears to indicate that regardless oflocation, the FPRA assumes a time to delayed detection of 10 minutes instead of the 15minutes used in Appendix P of NUREG/CR-6850 for fire scenarios should automatic detection be unavailable or a fire watch not be present.Please use the generic 15 minutes or provide justification for using 10 minutes.NSPM Response (PRA RAI 11):The PINGP Fire PRA no longer uses the delayed detection time of 10 minutes.

The time todelayed detection has been updated to 15 minutes.An assumption has been included within the FPRA-PI-SCA notebook to state the usage of a 15minute delayed detection time. The Integrated analysis associated with the response to PRARAI 03 will reflect the use of a 15 minute delayed detection time.PRA RAI 12 -MCR Abandonment Response to PRA RAI 12 will be provided in separate correspondence by June 26, 2015.PRA RAI 13 -Calculation of ACDF, ALERF and Additional Risk of RecoveryActionsSection 2.4.3.3 of NFPA 805 states that the PRA approach,

methods, and data shall beacceptable to the NRC. Section 2.4.4.1 of NFPA 805 further states that the change in publichealth risk arising from transition from the current fire protection program to an NFPA 805 basedprogram, and all future plant changes to the program, shall be acceptable to the NRC. TheRG 1. 174 provides quantitative guidelines on CDF, LERF, and identifies acceptable changes tothese frequencies that result from proposed changes to the plant's licensing basis anddescribes a general framework to determine the acceptability of risk-informed changes.

TheNRC staff's review of the information in the LAR has identified additional information that isrequired to fully characterize the risk estimates.

Section W.2 of the LAR provides some description of how the change-in-risk and the additional risk of recovery actions associated with variances from deterministic requirements (VFDRs) isdetermined but not enough detail to make the approach completely understood.

As a result,please provide the following:

[This RAI includes Subparts a, b, and c, as shown below along with NSPM responses]

IPage 71 of 81 L-PI-15-041 NSPMEnclosure 1NRC Request (PRA RAI 13.a):a. A summary of how the change in risk was determined for fire areas that credit MCRabandonment due to loss of habitability and due to loss of control (e.g., Relay Room).Include a discussion of how the CCDPs/CLERPs were determined for both the variantplant and the compliant plant models for these areas.NSPM Response (PRA RAI 13.a):a. The change in risk (i.e., delta risk) associated with VFDRs was calculated as thedifference in risk between the variant plant (i.e., the post-transition plant including plantmodifications) and the compliant plant (i.e., the post-transition plant in which the VFDRsare assumed to be deterministically resolved).

The change in risk in Fire Areas 13 and 18 was calculated as a delta core damagefrequency (delta CDF) and delta large early release frequency (delta LERF) for eachUnit in the same manner as for other fire areas and was evaluated as follows.A determination of whether the control room would be abandoned or not was made foreach fire scenario within the Fire PRA. A fire scenario that was found to lead toabandonment in the variant plant was also considered to lead to abandonment in thecompliant plant. MCR abandonment was credited only for some scenarios occurring inFire Areas 13 (Control Room) and 18 (Relay and Cable Spreading Room).For each fire scenario, the conditional core damage probability (CCDP) and conditional large early release probability (CLERP) of the variant plant were calculated.

To calculate the CCDP (CLERP) associated with a given VFDR, the compliant plant was modeled byaltering the model logic to mimic a plant where the VFDR was deterministically

resolved, i.e., it no longer existed.

In particular:

• Components credited to establish the relied-upon success path and identified asimpacted in the VFDR were assumed to be unaffected by fire impacts." For recovery actions credited to help resolve the VFDR, the execution portion ofthe human error probability (HEP) representing the recovery action failure wasset to zero. This approach was selected to represent a compliant plant where theoperator action is assumed to take place in the control room (when it is notabandoned) or at a hypothetical primary control station (when the control room isabandoned),

and therefore is not a recovery action. Setting the execution portionof the HEP to zero is conservative because operator

actions, even when takingplace under optimal conditions in the control room, may fail to be executedproperly.

Setting the execution portion to zero is otherwise an adequate proxy torepresent the fact that, in the compliant plant, operator actions would be easier toexecute than in the variant plant, where these actions are performed away fromthe control room. The potential for failing to take a required action (cognitive failure) exists both in the variant and compliant plant; therefore, this portion of theHEP is kept in both the variant and compliant plant models. For some VFDRs,the cognitive failure portion of the HEP was also set to zero in the compliant plant, in effect conservatively setting the entire HEP of the recovery to zero in thecompliant plant.Page 72 of 81 L-PI-15-041 NSPMEnclosure 1* VFDRs for which a credited component or recovery action was not modeled inthe Fire PRA are addressed separately in the response to Item c of this RAI.The changes described above lead to a CCDP (CLERP) in the variant plant higher thanin the compliant plant. The change in risk for a given VFDR is calculated as the sum,over all the fire scenarios impacted by the VFDR, of the difference in CDF (LERF) ofeach fire scenario, calculated as the fire scenario frequency multiplied by the difference in variant plant CCDP (CLERP) minus compliant plant CCDP (CLERP).NRC Request (PRA RAI 13.b):b. A description of how the reported additional risk of recovery actions was calculated, including any special calculations performed for the MCR and other abandonment areas(if applicable).

Note that it is unclear why the discussion provided in Section W2.2 ofthe LAR states that "the additional risk of recovery actions is calculated separately bycomparing the fire area CDF and LERF of the variant and compliant plant" when theadditional risk of recovery actions is not equivalent to the delta risk for all fire areas.NSPM Response (PRA RAI 13.b):b. For a given fire area, the additional risk of recovery actions was calculated as thedifference in CDF (LERF) in the variant plant (post-transition plant where the creditedrecovery actions take their nominal HEPs), and the CDF (LERF) of the post-transition plant where the credited recovery actions have the execution portion of their HEP set to0 (or, for some VFDRs, the entire HEP of the associated credited recovery actionconservatively set to 0). This method is applied to all fire areas, regardless of whether afire area credits, or not, control room abandonment.

