W3F1-2015-0015, Responses to Request for Additional Information Regarding Adoption of National Fire Protection Association Standard NFPA 805 License Amendment Request

Responses to Request for Additional Information Regarding Adoption of National Fire Protection Association Standard NFPA 805 License Amendment Request

ML15071A296
Person / Time
Site:	Waterford
Issue date:	03/12/2015
From:	Chisum M Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC ME7602, W3F1-2015-0015
Download: ML15071A296 (51)

Text

W3F1-2015-0015March 12, 2015U.S. Nuclear Regulatory CommissionAttn: Document Control DeskWashington, DC 20555-0001

SUBJECT:

Responses to Request for Additional Information Regarding Adoption ofNational Fire Protection Association Standard NFPA 805 LicenseAmendment Request (LAR) Waterford Steam Electric Station, Unit 3(Waterford 3)Docket No. 50-382License No. NPF-38REFERENCES

1. Entergy letter W3F1-2011-0074 "License AmendmentRequest to Adopt NFPA 805 Performance-Based Standardfor Fire Protection for Light Water Reactor Generating Plants(2001 Edition)", Waterford Steam Electric Station, Unit 3dated November 17, 2011 [ML113220230]2. Entergy letter W3F1-2012-0005 "Supplemental Information inSupport of the NRC Acceptance Review of Waterford 3License Amendment Request to Adopt NFPA 805, WaterfordSteam Electric Station, Unit 3" dated January 26, 2012[ML12027A049]3. Entergy letter W3F1-2012-0064 "Response to Request forAdditional Information Regarding Adoption of National FireProtection Association Standard NFPA 805 LicenseAmendment Request, Waterford Steam Electric Station, Unit3" dated September 27, 2012 [ML12272A099]4. Entergy letter W3F1-2012-0083 "90 Day Response toRequest for Additional Information Regarding Adoption ofNational Fire Protection Association Standard NFPA 805License Amendment Request, Waterford Steam ElectricStation, Unit 3" dated October 16, 2012 [ML12290A216]5. Entergy letter W3F1-2013-0022 "Response to 2 nd RoundRequest for Additional Information Regarding Adoption ofNational Fire Protection Association Standard NFPA 805License Amendment Request, Waterford Steam ElectricStation, Unit 3" dated May 16, 2013 [ML13137A128]Entergy Operations, Inc.17265 River RoadKillona, LA 70057-3093Tel 504-739-6660Fax 504-739-6698mchisum@entergy.comMichael R. ChisumVice President - OperationsWaterford 3 W3F1-2015-0015Page 26. Entergy letter W3F1-2013-0048 " Supplement to NFPA 805License Amendment Request (LAR) Waterford Steam ElectricStation, Unit 3" dated December 18, 2013 [ML13365A325]7. NRC letter to Entergy dated February 6, 2015,"Request forAdditional Information RE: License Amendment Request toTransition to National Fire Protection Association Standard 805(TAC NO. ME7602) [ML15022A239]

Dear Sir or Madam:

By letter dated November 17, 2011, as supplemented by letters dated January 26,September 27, October 16, 2012, May 16, 2013, and December 18, 2013 (References 1through 6 respectively), Entergy Operations, Inc. (Entergy), submitted a license amendmentrequest (LAR) to transition its fire protection license basis at the Waterford Steam ElectricStation, Unit 3, from paragraph 50.48(b) of Title 10 of theCode of Federal Regulations(10CFR) to 10 CFR 50.48(c), "National Fire Protection Association Standard 805" (NFPA 805).The LAR Supplement provided in Reference 6 represents changes to specified LARAttachments and supporting calculations primarily as a result of performing extensivereanalysis utilizing only NRC-accepted methods. An NRC site audit was conducted the weekof January 12, 2015 followed by Request for Additional Information (RAI) letter (Reference 7)received February 6, 2015. These RAIs were divided into 60, 90 and 120 day responses.Enclosure 1 contains responses to the 60 day RAIs.There are no new regulatory commitments contained in this submittal.If you require additional information, please contact the Regulatory Assurance Manager,John Jarrell at 504-739-6685.I declare under penalty of perjury that the foregoing is true and correct. Executed onMarch 12, 2015.Sincerely,MRC/ajh

Enclosures:

1. 60 Day RAI Responses2. Revised Attachment A - NFPA 805 Chapter 3 Requirement 3.11.3 W3F1-2015-0015Page 3cc: Marc L. DapasRegional AdministratorU. S. Nuclear Regulatory CommissionRegion IV1600 E. Lamar Blvd.Arlington, TX 76011-4511RidsRgn4MailCenter@nrc.govNRC Senior Resident InspectorWaterford Steam Electric Station Unit 3P.O. Box 822Killona, LA 70066-0751Frances.Ramirez@nrc.govU. S. Nuclear Regulatory CommissionAttn: Mr. Michael OrenakMail Stop 8-G9AWashington, DC 20555-0001Michael.Orenak@nrc.govLouisiana Department of Environmental QualityOffice of Environmental ComplianceSurveillance DivisionP.O. Box 4312Baton Rouge, LA 70821-4312Ji.Wiley@LA.gov toW3F1-2015-001560 Day RAI Responses Waterford 3 NFPA 805 License Amendment Request to W3F1-2015-0015Page 1 of 44FPE RAI S01LAR Attachment A, identifies both "open items" and "confirmatory items" that are necessary toachieve compliance with NFPA 805. In the RAI dated July 18, 2012 (ADAMS Accession No.ML12185A212), FPE RAI 07 had requested the licensee to confirm which "open items" are closedand which should be included as implementation items in the licensee's LAR applicationAttachment S. By letter dated September 27, 2012, the licensee responded to FPE RAI 07 andstated that "confirmatory items" are identified as implementation items because these items aresolely required for NFPA 805 implementation. The licensee also stated that "open items" arethose required for the current Appendix R licensing basis, as well as NFPA 805, and must becorrected regardless of transitioning to NFPA 805. These "open items" are tracked in the plantcorrective action program. The licensee further stated that the "open items" are considered asrequired actions to transfer to NFPA 805, but do not rise to the level of being categorized as"confirmatory items" needing to be listed in Attachment S.Since the "open items" are required to comply with NFPA 805, and LAR Attachment S is used todocument items necessary to complete NFPA 805 implementation, per the proposed licenseecondition, provide the following:

a.A revised LAR Attachment S that includes these "open items" as implementationitems, or justification for their exclusion.

b.LAR Section 4.1.2.1, "NFPA 805 Chapter 3 Requirements Met or Previously Approvedby the NRC," indicates that the "confirmatory items" and "open items" are associatedwith the "Complies" compliance statement. However, in the LAR, Attachment A,several of these items are included with elements that state "Complies with use ofEEEEs [Existing Engineering Equivalency Evaluations]." For these cases, whichinclude NFPA 805 Chapter 3 Sections: 3.3.1.2(5), 3.3.7.1, 3.4.1(a)(1), 3.5.3, 3.7,3.8.1, 3.8.2, 3.9.1(1), 3.9.1(2), 3.11.3(2), and 3.11.5, clarify if these "confirmatoryitems" and "open items" are necessary to comply directly with the NFPA 805 Chapter 3sections, or if they are associated with the EEEEs that are cited in the compliancebasis.Waterford 3 Responsea. Attachment S is tracking the 'Open' items given in Attachment A with the exception ofthe following NFPA 805 Chapter 3 sections: 3.3.1.2(5), 3.3.5.1, 3.3.10, 3.3.11,3.3.12(1), 3.5.3-1, 3.7, 3.9.1(1), 3.9.1(2)-1, 3.9.1(2)-3, and 3.11.3(2). All of these'Open' items have been completed with the exception of NFPA 805 Chapter 3Sections 3.3.5.1 and 3.3.11. These remaining 'Open' Items are procedure/documentupdates being tracked by Attachment S, Item S2-13.