This calculation method was selected based on the following considerations:

" Because the fire area CDFs (LERFs) calculated with that method differ only bytheir HEP values, the resulting differences in risk are due to credited recoveryactions only. This differs from the general calculation of changes in risk (deltarisks) described in the response to Item a of this RAI. The overall change in riskof a fire area could be used as an adequate, bounding proxy for the additional risks of its recovery

actions, but would suffer from the drawback that some of itsunderlying cutsets pertain to equipment lost to the fire in the variant plant, butotherwise not related to the recovery actions credited in the fire area.* Setting the execution portion of the HEP to 0 is conservative because norecovery action is perfectly reliable.

That is, in reality, the execution portion ofthe HEP is always greater than 0. Consequently, the differences in CDF andLERF calculated with this method are conservative approximations of theadditional risk of recovery actions.

There is added conservatism when the entireHEP is set to 0.In Section W.2.2 of the LAR, the sentence stating:

"the additional risk of recovery actionsis calculated separately by comparing the fire area CDF and LERF of the variant andcompliant plant" was intended to outline the fact that the additional risk of recoveryactions was not approximated by the change in risk (delta CDF and delta LERF due toVFDRs in each fire area, calculated as described in response to Item a of this RAI).Page 73 of 81 L-PI-15-041 NSPMEnclosure 1While such an approximation would be adequate and bounding, it was judged that aseparate calculation focusing on the changes in CDF and LERF due only to changes inHEPs of credited recovery actions (representing the differences in HEPs between thevariant and compliant plant) was a more accurate representation of the additional risk ofrecovery actions.NRC Request (PRA RAI 13.c):c. A summary of the types of VFDRs that were identified but not modeled in the FPRA.Include any qualitative rationale for excluding these from the change-in-risk calculations.

NSPM Response (PRA RAI 13.c):c. There are essentially two types of VFDRs that were identified but not modeled in theFPRA, "not modeled" here meaning that no change to the compliant plant logic wasneeded to model the VFDR. They are as follows:" VFDRs resolved by plant modification.

These VFDRs correspond to the caseswhere a proposed plant modification provides a success path for the nuclearsafety performance criteria previously identified as challenged in the VFDR,thereby indicating that the VFDR will no longer exist when the modification isimplemented.

In these cases, the delta risk is zero.* VFDRs with an insignificant delta risk based on a qualitative evaluation.

Somecomponents identified in a VFDR as potentially failed due to fire effects may notbe modeled in the Fire PRA. This was done in the situations where it was shownthat the change in risk associated with the component is insignificant.

Anexample would be the situation where there is sufficient diversity and redundancy of instrumentation to ensure that a monitoring parameter, identified as lost in aVFDR, would be in fact available.

In effect, the delta risk is zero.PRA RAI 14- Attachment W Inconsistencies Inconsistencies were noted within Attachment W for particular fire areas. In light of this,[This RAI includes Subparts a, b, and c, as shown below along with NSPM responses]

NRC Request (PRA RAI 14.a.i):a. Provide clarification on the following inconsistencies, and discuss their significance tothe risk results reported in Tables W-6 and W-7.(i) In Table W-6, Unit 1 Fire Area 84 is indicated as having VFDRs (i.e., there is a "Yes"'under the "VFDR" column);

however, there is an "N/A"in the column forA CDF/A LERFPage 74 of 81 L-PI-15-041 NSPMEnclosure 1NSPM Response (PRA RAI 14.a.i):(i) This is a typo; there is no VFDR for Fire Area 84. Documentation for Attachment Whas been updated to reflect that there should be a "No" in the VFDR column of TableW-6 for Fire Area 84. A revision to LAR Attachment W will be provided with theresponse to PRA RAI 03.NRC Request (PRA RAI 14.a.ii):

(ii) In Table W-7, Unit 2 Fire Areas 1 and 20 are indicated as having no VFDRs (i.e.,there is a "No" under the "VFDR" column);

however, there is an "e, or epsilon in thecolumn for A CDF/ALERF.

NSPM Response (PRA RAI 14.a.ii):

(ii) This is a typo; VFDRs exist for Fire Areas 1 and 20. Documentation for Attachment W has been updated to reflect that there should be a "Yes" in the VFDR column ofTable W-7 for Fire Areas 1 and 20. A revision to LAR Attachment W will be providedwith the response to PRA RAI 03.NRC Request (PRA RAI 14.b):b. Describe what is meant by the use of "E," or epsilon, in columns for Fire AreaACDF/ALERF and additional risk of RAs. Address if epsilon is defined by a specific cut-off value(s).

NSPM Response (PRA RAI 14.b):b. Epsilon or "E" is a common mathematical term used to represent a small positiveinfinitesimal

quantity, whose limit is usually taken as F -+ 0. There was no "cutoff" valueused for epsilon, other than the FRANX truncation limits (1 E-1 2 for CDF and 1 E-1 3 forLERF) applied during the baseline and compliant case quantifications.

In the specificcase of this analysis, the variable epsilon was used if either there was no impact to PRA-modeled equipment due to the VFDR, or if the impact was too small to survivetruncation.

A note was added to Attachment W explaining

epsilon, as follows:"E" represents a small positive infinitesimal quantity whose impact is too small toaffect the analysis.

An updated version of Attachment W will be provided with the response to PRA RAI 03.NRC Request (PRA RAI 14.c):c. Describe what is meant by the use of "N/A"in columns for Fire Area CDF/LERF, A CDF/ALERF and additional risk of RAs.Page 75 of 81 L-PI-15-041 NSPMEnclosure 1NSPM Response (PRA RAI 14.c):c. If no Fire Risk Evaluation was required for the specific Fire Area an "N/A" (i.e., NotApplicable) was assigned for the columns CDF/LERF, ACDF/ALERF, and Additional Risk of RAs.For some Fire Areas there was an FRE (i.e., the Fire Area had VFDRs that wereevaluated using performance based methods) but had no recovery actions credited.

Inthose cases there was either a delta-risk or epsilon provided for the Fire Area delta-risk

columns, but "N/A" was used for the "Additional risk of recovery actions" columns.A note was added to Attachment W explaining "N/A," as follows:"N/A" indicates that no Fire Risk Evaluation was required or no Recovery Actionswere credited.

An updated version of Attachment W will be provided with the response to PRA RAI 03.PRA RAI 15- Implementation Item Impact on Risk Estimates Section 2.4.3.3 of NFPA 805 states that the PRA approach,

methods, and data shall beacceptable to the NRC. Section 2.4.4.1 of NFPA 805 further states that the change in publichealth risk arising from transition from the current fire protection program to an NFPA 805 basedprogram, and all future plant changes to the program, shall be acceptable to the NRC. TheRG 1. 174 provides quantitative guidelines on CDF, LERF, and identifies acceptable changes tothese frequencies that result from proposed changes to the plant's licensing basis anddescribes a general framework to determine the acceptability of risk-informed changes.

TheNRC staff's review of the information in the LAR has identified additional information that isrequired to fully characterize the risk estimates.

Implementation Item 20 in Table S-3 of the LAR commits to updating the FPRA and verifying the risk results after Table S-2 plant modifications have been incorporated.

However, Table S-3includes a number of procedural modifications that may affect the as-built and as-operated plantrisk models.[This RAI includes Subparts a, b, and c, as shown below along with NSPM responses]

NRC Request (PRA RAI 15.a):a. Please update Implementation Item 20 to include incorporation of all risk relevantmodifications in Tables S- 1, S-2, and S-3 into the FPRA before the final risk resultverification.

NSPM Response (PRA RAI 15.a):a. An updated Implementation Item 20 that commits to including all risk-relevant items inTables S-1, S-2, and S-3 into the FPRA before the final risk result verification is providedbelow. The Item is also updated for clarity with respect to the actions that will be takenshould the quantified risk measures for either unit exceed the RG 1.205 acceptance guidelines (see response to PRA RAI 15(b) below):Page 76 of 81 L-PI-15-041 NSPMEnclosure 1Current description provided in Table S-3 of the April 30, 2014 LAR Supplement:

Item 20:Update the Fire PRA Model, as necessary, after all modifications identified inTable S-2 are complete and as-built.

If the revised Fire PRA indicates anincrease in risk metrics such that the RG 1.205 acceptance guidelines are notmet, the configuration control process described in LAR Section 4.7.2 will beimplemented.

New, revised Table S-3 description:

Item 20:Update the Fire PRA Model, as necessary, after all modifications, procedure

changes, and other risk-relevant items identified in Tables S-1, S-2, and S-3 arecomplete and as-built.

If the revised Fire PRA indicates an increase in riskmetrics such that the RG 1.205 acceptance guidelines are not met, changes willbe made such that the Fire PRA results will fall within the acceptance guidelines.

These changes may include additional

analysis, procedure enhancements, plantmodifications, or other changqes determined to be necessary to reduce the overallrisk metrics to within the acceptance guidelines.

NRC Request (PRA RAI 15.b):b. Implementation Item 20 states, "If the revised Fire PRA indicates an increase in riskmetrics such that the RG 1.205 acceptance guidelines are not met, the configuration control process described in LAR Section 4.7.2 will be implemented."

Please clarify this statement.

NSPM Response (PRA RAI 15.b):b. The intent of Table S-3 Implementation Item 20 is to state that after implementation of allplant modifications (and after implementation of all other risk-relevant S-1, S-2 and S-3table items; see response to PRA RAI 15.a above), the Fire PRA will be revised toincorporate these changes, and the internal fires risk metrics will be recomputed.

Ifthese metrics are found to exceed the RG 1.205 acceptance guidelines, then anynecessary changes will be made to reduce the overall risk metrics to within theacceptance guidelines.

Table S-3 Implementation Item 20 is revised as shown in theresponse to PRA RAI 15.a above to better describe the analysis that will be performed.

NRC Request (PRA RAI 15.c):c. Tables S- 1 and S-2 include the new RCP seals for which no acceptable PRA modelexists yet and the time until an acceptable model exists is difficult to determine.

Please clarify how transition to NFPA-805 could be achieved with the currentimplementation items if an acceptable RCP seal model is delayed for an extended time.Page 77 of 81 L-PI-15-041 NSPMEnclosure 1NSPM Response (PRA RAI 15.c):c. All of the Reactor Coolant Pump (RCP) seals in both Prairie Island units have now beenreplaced with the Flowserve Three-Stage N-9000 Seal package with the Abeyance Seal.The new RCP seals for the Prairie Island Nuclear Generating plant are modelled usinggeneral PRA methods consistent with the consensus model endorsed by the NRC:WCAP-1 6175-P-A, Model for Failure of RCP Seals Given Loss of Seal Cooling in CENSSS Plants. The model developed for use in the Prairie Island internal events PRAmodels and for the Fire PRA uses the failure model of WCAP-1 6175 as the basis andthen incorporates the effect of failure modes and the Abeyance Seal on resulting sealfailure flow rates. Additionally, the SER for WCAP-16175 acknowledges that no damageto the seals is expected if the pumps are tripped within 20 minutes of the loss of sealcooling.

However, WCAP-1 6175 evaluated pumps with a smaller volume of cold wateraround the thermal barrier heat exchanger.

Therefore, the PRA model for Prairie Islandhas provided for additional time available to purge the cold water from Westinghouse pumps, consistent with the analyses in WCAP-16175.

The Flowserve report, PRAModel for Flowserve 3 Stage N-Seals with Abeyance Seal, Revision 0, dated 12/20/13, provides the bases for the development of the PRA model used in the Prairie Island firePRA. Transition to NFPA-805 can be achieved with the current implementation items byrevising Table S-3 to include the following new Implementation Item #66:Item 66:Upon NRC approval of the Flowserve topical report for the Reactor Coolant Pump(RCP) seals and related PRA model, the Prairie Island PRA model shall bereviewed using the final version of the topical report as well as anyexceptions/clarifications included in the NRC approval to determine if the internalevents and Fire PRA require a revision.