b.These "confirmatory and open items" are necessary to comply directly with theaffected NFPA 805 Chapter 3 Section. to W3F1-2015-0015Page 2 of 44FPE RAI S0290 day responseFPE RAI S03LAR Attachment A, Section 3.4.1(c) states compliance with NFPA 805 Section 3.4.1(c) whichrequires that the fire brigade leader and at least two brigade members have sufficient training andknowledge of nuclear safety systems to understand the effects of fire and fire suppressants onnuclear safety performance criteria. In RG 1.189, "Fire Protection for Nuclear Power Plants,"Revision 2, dated October 2009 (ADAMS Accession No. ML092580550), Section 1.6.4.1"Qualifications," the NRC staff acknowledged the following example for the fire brigade leader assufficient that states, in part:The brigade leader should be competent to assess the potential safetyconsequences of a fire and advise control room personnel. Such competence bythe brigade leader may be evidenced by possession of an operator's license orequivalent knowledge of plant systems.Provide additional detail regarding the training provided to the fire brigade leader and brigademembers that addresses their ability to assess the effects of fire and fire suppressants onNFPA 805 nuclear safety performance criteria. Include the justification for how the training meetsNFPA 805 Section 3.4.1.Waterford 3 ResponseAll Waterford 3 Fire Brigade Members are part of the Operations Department and are NuclearAuxiliary Operators (NAO). The Fire Brigade Leader and two of the Fire Brigade Members onevery Operations Shift are at least Level A NAO's who are qualified on Safety Systems. TheseFire Brigade Members are qualified to stand watch in the Reactor Auxiliary Building (RAB)including the Radiation Controlled Area (RCA). As such, they are qualified on all aspects of theoperation of every safety system including the potential impact of firefighting activities on thosesystems. All Level A qualified NAO's are qualified to be Fire Brigade Leaders (FBL) andHazardous Material (HAZMAT) incident commanders.Each Fire Brigade Leader and Fire Brigade Member, as qualified Operations watch standers andradiation workers, are physically qualified to stand watch including use of respiratory equipmentused during firefighting, HAZMAT and Emergency Planning events.Fire Brigade Training is provided in accordance with procedure NTP-202, Fire Protection Training,Attachment 7.6 and lesson plan training documents (WLP-FPFB-FFR04, WLP-FPFB-IFB01 andWLP-FPFB-FBL02). This training ensures the Waterford 3 Fire Brigade Leaders and Fire BrigadeMembers meet the requirements of Regulatory Guide 1.189, Rev. 2, [Section 1.6.4.1,Qualifications] and LAR Attachment A, Section 3.4.1(c). These training documents and theguidance contained within EN-OP-115, Conduct of Operations and OI-042-000, Watch StationProcess ensure the Fire Brigade's ability to assess the effects of fire and fire suppressants on theoperability of safety related systems. Both the initial and requalification programs for Fire Brigade to W3F1-2015-0015Page 3 of 44Members and Leaders are rigorous accredited programs conducted in compliance with EN-TQ-201, Systematic Approach to Training Process.The site Fire Brigade training coordinator is a qualified instructor responsible for managing FireBrigade training and education. This includes: Classroom (initial and requalification) Fire Field Practical (initial and requalification) Conduct of Fire Drills.FPE RAI S04NFPA 805 Section 3.4.1(a) requires a minimum fire brigade staff of five persons on duty at alltimes. LAR Attachment A, the licensee states it complies with clarification that the fire brigademay be less than the minimum complement of five persons for a maximum of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> toaccommodate unexpected absences. As described in NEI 04-02, Section 4.3.1, "Complies withClarification," statements are intended to address compliance differences that are generallyeditorial in nature. The less-than-minimum staffing statement in the LAR is not considerededitorial.NFPA 805 states that previously approved alternatives to the fundamental fire protection programattributes, identified in Chapter 3, take precedence over the requirements in NFPA 805, Chapter 3.Clarify whether the NRC has previously approved less-than-minimum fire brigade staffing forWaterford, for this 2-hour grace period, and if so, confirm that this approval has been assessed for"continued applicability and validity." If not, specifically request NRC staff approval of thiscondition and provide adequate justification.Waterford 3 ResponseThe current Waterford 3 Fire Brigade Staffing requirement, including the two hour allowance forless than full staffing, was previously contained in Waterford 3's Technical Specifications (TS) andthus had received commission approval. Amendment 50 to the Waterford 3 Operating LicenseNPF-38, dated 02/07/1989 approved removing this requirement from Technical Specification, tobe controlled by the site (currently contained in Operations Procedure OI-042-000, "Watch StationProcesses"). The intent of the Fire Brigade Staffing requirements is identical to that previouslycontained in TS. The need for this exception is unchanged from the original TechnicalSpecification allowance and is consistent with the similar exception for Operations and RadiationProtection Shift staffing currently contained in Technical Specifications Section 6.0, Table 6.2-1,Minimum Shift Crew Composition. to W3F1-2015-0015Page 4 of 44FPE RAI S05LAR Attachment A, Element 3.11.3(2) refers to LAR Attachment T, for clarification in theCompliance Basis column. In the LAR supplement dated December 18, 2013, the content of LARAttachment T was reported as "deleted." Clarify the compliance basis for NFPA 805 Element3.11.3(2).Waterford 3 ResponseNFPA 805 Element 3.11.3(2) in Attachment A of the LAR Supplement is revised per Enclosure 2.The revision includes the removal of Attachment T as a reference and updates Element 3.11.3(2)to include Attachment K, Deviations 16 and 42, which were omitted in the original LARAttachment A. Also included in the revision is removal of VFDR 3.11.3 (2) which was closed viaEC 39570. to W3F1-2015-0015Page 5 of 44SSA RAI S01By letter dated December 18, 2013, the LAR supplement provided a revised Attachment S.Clarify or provide a revised status regarding the following Implementation Items:a. The updated response in the licensee's letter dated June 11, 2014, to RAI SSA 12 statesthat Implementation Item S2-17 has been revised to ensure that all feasibility criteria inFAQ 07-0030, "Establishing Recovery Actions," are addressed for Nuclear SafetyCapability Assessment (NSCA) recovery actions (RAs). The updated responseto RAI SSA 08.01, however, states that the only RAs in Attachment G of the revisedLAR are associated with tripping Reactor Coolant Pumps, and the RAs were confirmed tobe feasible based on all 11 criteria of FAQ 07-0030. From the above RAI responses, it isunclear if all RAs feasibility analysis to FAQ 07-0030 have been completed or if there willbe some that are completed during the implementation period. Confirm the time periodfor the feasibility analyses for Waterford NSCA and non-power operations RAs that arecompleted and those to be completed per Implementation Item S2-17.b. Implementation Item S2-20 also involves evaluating a revised list of RAs for feasibilityusing the criteria of FAQ 07-0030 (see Item a. above) and revising LAR Attachments C,G, S, and W. This implementation item references SSA RAIs 08.01 and 13. However,the licensee's letter dated June 11, 2014, indicates that the actions described in thisimplementation item have been completed. Clarify the status of which items have beenrevised and which items remain to be revised per Implementation Item S2-20.Waterford 3 ResponseThe status of the LAR Attachment S, Table S-2, Implementation Items S2-17 and S2-20 areidentified below:a. LAR Attachment S, Table S-2, Implementation Item S2-17 action to update the recoveryfeasibility process against the 11 criteria of FAQ 07-0030 which includes the incorporationof drills into the fire protection program has been completed and is documented inEngineering Report WF3-FP-13-00003, Recovery Action Feasibility & Reliability Review.b. LAR Attachment S, Table S-2, Implementation Items S2-20 action to evaluate the revisedlist of Recovery Actions and revise the LAR Attachments C, G, S, and W has beencompleted. A Recovery Action (RA) is required in four fire areas (RAB 1, RAB 7, RAB 8and TGB) to trip the Reactor Coolant Pumps (RC-MPMP-0001A, B, 2A & 2B.) Thefeasibility of the RA in each of the affected Fire Areas using the criteria of FAQ 07-0030has been performed and is documented in Engineering Report WF3-FP-13-00003,Recovery Action Feasibility & Reliability Review. LAR Attachments C, G, S and W reflectthis information. to W3F1-2015-0015Page 6 of 44SSA RAI S02By letter dated December 18, 2013, the LAR supplement, Attachment C, VFDR 1-045 does nothave a disposition. Provide the disposition of this variance from deterministic requirements (VFDR).Waterford 3 ResponseThe disposition of VFDR 1-045 in EC-F10-002, WF3 Fire Area RAB-1 Fire Risk Evaluation, is:This condition was evaluated for compliance using the performance-based approach of NFPA805, Section 4.2.4. A fire risk evaluation determined that applicable risk, defense-in-depth, andsafety margin criteria were satisfied without further action.SSA RAI S03By letter dated December 18, 2013, the LAR supplement, Attachment G, the recovery action forRAB 7 references VFDR 7-074. In Attachment C of the LAR supplement, the recovery action forRAB 7 references VFDR 7-070. Confirm which VFDR should be referenced.Waterford 3 ResponseThe correct reference in Attachment G for the recovery action for RAB 7 is VFDR-070.SSA RAI S04By letter dated December 18, 2013, the LAR supplement, Attachment C, Tables C-1 and C-2,identifies the Fire Suppression and Detection Systems required for compliance by an EEEE or tomeet a licensing action. In particular, Fire Areas RAB 15A, RAB 16A, RAB 40 and RAB 41 eachreference "Attachment A, Section 3.3.8." However, the LAR supplement revised a number ofAttachment A, Section 3.3.8, compliance statements from "Complies with EEEE" and "Compliesby Previous Approval," to "Complies." Therefore, in Attachment C, clarify the basis for therequired Fire Suppression and Detection Systems to align with Attachment A.Waterford 3 ResponseThe Diesel Fuel Oil Tanks in Fire Areas RAB 15A, RAB 16A, RAB 40 and RAB 41 are part of theplant's diesel supply system. These tanks are inside the power block, however these oil tanks arenot for bulk storage and are not applicable to NFPA 805 Section 3.3.8, Bulk Storage.For Fire Areas RAB 15A and RAB 16A, the Required Fire Protection Systems and FeaturesPages (275 and 282) are removed from Enclosure 2 to W3F1-2013-0048 LAR Attachment CTable C-1. The basis for the required Fire Detection and Suppression provided inAttachment C, Table C-2 Pages (5 and 6) is changed from "L" to "NONE". References toLAR Attachment A, NFPA 805 Section 3.3.8 are removed from this table for these FireAreas. For Fire Area RAB 15A, the Required Fire Protection Feature Page (5) is changedfrom "E" to "NONE". to W3F1-2015-0015Page 7 of 44For Fire Areas RAB 40 and RAB 41, the Required Fire Protection Systems and FeaturesPages (421 and 423) are removed from Enclosure 2 to W3F1-2013-0048 LAR Attachment CTable C-1. The basis for the required Fire Suppression provided in Attachment C, Table C-2 Page (11) is changed from "L" to "NONE". References to LAR Attachment A, NFPA 805Section 3.3.8 are removed from this table for these Fire Areas..SSA RAI S05In the LAR, Section 4.8.1 and Attachment C, the licensee designates the basis for requiring a fireprotection system or feature as follows: S (Separation), L (Licensing Action), E (EEEE Criteria), R(Risk Criteria), and D (Defense-in-Depth). LAR Attachment C, Table C-2, provides thecompilation of required systems and features for all fire areas and follows the Section 4.8.1designations described above. However, in LAR Attachment C, Table C-1, systems or featuresrequired for licensing action or EEEE in individual fire areas are designated "LA/EEEE." Becausethe LA/EEEE designation in Table C-1 does not follow the same convention as Section 4.8.1 andTable C-2, confirm the designations (i.e., S, L, E, R, D) for the required features and systems inTable C-2 are correct and appropriate for use in determining if the systems and features arerequired for either a licensing action or an EEEE.Waterford 3 ResponseThe software tool utilized to prepare LAR Attachment C-1 only allowed the choice of "EEEE/LA"; ifa fire protection feature/system was required for the "Existing Engineering EquivalencyEvaluations" and/or "Licensing Actions". However, LAR Attachment C-2 provides further detail bydesignating the requirement for the specific fire protection feature/system as "E" for "ExistingEngineering Equivalency Evaluations" and/or "L" for Licensing Actions" for the applicable fireareas.SSA RAI S06By letter dated December 18, 2013, the LAR supplement, Attachment G, states, "Once control ofthe plant has been established at the primary control station (PCS), operators will use availablecontrols as dictated by operations procedures to maintain the plant in a safe and stable condition.The location of the fire and impact on specific equipment will determine which equipment andactions will be available to the operators." Provide the following clarifications:

a.Clarify whether additional local actions are necessary outside the PCS (i.e., remoteshutdown station) to maintain safe and stable conditions once control of the plant hasbeen established at the PCS.

b.If additional actions are necessary to control the plant beyond those performed at thePCS, describe these actions and either: 1) include them in a revised Table G-1 andassess the risk of the additional actions, or 2) provide justification for not includingthese actions in Table G-1. to W3F1-2015-0015Page 8 of 44Waterford 3 Response a.There are no additional local actions outside the PCS required to maintain plant safeand stable other than the actions listed in Attachment G.

b.Not Applicable. toW3F1-2015-0015Page 8 of 44FM RAI S01NFPA 805, Section 2.4.3.3 states that the Probabilistic Risk Assessment (PRA) approach,methods, and data shall be acceptable to the NRC. The NRC staff noted that fire modelingcomprised the following: The algebraic equations implemented in FDTs [Fire Dynamics Tools] were used tocharacterize flame radiation (heat flux), flame height, plume temperature, ceiling jettemperature, and hot gas layer (HGL) temperature, the latter of which is used in the multi-compartment analysis. The Consolidated Model of Fire Growth and Smoke Transport (CFAST) was used to assessmain control room (MCR) habitability and to calculate HGL temperature in selected multi-compartment scenarios.LAR Section 4.5.1.2, "Fire PRA" states, in part, that "[f]ire modeling was performed as part of theFire PRA development (NFPA 805 Section 4.2.4.2)." Reference is made to LAR Attachment J,"Fire Modeling V&V [verification and validation]," for a discussion of the acceptability of the firemodels that were used.Regarding the acceptability of the PRA approach, methods, and data:

a.Identify whether any fire modeling tools and methods have been used in the developmentof the LAR that are not discussed in LAR Attachment J. Conversely, identify any firemodeling tools and methods discussed in LAR Attachment J that have not been used inthe fire modeling analyses performed at Waterford.

b.90 day response c.90 day response d.Typically, during maintenance or measurement activities in the plant, electrical cabinetdoors are opened for a certain period of time. Explain what administrative controls are inplace to minimize the likelihood of fires involving such an open cabinet, and describe howcabinets with temporarily open doors were treated in the fire modeling analyses.

e.90 day response f.For calculation of fixed ignition source ZOIs, the fire area for all cabinet fires was fixed at 0.5 m 2. Justify why using a fixed area is representative of all fixed ignition sources in theplant or demonstrate that the findings from this analysis are not sensitive to the fire areasize or that the obtained results are bounding.In addition, provide justification for the assumed fire areas and elevations that were usedin the transient ZOI calculations. Explain how the model assumptions in terms of locationand HRR of transient combustibles in a fire area or zone will not be violated during andpost-transition.

g.90 day response toW3F1-2015-0015Page 9 of 44 h.Specifically, regarding the use of CFAST in the MCR abandonment calculations:

i.120 day response ii.120 day response iii.90 day response iv.Provide the technical basis for the material properties that were specified inCFAST for the transient combustibles in the MCR. Provide confirmation that theassumed soot yield and heat of combustion values are representative of thetransient materials that are present in the MCR, or lead to conservative estimatesof the soot generation rate.

v.It appears that the cables in the electrical cabinets are assumed to be an equalmix of ethylene propylene rubber (EPR), Hypalon, and Neoprene. Confirm thatthis assumption is consistent with the actual cable mixture present in the plant.Provide the technical basis for the material properties that were specified inCFAST for the cables inside the cabinets in the MCR. Provide confirmation thatthe assumed soot yield and heat of combustion values lead to conservativeestimates of the soot generation rate.

i.Specifically, regarding the use of CFAST in the RAB 7A, 7B, 7C and 7D calculations:

i.90 day response ii.120 day response iii.90 day response iv.Temperature values obtained from the CFAST analysis were used to determinewhether a cabinet or sub-PAU [Physical Analysis Unit] fails. Clarify where(location within the surface or volume of the cabinets) these temperature valueswere recorded and provide the basis for selecting these particular locations.

j.Specifically regarding the multi-compartment analysis (MCA):

i.Describe the criteria that were used to qualitatively screen multi-compartment scenarios.

ii.Explain how the methods described in Chapter 2 of NUREG-1805, "Fire DynamicsTools (FDTs) Quantitative Fire Hazard Analysis Methods for the U.S. NuclearRegulatory Commission Fire Protection Inspection Program," dated December2004 (ADAMS Accession No. ML043290075) were used in the calculations toscreen an ignition source based on insufficient HRR to generate a HGL conditionin the exposing compartment.