The Prairie Island internal events and FirePRA will be updated, if applicable, with the latest RCP seal information.

If theupdates result in a risk increase greater than the self-approval limits (1 E-07/yr forCDF and 1 E-08/yr for LERF), NSPM will take action to reduce the risk results towithin the self-approval limits. Compensatory measures established prior to theRCP seal replacement shall remain in place until the calculated risk increase iswithin the self-approval limits.PRA RAI 16- Use of Incipient Detection Response to PRA RAI 16 will be provided in separate correspondence by June 26, 2015.PRA RAI 17 -RCP Seal PRA ModelingResponse to PRA RAI 17 will be provided in separate correspondence by June 26, 2015.PRA RAI 18 -Deviations from Acceptable MethodsResponse to PRA RAI 18 will be provided in separate correspondence by June 26, 2015.Page 78 of 81 L-PI-15-041 NSPMEnclosure 1PRA RAI 19 -Defense-in-Depth and Safety MarginSection 2.4.3.3 of NFPA 805 states that the PRA approach,

methods, and data shall beacceptable to the NRC. Section 2.4.4.1 of NFPA 805 further states that the change in publichealth risk arising from transition from the current fire protection program to an NFPA 805 basedprogram, and all future plant changes to the program, shall be acceptable to the NRC. TheRG 1.174 provides quantitative guidelines on CDF, LERF, and identifies acceptable changes tothese frequencies that result from proposed changes to the plant's licensing basis anddescribes a general framework to determine the acceptability of risk-informed changes.

TheNRC staff's review of the information in the LAR has identified additional information that isrequired to fully characterize the risk estimates.

Section 4.5.2.2 of the LAR provides a high-level description of how the impact of transition toNFPA 805 impacts defense-in-depth (DID) and safety margin was reviewed, including using thecriteria from Section 5.3.5 of NEI 04-02 and from RG 1.205. However, no explanation isprovided of how specifically the criteria in these documents were utilized and/or applied in theseassessments.

[This RAI includes Subparts a and b, as shown below along with NSPM responses]

NRC Request (PRA RAI 19.a):a. Please provide further explanation of the method(s) or criteria used to determine when asubstantial imbalance between DID echelons existed in the Fire Risk Evaluations (FREs), and identify the types of plant improvements made in response to thisassessment.

NSPM Response (PRA RAI 19.a):a. Section 1.2 of NFPA 805 defines defense-in-depth (DID) as:1. Preventing fires from starting2. Rapidly detecting fires and controlling and extinguishing promptly those fires that dooccur, thereby limiting fire damage3. Providing an adequate level of fire protection for structures,

systems, andcomponents important to safety, so that a fire that is not promptly extinguished willnot prevent essential plant safety functions from being performed.

In general, DID is considered satisfied if the proposed licensing change does not resultin a substantial imbalance among these elements (or echelons).

The review of DID isqualitative and addresses each of the elements with respect to the proposed licensing change. It involves a review of plant documents such as the fire protection program,pre-fire plans, and administrative procedures.

In the context of the NFPA 805 transition, the DID evaluation accounts for the fact thatthe fundamental elements of the fire protection program and the design requirements forfire protection systems and features have been addressed in a manner consistent withthe requirements of Chapter 3 of NFPA 805. Accordingly, the DID evaluation focuses onpotential enhancements that may be required to maintain the balance of DID echelons.

Page 79 of 81 L-PI-1 5-041 NSPMEnclosure 1Systems and features associated with the first echelon of Defense in Depth (DID) dealwith control of combustibles and control of hot work. Generally, the Fire PRA assumeslimited credit for Administrative Controls that may limit the size, placement, andfrequency of transients.

Unless transient placement is precluded by design, the Fire PRAassumes transients can be placed near targets and hot work transients anywhere in the3-dimensional space near pinch points. With this approach, the Fire PRA helps uncoverpotential plant vulnerabilities regarding transients or hot work locations and providesinsights on the adequacy of Echelon 1. Combustible Control and Hot Work Control wereindirectly credited in the Fire PRA through the ignition frequencies for transient combustibles and hot work. Because these administrative controls reduce risk of fires inall nuclear plants, generic ignition frequencies used in the Fire PRA credit these controlsto be in place. The fire risk evaluations (FREs) found that the existing combustible andhot work controls were adequate and no plant improvement was required for thisechelon.The second echelon of DID involves the detection and suppression of fires. To evaluatethe adequacy of this echelon, insights from the Fire PRA can be used as a reference point. For example, fire detection and associated fire pre-plan procedures in a given firearea may be credited in order to meet DID considerations for this area, but otherwise may not have been credited in the Fire PRA. This could be based on the consideration that, for example, the relatively significant amount of diesel fuel oil present in the areacould challenge firefighting activities, thereby making it beneficial to credit detection forDID, to ensure that the fire is promptly controlled and extinguished.

The fire riskevaluations (FREs) found that the existing fire protection features and pre-fire planswere adequate and that no plant improvement was required for this echelon.The third echelon of DID deals with the adequacy of fire barriers, fire rated cables andsystems free of fire damage including procedural controls.

This echelon also deals withoperator actions such as recovery actions aimed at mitigating the adverse effects of thefire. As for the other echelons, the Fire PRA can be used as a reference point to identifypotential improvements for this echelon.

The FREs identified improvements for thisechelon.

For example, procedure F5 Appendix D, Impact of Fire Outside Control/Relay Room, was updated to include several additional fire areas in the list of fire areas wherea fire could potentially require manual operation of cooling water strainers.

NRC Request (PRA RAI 19.b):b. Please provide further discussion of the approach in applying the criteria for assessing safety margin in the FREs as described in NEI 04-02, "Guidance for Implementing aRisk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c),"

Revision 2, dated April2008 (ADAMS Accession No. ML081130188).