iii.Explain how the possibility of damaging hot gases spreading to a thirdcompartment was considered. toW3F1-2015-0015Page 10 of 44Waterford 3 ResponseFM RAI S01 a.Attachment J of the Waterford 3 LAR Supplement (Waterford 3 Letter W3F1-2013-0048) was re-reviewed in response to this RAI, and the only instance of a fire modeling tool or method being usedand not presented in Attachment J is the use of NUREG-1805 Fire Dynamics Tools (FDTs)spreadsheet 05.1_Heat_Flux_Calculations_Wind_Free.xls (v. 1805.0) Solid Flame 1 tool whichcalculates the heat flux at a specific distance from a fire source. This tool and its methods are usedin the fire PRA to generate a comparison plot of various heat flux calculational methods in the fixedignition source and transient ignition source methodology reports PRA-W3-05-006F and PRA-W3-05-013, respectively. This tool is not used in the heat flux level calculations for ignition source zoneof influence (ZOI) analyses which uses the Solid Flame 2 tool of the same spreadsheet and isalready documented in Attachment J of the Waterford 3 LAR Supplement.The prediction tool for calculating heat flux at a specific distance from a fire source for wall andcorner locations (NUREG-1805 FDTs spreadsheet 05.1_Heat_Flux_Calculations_Wind_Free.xls (v.1805.0) Point Source Method tool) is listed in Attachment J of the LAR as being used to generate acomparison plot of various heat flux calculational methods. However, this comparison plot is notused in the fixed ignition source or transient ignition source methodology reports PRA-W3-05-006For PRA-W3-05-013, respectively.Additionally, hot gas layer (HGL) temperature predictions using the methods of NUREG-1805 FDTsspreadsheets 02.1_Temperature_NV.xls (v. 1805.0), listed in Attachment J of the LAR Supplementas the Method of McCaffrey, Quintiere, and Harkleroad (MQH method) which calculates the HGLtemperature in a compartment with natural ventilation (i.e. an open compartment) is listed as apossible tool for use based on physical analysis unit (PAU) arrangements. This FDT is listed as anoption for an analyst's use when a HGL prediction is needed in the methodology for zone ofinfluence (ZOI) determination for the various fixed ignition source bins as presented in PRA-W3-05-006F and for the transient ignition source bin as presented in PRA-W3-05-013. However, this HGLtool is subsequently not used in the Waterford 3 fire PRA analyses of fixed or transient ignitionsources.Similarly, HGL temperature predictions using the methods of NUREG-1805 FDTs spreadsheets02.2_Temperature_FV.xls (v. 1805.0), listed in Attachment J of the LAR Supplement as the Methodof Foote, Pagni, and Alvares (FPA) and the Method of Deal and Beyler, which calculates the HGLtemperature in a compartment with forced ventilation is listed as a possible tool for use based onPAU arrangements. This FDT is also listed as an option for an analyst's use when a HGL predictionis needed in the methodology for zone of influence (ZOI) determination for the various fixed ignitionsource bins as presented in PRA-W3-05-006F and for the transient ignition source bin as presentedin PRA-W3-05-013. However, these HGL tools are subsequently not used in the Waterford 3 firePRA analyses of fixed or transient ignition sources.A revised Attachment J of the LAR Supplement will be provided with the 120 day RAI responses.FM RAI S01 d.Administrative controls to minimize likelihood of fires involving open cabinets include writtenprocedure guidance in EN-IS-123, Electrical Safety; EN-OP-115, Conduct of Operations; EN-MA-101, Conduct of Maintenance and EN-MA-118, Foreign Material Exclusion and EN-DC-127, Controlof Hot Work and Ignition Sources.Electrical and instrument cabinet doors remain closed at Waterford 3. Procedures require stationpersonnel to ensure equipment doors and covers designed to be closed remain so. Operations toW3F1-2015-0015Page 11 of 44procedures require Operations personnel to 1) be alert for any equipment, including electrical andinstrument doors, panels and covers, to be in their normal configuration, 2) report any variances fromthis standard, and 3) correct the deficiency.Written guidance exists to prevent ignition of fires within electrical and instrument cabinets whichmust remain open to support work. Any electrical or instrument cabinet which must remain open toconduct work or testing is controlled by procedure or work order. These procedures and work ordersinclude precautions on restricting access to the cabinet interiors by signage and barricading,preventing accumulation of combustible materials near the open cabinet through good workpractices and housekeeping, and close out of cabinets to ensure no tools or foreign material remainwithin the cabinet which might trigger fire ignition.Any hot work, including hot work within or near open electrical and instrument enclosures, isconducted in accordance with approved procedures and has received required Fire Protectionpersonnel review.In the fire modeling analyses, all electrical panel (cabinet) doors were treated as closed. Even if acabinet was found to be sealed (few instances exist), all of the cabinets modeled in the FPRA areconservatively considered vented. The associated heat release rate is based on the presence ofthermoset (qualified) or thermoplastic (unqualified) cables with the applicable single or multiple cablebundle characteristic for the particular cabinet being analyzed.FM RAI S01 fThe zone of influence (ZOI) analyses for each fixed ignition source bin as well as for transientignition sources are based