NSPM Response (PRA RAI 19.b):b. In accordance with Section 5.3.5.3 of NEI 04-02, the adequacy of Safety Margin isassessed by the consideration of categories of analyses utilized by the FRE. Safetymargins are considered to be maintained if:* Codes and standards or their alternatives accepted for use by the NRC are met,andPage 80 of 81 L-PI-15-041 NSPMEnclosure 1* Safety analysis acceptance criteria in the licensing basis (e.g., FSAR, supporting analyses) are met, or provide sufficient margin to account for analysis and datauncertainty.

The following summarizes the bases for ensuring the maintenance of safety margins:" The risk-informed, performance based processes utilized are based upon NFPA805, endorsed by the NRC in 10 CFR 50.48(c).

" The fire risk evaluation process is in accordance with NEI 04-02, which isendorsed by the NRC in Regulatory Guide 1.205.* The Fire PRA is developed in accordance with NUREG/CR-6850, which wasdeveloped jointly between the NRC and EPRI.* The internal events PRA and Fire PRA have received a formal industry peerreview based on the NEI guidelines, in order to ensure the Fire PRA meets theappropriate quality standards of ASME/ANS RA-Sa-2009.

Peer review of the FirePRA model has been conducted by diverse groups of PRA practitioners fromother PWR plants and industry.

These reviews generally cover all aspects of theFire PRA model and the administrative processes used to maintain and updatethe model. The peer review has generated specific recommendations for modelchanges, as well as guidance for improvements to processes and methodologies used in the Fire PRA model, and enhancements to the documentation of themodel and the administrative procedures used for model updates.

The Fire PRAmodel and administrative requirements are assessed and revised or clarified toaddress the issues identified through peer reviews.* Fire protection systems and features determined to be required by NFPA 805Chapter 4 have been confirmed to meet the requirements of NFPA 805 Chapter3 and their associated referenced codes and listings, or provided with acceptable alternatives using processes accepted for use by the NRC." Fire modeling performed in support of the transition has been performed withinthe Fire PRA utilizing codes and standards developed by industry and NRC staffwhich have been verified and validated in authoritative publications, such asNUREG-1824, "Verification and Validation of Selected Fire Models for NuclearPower Plant Applications."

In general, the fire modeling performed in support ofthe fire risk evaluations has been performed using conservative methods andinput parameters that are based upon NUREG/CR-6850.

Page 81 of 81 L-PI-15-041 NSPMEnclosure 2Enclosure 2Licensee Identified ChangesThis Enclosure identifies changes to LAR sections not directly related to RAI responses, andincludes the following:

LAR Section ChangeAttachment S Revise schedule statements on Page S-2 to clarify that two fullrefueling cycles are needed for completion of modifications Attachment S Delete S-3 item 30Licensee Identified Issue #11: Modification ScheduleThe implementation schedule in Attachment S, page S-2, should be revised for consistency withSection 5.5, Transition Implementation Schedule.

NSPM is requesting two full operating cycles,per unit, to complete the modifications listed in Table S-2, Plant Modifications Committed.

Thisavoids the need to work on both protection trains in the same outage, consistent with PINGPoutage scheduling practices.

Section 5.5 identifies that Table S-2 modifications will be completed before the end of thesecond full operating cycle for each unit after approval of the LAR. Also, Table S-2, page S-2includes a statement that NSPM is requesting "two full refueling cycles" beyond SE issuance tofully implement modifications.

However, page S-2 also includes a statement that S-2modifications will be complete by the "completion of the second refueling outage" per unit afterissuance of the NFPA 805 license amendment; this would not allow the two full operating cycles(which are the same as refueling cycles) that are described in Section 5.5 and were intended byNSPM. The modification schedule statement on page S-2 of the LAR should be changed toread as follows:NSPM will complete implementation of the modifications described in Table S-2 as follows:" By the completion of the first refueling outage per unit after issuance of the NFPA805 license amendment:

Items 8, 9, and 16" By the completion of the refueling outage after the second full refueling cycle per unitafter issuance of the NFPA 805 license amendment:

all remaining itemsAttachment S will be revised and submitted with the response to PRA RAI 03, and this updatewill include the above change.1 L-PI-15-041 NSPMEnclosure 2Licensee Identified Issue #2: Delete S-3 Item Regarding Compressed Air BottlesTable S-3, Item 30 is an Implementation Item to "Revise/initiate Orocedures and/or procureadditional compressed air bottles to achieve 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> to ensure we are 'safe and stable' at 24hours." This item was included in the initial 2012 LAR submittal (Reference 1 to the cover letterfor this enclosure).

However, the revised analyses for the 2014 LAR Supplement (Reference 2to the cover letter for this enclosure) no longer require compressed air and this item wasinadvertently not deleted from Attachment S. Attachment S will be revised and submitted withthe response to PRA RAI 03, and the update will include the deletion of Item 30 from Table S-3.2 L-PI-15-041 NSPMEnclosure 3Enclosure 3LAR Attachment L (revised)

The following pages include a revision toLAR Attachment L, NFPA 805 Chapter 3 Requirements for Approval, and includes information in response to FPE RAI 05.7 pages follow Northern States Power -Minnesota Attachment LNFPA 805 Chapter 3 Requirements for ApprovalL. NFPA 805 Chapter 3 Requirements for Approval(10 CFR 50.48(c)(2)(vii))

6 Pages AttachedNote that Revision 1 is a complete revision andno revision indicators are included.

PINGP Page L-1 -Revision 1PINGPPage L-1 -Revision 1

Northern States Power -Minnesota Attachment LNFPA 805 Chapter 3 Requirements for ApprovalApproval Request 1NFPA 805 Section 3.5.16NFPA 805 Section 3.5.16 states:"The fire protection water supply system shall be dedicated for fire protection use only.Exception No. 1: Fire protection water supply systems shall be permitted to be used toprovide backup to nuclear safety systems, provided the fire protection water supplysystems are designed and maintained to deliver the combined fire and nuclear safetyflow demands for the duration specified by the applicable analysis.