on the noted 0.5 m 2 fire source area. The methodology and resultantZOIs for the various fixed ignition source bins are presented in PRA-W3-05-006F . The methodologyand resultant ZOI for the transient ignition source bin is presented in PRA-W3-05-013.As noted in those reports, the fire size (area) is selected for several reasons. First, when heatrelease rate (HRR) per unit area (HRRPUA) is considered, a fire source must have a HRRPUAgreater than 600 kW/m 2 (52.9 Btu/s-ft²) in order to remain within the NUREG/CR-1824 validationbasis fire Froude Number range. Using a fire area of 0.5 m 2 results in a HRRPUA of 300 kW / 0.5 m 2 (or 600 kW/m 2), falling within the validation range of NUREG/CR-1824 for any ignition source binHRR equal to or greater than 300 kW. Second, FDTs are sensitive to the fire area used and the useof 0.5 m 2 compared to a larger area such as 1.0 m 2 serves to produce more restrictive orconservative (bounding) outputs from the FDTs by predicting higher plume temperatures at the sameelevation above the fire source for a smaller fire source area as compared to a larger fire sourcearea. Third, it is a practical size for a cabinet's physical size and expected size of pump motors andother fixed ignition sources being defined for the ZOI analyses.The fire source area is also a practical size for expected transients with an overall HRR of 317 kW,taken from the 98 th percentile HRR from NUREG/CR-6850, such as a trash can or smaller amount ofmaterial staged in a plant area. Larger transient fire source areas could be used if they weredeemed applicable, though it is likely that a larger area transient fire would not correlate to theoverall 317 kW HRR as defined in NUREG/CR-6850.The transient fire sources are located at the floor level of the particular physical analysis unit (PAU)under consideration as it is representative of the location of materials that would typically be broughtinto a PAU for temporary work, cleaning, or storage. If curbs, risers, or other elevated areas arepresent where transient combustibles could possibly be located, analyst judgment is used to locatethe fire in the location with the most limiting impacts. However, one characteristic of Waterford 3 is toW3F1-2015-0015Page 12 of 44that the presence of such areas is very limited; therefore, instances of such elevated cases arelimited as well.Note that oil source fires, as documented in PRA-W3-05-029 for oil source ZOIs, have a fire sourcearea determined based on the quantity of oil spilled using guidance from NUREG/CR-6850 for a spilldepth also based on the quantity of oil spilled.Transient combustibles during and post-transition will be subject to operational procedures whichlimit the total amount and locations where potential transients can be within the Waterford 3 plant byprocedure EN-DC-161.FM RAI S01 h.ivThe soot yield, heat of combustion, and carbon dioxide yield for Polyethylene (PE) are based ondata provided in Table 3-4.16 of the 4 th Edition of theSFPE Handbook of Fire ProtectionEngineering. The soot yield, heat of combustion, and carbon dioxide yield for wood are based ondata provided in Table 3-4.16 of the 4 th Edition of theSFPE Handbook of Fire Protection Engineeringfor natural materials. Specifically, the properties for the highest soot yield natural material (wood, redoak) are assumed in the CFAST model and are thus conservative for this material type. The heat ofcombustion used in the CFAST model was increased from the listed value of 12,400 kJ/kg (5,340Btu/lb) to 13,700 kJ/kg (5,900 Btu/lb) to avoid the lower heat of combustion limit of 13,100 kJ/kg(5,640 Btu/lb) in CFAST, Version 6.1.1. This results in a slightly non-conservative soot mass lossrate since the soot yield was not proportionately increased in the CFAST, Version 6.1.1 model tomaintain a constant soot yield to heat of combustion ratio. However, based on Table B-8 in AppendixB of PRA-W3-05-026, Rev. 0, the abandonment times for the transient fire scenarios are notsensitive to variations in the heat of combustion over a range that bounds the CFAST lower limitadjustment and a correction to the assumed soot yield for transient materials is thus not necessary.Finally, the carbon monoxide (CO) yield for wood listed in Table 3-1 of PRA-W3-05-026, Rev. 0 andused in the CFAST model is 0.0305 kg CO/kg CO 2 produced whereas the value based on Table 3-4.16 of the 4 th Edition of theSFPE Handbook of Fire Protection Engineering for red oak is 0.00315kg CO/kg CO 2 produced. The CO yield can have a minor effect on the temperature of the hot gaslayer via altering the radiant absorption characteristics of the combustion gases, but otherwise doesnot affect the predicted abandonment conditions given that toxicity of the modeled environment isnot considered. The effect may be conservative or non-conservative, depending on whether visibilityor temperature leads to abandonment. Note that the yield properties for wood and PE are reversedin Table 3-1 of the MCR abandonment report (PRA-W3-05-026, Rev. 0), but the averaged propertiesare used in the CFAST calculations, and these are not affected by the reversed data in Table 3-1 ofPRA-W3-05-026, Rev. 0.The sensitivity analyses provided in PRA-W3-05-026, Rev. 0 will be updated to include sensitivitycases on the assumed CO yield. The updated version of the calculation will be used to complete theresponse to PRA RAI S04 (120 day response).It should be also noted that the abandonment times for all baseline fire scenarios are determinedusing the NUREG/CR-6850 Optical Density (OD) threshold of 3 m