Exception No. 2: Fire protection water storage can be provided by plant systems servingother functions, provided the storage has a dedicated capacity capable of providing themaximum fire protection demand for the specified duration as determined in thissection."

Basis for Request:Contrary to the requirements of NFPA 805 Section 3.5.16, the fire protection water supplysystem at PINGP may periodically be utilized to supply water for non-fire protection purposes.

NRC approval is being requested since the fire water system can be aligned for screenwash system use and spent fuel pool makeup, and as such it does not meet the requirement orallowed exceptions.

Acceptance Criteria Evaluation:

The Mississippi river provides fire protection water. The system consists of two horizontal centrifugal fire pumps each rated at 2,000 gpm at 125 psig. One pump is motor driven (MDFP)and the other pump is diesel driven (DDFP). The 10" fire header is maintained between 108and 113 psig by a jockey pump. The motor driven fire pump will automatically start at 95 psig.If the header pressure drops to 90 psig, the diesel-driven fire pump will start. The motor anddiesel-driven fire pumps are designed to pump 2,000 gpm at a discharge pressure of 125 psi.The screenwash pump can be aligned to the fire protection header to supply 2,000 gpm at adischarge pressure of 125 psi.Each of the requested non-fire protection uses is discussed below.Fire protection water for screenwash function:

The motor-driven fire pump can be aligned to provide a backup water supply to theScreenhouse screenwash system in the event the screenwash pump is unavailable.

To use the fire pump to supplement the screenwash header flow, the pump must be startedmanually either locally or remotely.

Once the pump is operating, if no auto start signal existsfrom the fire protection header (i.e., if pressure in the fire water header has not been reduced byother uses), the discharge valve to the screenwash header (Control Valve CV-31131) opensautomatically via Solenoid Valve SV-33049 and supplies water to the screenwash header whenrequired.

If a demand is placed on the fire header from a suppression system actuation, the fireprotection header low pressure signal will cause SV-33049 to close CV-31131 therebyrealigning all water flow to the fire protection header. The diesel driven fire pump (DDFP) willalso be available to supply water to the fire protection water piping.There is also a bypass line around control valve CV-31131.

If manual valve FP-27-1 in thebypass line around CV-31131 is opened, it will be manually closed in the event of a fire toPINGPPage L-2 -Revision 1

Northern States Power -Minnesota Attachment LNFPA 805 Chapter 3 Requirements for Approvalrealign water flow to the fire protection header. If the manual bypass valve (FP-27-1) to divertfire water to the screenwash system is opened, it is tracked as a fire impairment to the MDFP,and therefore operators are aware and can shut the manual bypass valve to restore fireprotection water flow and pressure to the fire protection header if needed.There are two potential impacts to the use of the fire protection water system for screenwash purposes

-one is a fire within the Screenhouse Fire Area 41, and one for all other fire areas, asdelineated below:For a fire in any fire area except Fire Area 41The plant P&ID drawing indicates that SV-33049 will only open CV-31131 if the 121 MDFP isnot running and the pressure demand on the fire header is not below 90 psi. If there is not ademand on the fire header and the pump is manually

started, the control valve will open andallow fire protection water into the Screenwash system to clean the screens.

If a demand isplaced on the fire header from a suppression system actuation, the pressure drop will cause SV-33049 to close CV-31131 thereby realigning the water flow to the fire protection header.The control cables for SV-33049/CV-31131 run from the MDFP room FA 41 B (elev. 670'screenhouse) to FA 41, (elev. 695' screenhouse).

A fire event in any fire area other than FA 41 will cause a pressure drop on the fire headerthereby closing CV-31131 via SV-33049 and re-aligning the MDFP water supply from thetraveling screens to the fire header.For a fire in Fire Area 41The control cables for SV-33049/CV-31131 run from the MDFP room FA 41 B (elev. 670'screenhouse) to FA 41, (elev. 695' screenhouse).

A postulated fire in FA 41 may cause cabledamage, thereby creating the potential for a hot short of the circuit resulting in CV-31131 failingin the open position.

This would divert some fire protection water from the MDFP away from thefire header. If this scenario occurs and there is a demand on the fire header, the DDFP willstart, providing the fire header with 2,000 gpm at 125 psi. FA 41 suppression system, PA-9, hasa demand of 1,094.1 gpm at 89.1 psi. This demand is within the design capacity of the DDFP.Check valve FP-28-2 will prevent water in the fire protection header from entering theScreenwash diversion piping network.The use of the MDFP for Screenwash cleaning will not impact the ability of the fire protection header to deliver the system demand for fire suppression activities in any plant fire area.Fire protection water to maintain spent fuel pool inventory:

Fire protection water is available through hose stations to supply water to the spent fuel pool.As described below, there are six other preferred sources of water that will be used first tomaintain inventory in the spent fuel pool (SFP). However, in the event that all other watersources are unavailable, fire protection water may be needed to maintain spent fuel poolinventory.

One of the justifications for this non-fire protection use of fire protection water is that it is unlikelythat fire protection water would be needed to simultaneously extinguish a fire using the highestsystem demand and supply makeup water to the spent fuel pool. NUREG/CR-6850, EPRI/NRC-RES, Fire PRA Methodology for Nuclear Power Facilities, states that multipleinitiating events from the same root cause may use a qualitative approach to address thoseinteractions resulting between fire and a seismic event. Consistent with this guidance, aPINGPPage L-3 -Revision 1

Northern States Power -Minnesota Attachment LNFPA 805 Chapter 3 Requirements for Approvalqualitative discussion is provided to address events requiring the simultaneous need forfirewater for fire suppression purposes and for spent fuel pool inventory makeup.Actions to makeup spent fuel pool inventory are well described in an existing abnormal plantprocedure, C16 AOP01, Loss of SFP Inventory.

There are four preferred borated sources formake-up water and two non-borated sources that will be depleted before fire protection water isutilized for spent fuel pool inventory replenishment.

Use of the other six water sources for SFP inventory makeup provides time to respond to thefire and/or get additional sources of equipment and resources.

The existing mutual aidagreement with the City of Red Wing Fire Department and their participation in annual fire drillswill ensure that they can provide a timely response and can also provide the ability toindependently draw water from the river and augment the fire protection water supply andpressure.

In the event that fire suppression water is needed after firewater flow to the SFP is established, control room operators will assess the relative significance of these two needs and balancewater flows as considered appropriate.

Factors in this consideration will include the location, size, and potential significance of the fire to ensuring safe shutdown of the plant, and theamount of water remaining above fuel stored in the spent fuel pool and the rate of change inSFP water level.Nuclear Safety and Radiological Release Performance Criteria:

An evaluation per 10 CFR 50.48, Fire Protection, Section 2.C(vii)(A) is provided below for eachof the requested non-fire protection uses.Fire protection water for screenwash function:

(A) Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological releaseThe ability to use fire protection water for the screenwash function does not impact therequirements of nuclear safety performance criteria as defined in NFPA 805 Section 1.5.1.This use will not directly result in compromising automatic fire suppression functions, manualfire suppression functions, or post-fire safe shutdown capability since water willautomatically or manually be re-directed to the fire protection header upon fire protection header demand. The diesel driven fire pump will also auto start if needed as a result of apressure drop in the fire header to provide for fire protection header demands.

Therefore there is no impact to performance goals or objectives as described in LAR Attachment 2 (B-2 Table).The ability to use fire protection water for the screenwash function has no impact on theradiological release performance criteria since water will be automatically or manually re-directed to the fire protection header upon fire protection header demand. The radiological release review addresses the release of firefighting water potentially containing radioactive materials and is not affected by the use of fire protection water to supplement thescreenwash function.

The ability to use fire protection water for the screenwash functiondoes not change the radiological release evaluation and does not add additional radiological materials to the area or challenge system boundaries.

PINGPPage L-4 -Revision 1

Northern States Power -Minnesota Attachment LNFPA 805 Chapter 3 Requirements for ApprovalFire protection water to maintain spent fuel pool inventory:

(A) Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release;In the unlikely event requiring fire protection water for suppression system use and use as anon-fire protection related source for spent fuel pool inventory makeup, there areprocedures in place and trained operators that will maintain the suppression system demandfor firefighting.

Therefore, nuclear safety performance criteria will not be impacted.

The ability to use fire protection water only after the other six water sources have beenexhausted supports the radiological release performance criteria.

Allowing the SFP waterlevel to decrease could result in spent fuel being uncovered reducing spent fuel decay heatremoval, and creating an extremely hazardous radiation environment.

Therefore, in thisworst case scenario the use of fire protection water for non-firefighting uses is justified toreduce the potential for radiological release.Safety Margin and Defense-in-Depth:

An evaluation per 10 CFR 50.48, Fire Protection, Section 2.C(vii)(B) and Section 2.C(vii)(C) isprovided below for each of the requested non-fire protection uses.Fire protection water for screenwash function:

(B) Maintains safety marginsThe ability to use fire protection water for the screenwash function does not change thesafety margin since it has no impact on safety analysis acceptance criteria or on functional capabilities of equipment relied upon for safe and stable plant conditions.

Due to theautomatic and manual actions to re-direct fire protection water flow, and the ability of theDDFP to supply additional water to the fire header, the use of fire protection water forcleaning intake screens will not impact the ability of the fire protection header to deliver thesystem demand for fire suppression requirements.

(C) Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability).

The ability to use fire protection water for the screenwash function does not compromise thedefense-in-depth

elements, as follows:" Fire prevention functions are maintained because controls such as combustible controls and hot work controls are not affected.

" Fire detection, automatic fire suppression, manual fire suppression, and mitigation functions are maintained because water will automatically or manually be re-directed to the fire protection header, and water from the DDFP will be supplied if needed,upon actuation of a fire suppression system in the event of a fire.* Post-fire safe shutdown capability is maintained because fire barriers are maintained so that a fire will not spread and prevent operation of the equipment required toestablish and maintain safe and stable plant conditions.

PINGP Page L-5 -Revision 1PINGPPage L-5 -Revision I

Northern States Power -Minnesota Attachment LNFPA 805 Chapter 3 Requirements for ApprovalFire protection water to maintain spent fuel pool inventory:

(B) Maintains safety margins:In the unlikely event of a fire and fuel pool loss of inventory, there will be no initial impact tothe safety margin of the fire protection water supply since there will be no immediate diversion of water to the spent fuel pool. There are multiple preferred borated and non-borated water sources that will be procedurally utilized prior to using fire suppression waterfor spent fuel pool inventory control.

If the preferred sources of water are depleted (orunavailable) and fire protection water must be used to maintain spent fuel pool inventory, procedures in place and trained operators will maintain the suppression system demanduntil fire extinguishment.

(C) Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability).

The ability to use fire protection water for the SFP makeup function does not compromise the defense-in-depth

elements, as follows:" Fire prevention functions are maintained because controls such as combustible controls and hot work controls are not affected.

• Fire detection, automatic fire suppression, manual fire suppression, and mitigation functions are maintained because fire protection water will only be used for SFPmakeup after all other sources of water are depleted or are otherwise unavailable, atwhich time fires will likely have been extinguished; also, by this time other sources offire protection water, e.g., support from the Red Wing Fire Department, should alsobe available if needed in the event of a fire.* Post-fire safe shutdown capability is maintained because fire barriers are maintained so that a fire will not spread and prevent operation of the equipment required toestablish and maintain safe and stable plant conditions.

Operators would balance the need to supply firefighting water and maintain SFP level. Theadditional time provided by the six other water supplies would allow additional equipment andresources to arrive and support the fire protection water supply demand. There will be noimpact to fire prevention, fire detection, automatic fire suppression functions, manual firesuppression functions, mitigation or post-fire safe shutdown capability.

Conclusion The use of fire water to perform the screenwash function and to maintain spent fuel poolinventory does not meet the requirement or allowed exceptions of NFPA 805 Section 3.5.16.The evaluation determined that the performance-based approach utilized to evaluate a variancefrom the requirements of NFPA 805 Chapter 3:(A) Satisfies the performance goals, performance objectives, and performance criteriaspecified 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, firesuppression, mitigation, and post-fire safe shutdown capability).

Therefore, the risk associated with the use of fire protection water to provide the screenwash function and maintain spent fuel pool inventory is low and PINGP can maintain safe and stablePINGPPage L-6 -Revision 1

Northern States Power -Minnesota Attachment LNFPA 805 Chapter 3 Requirements for Approvalconditions.

NRC approval is being requested to permit the fire water system to be aligned forthe screenwash function and to maintain spent fuel pool inventory.

PINGP Page L Revision 1PINGPPage L-7 -Revision 1

L-PI-15-041 Enclosure 4NSPMEnclosure 4LAR Attachment T (revised)

The following pages include a revision toLAR Attachment T, Clarification of Prior NRC Approvals, and includes information in response to SSA RAI 02.3 pages follow Northern States Power -Minnesota Attachment T- Clarification of Prior NRC Approvals T. Clarification of Prior NRC Approvals 2 Pages AttachedPINGP Page T-1 -Revision 1PINGPPage T-1 -Revision I

Northern States Power -Minnesota Attachment T -Clarification of Prior NRC Approvals Introduction The elements of the pre-transition fire protection program licensing basis for which specific NRCprevious approval is uncertain are included in this attachment.

Also included is sufficient detailto demonstrate how those elements of the pre-transition fire protection program licensing basismeet the requirements in 10 CFR 50.48(c)

(RG 1.205, Revision 1, Regulatory Position 2.2.1).Prior Approval Clarification Request I of 1: Operator Action to Isolate Power toPORV Control CircuitsPre-transition Fire Protection Program Licensing Basis:The Prairie Island Nuclear Generating Plant (PINGP) pre-transition licensing basis relative tothe preclusion of spurious operation of pressurizer power operated relief valve (PORV) flowpaths, for fires involving control room evacuation, included a previously approved exemption from the requirements of Section III.G.1 of Appendix R to 10 CFR 50. Specifically, theexemption allowed operators to close the Unit 1 and 2 PORV block valves prior to evacuating the control room, and then taking the follow-on action to remove control power fuses from thePORV control circuits for both units at their respective branch circuit panels.The exemption was required because the removal of fuses involved the use of a fuse-pulling tool, which was considered to be a "repair action."

This repair action was interpreted as a non-compliance to Section Ill.G.1 of Appendix R to 10 CFR 50 which requires, in part, that fireprotection features shall be provided for structures,

systems, and components important to safeshutdown so that one train of systems necessary to achieve and maintain hot shutdownconditions be free of fire damage.This exemption was approved in a letter dated February 21, 1995. In 1999, PINGP performed aplant modification (99DC03),

which included modification of the PORV control power suppliessuch that disconnect switches could be used in lieu of pulling control power fuses. Thefeasibility of utilizing the disconnect switches (no tool required) has been validated and hasproven to be a beneficial change with respect to this activity.

Background/Basis:

NSP Exemption Request Letter, dated May 2, 1994NSP requested an exemption from the requirements of Section III.G.2 of Appendix R to 10 CFR50, to allow the manual removal of fuses from the PORV control circuits in the event of a fire, inlieu of modifying plant hardware.

The reference to III.G.2 was later revised to III.G.1 during afollow-up phone call between NSP and NRR.Issuance of Exemption Letter, dated February 21, 1995The NRC issued an exemption from certain requirements of Appendix R to 10 CFR Part 50 toallow NSP to remove fuses from the PORV control circuits as a means of ensuring the reactorcoolant system inventory in the event of a control room fire.PINGP Page T-2 -Revision 1PINGPPage T-2 -Revision 1

Northern States Power -Minnesota Attachment T -Clarification of Prior NRC Approvals PINGP Plant Modification (99DC03)Summary:This modification relocated EQ circuit power supplies from harsh environments to mildenvironments.

This modification repowered the Unit 1 and 2 PORV control circuits, from newdistribution panels PNL 171, PNL 181, PNL 271, and PNL 281 respectively, which were, in turn,powered by upstream feeder distribution panels PNL 11, PNL 12, PNL 21, and PNL 22respectively.

An added benefit of this modification is that it allowed the PORV control circuits tobe de-energized via disconnect switches in the feeder distribution panels, thus eliminating theneed to pull control power fuses for fire events requiring control room evacuation.

RequestAs part of this LAR submittal and transition to NFPA 805, it is requested that the NRC acceptthe following clarification of a prior NRC approval, with respect to the exemption granted to NSPon February 21, 1995:This operator action (recovery action) to preclude PORV opening remains a requiredaction for PINGP under NFPA 805. The use of recovery actions is not allowed underthe deterministic requirements of NFPA 805 Section 4.2.3.1.

Clarification is requested toallow the previous exemption for operator actions to be extended to the NFPA 805program.In addition, clarification is requested to extend the previous allowance to pull fuses toinstead allow the operation of disconnect switches.

The manual operation to opendisconnect

switches, demonstrated by PINGP to be feasible and reliable, is simpler thanpulling fuses and therefore, for the purposes of this request, is requested to be deemedequivalent in intent and function.

Clarification is also requested that the term "control room fire", as referred to in theexemption letter, applies to fires occurring in Fire Area 013 (Control Room) and FireArea 018 (Relay Room). Under the pre-transition (Appendix R) program, both Fire Area013 and Fire Area 018 were analyzed as one analysis area.PINGP Page T-3 -Revision 1PINGPPage T-3 -Revision 1