-1 (0.9 ft-1) (when the hot gas layerheight is at or below 1.83 m [6.0 ft]), the NUREG/CR-6850 hot gas layer temperature threshold of95°C (200°F), or an additional immersion temperature threshold of 50°C (122°F) when the hot gaslayer height is at or below 1.83 m (6.0 ft), whichever occurs first. The immersion temperaturethreshold is significantly more conservative than the NUREG/CR-6850 criteria and in most cases isthe limiting abandonment criterion. This introduces a significant conservative bias across all baselinefire scenarios and provides a conservative margin for bounding parameter uncertainty. toW3F1-2015-0015Page 13 of 44FM RAI S01 h.vThe cable mixture for the control room is assumed to be conservative based on input fromknowledgeable plant personnel and a review of the content of the more prevalent cable qualificationcategories which show only one of ten cables containing Neoprene (jacket only).The sensitivity of the predicted control room abandonment times to the assumed cable mix wasevaluated as part of the response to RAI FM 01e (see PRA-W3-05-027, Rev. 0). The most adversecable mix considered was one represented entirely by the properties of a Cross-Linked Polyethylene(XPLE)/Neoprene cable. It was shown that the predicted control room abandonment times can besensitive to the assumed cable mix, in particular if the XLPE/Neoprene cable properties areassumed; however, when considering the data for the same types of cables provided in NUREG-7010, Volume 1 and the CFAST tool's uncertainty for visibility predictions, it was concluded that thebaseline results as characterized by an equal mix of XLPE/Neoprene, Ethylene Propylene Rubber(EPR), and Hypalon cables are conservative. Because the XPLE/Neoprene cable has the largestsoot generation rate, as defined by the ratio of the soot yield to the heat of combustion, among allthermoset cables tabulated in Table 3-4.16 of the SFPEHandbook of Fire Protection Engineering (4 th Edition), there are no applicable cable fuels that are more adverse than the sensitivity casedescribed in the response to RAI FM 1e (see PRA-W3-05-027, Rev. 0). As such, it is concluded thatthe baseline abandonment results provided in PRA-W3-05-026, Rev. 0 are conservative with respectto the soot generation rate, or are not significantly sensitive to variations in the actual cable mix inthe control room when the fire model uncertainty with respect to the optical density (visibility)parameter is considered.FM RAI S01 i.ivThe failure criteria were based on predictions taken on the exterior of the upper surface of the panel(cabinet). This location was based on the failure mechanism of interest in the multi-compartmentanalysis (MCA), as documented in PRA-W3-05-005, of physical analysis unit (PAU) RAB7 where theCFAST analysis, as documented in PRA-W3-05-030, served to investigate the impacts from apotential hot gas layer, which would be expected to form in the ceiling area above the panels(cabinets) and descend to the top of the panels (cabinets), thus impacting the top surface of thepanel (cabinet) first. No credit for shielding is taken during the MCA analysis of PAU RAB7 for thepredicted temperature values by establishing the panel (cabinet) failure impacts on the outer surfaceof the panel (cabinet).FM RAI S01 j.iQualitative screenings in the multi-compartment analysis (MCA), as documented in PRA-W3-05-005,are based on the principles of Step 2.c in NUREG/CR-6850 for MCA qualitative screening.Qualitative screening of plant areas is performed before building the compartment adjacency matrixto eliminate those compartments that can be screened as an exposed compartment on the criteriondescribed in NUREG/CR-6850 that the exposed compartment does not contain any PRA-relatedcomponents and/or cables. In addition, a compartment is qualitatively screened as an exposingcompartment if the compartment does not share a common boundary with an exposed compartmentthat contains PRA related equipment and/or cables.Several plant areas were qualitatively screened in other fire PRA analyses and are noted in the PlantPartitioning, Qualitative Screening, and Ignition Frequency Development Notebook, as documentedin PRA-W3-05-001, due to their lack of PRA-modeled equipment that serve as potential fire scenariotarget failures. Although these areas can be screened as exposed compartments in the MCA, theirpotential to be exposing compartments was re- evaluated in the MCA. These areas were reviewedto confirm that their screening from consideration in the MCA is appropriate based on the principlesabove. toW3F1-2015-0015Page 14 of 44FM RAI S01 j.iiFor hot gas layer (HGL) temperature predictions, Chapter 2 of NUREG-1805 Fire Dynamics Tools(FDTs) spreadsheet 02.3_Temperature_CC.xls (v. 1805.1), which calculates the HGL temperature ina closed compartment given input parameters such as ambient air temperature, ignition source heatrelease rate (HRR), duration of the fire, and compartment geometry, is used. The use of a closedcompartment calculation is a conservative assumption for rooms that may have small ceilingopenings.As stated in PRA-W3-05-005, the following inputs are used for the MCA HGL calculations:For most compartments, the floor size is based on an equivalent square of the total surfacearea of the compartment. Some plant areas reflect unique geometries where a square shapeis not used.The compartment height is presented with the assumption that the ceiling/floor barrier fromone compartment to another is 1.5 ft. thick and is composed of concrete.The ambient room temperature for all plant areas is assumed to be 77° F unless otherwisenoted.The interior lining thickness (or barrier thickness) for walls is assumed to be 12 inches unlessotherwise noted. The interior lining material is taken as concrete unless otherwise noted.The duration of a fire is limited by either the amount of fuel in the exposing compartment orthe postulated time to successful suppression. A time of 25 minutes (1500 seconds) isselected as a conservative value that is representative of both the time to fuel burnout and theconfidence of suppression. Oil fire sources, as documented in PRA-W3-05-029, are based onthe quantity of oil spilled for burnout duration determination.The bounding heat release rates for various fixed and transient ignition sources are providedin NUREG/CR-6850 and use the 98th percentile HRRs for the various ignition source bins.Oil fire sources are based on the quantity of oil spilled for a HRR determination.The HGL analysis was conducted under the general assumption that the potential for target damagein an exposed compartment occurs when a severe HGL forms in the exposing compartment andpropagates to an exposed compartment. If the exposing compartment itself cannot form asufficiently high HGL temperature, it may be screened from consideration as an impact in theexposed compartment as the combined exposing plus exposed compartment volumes will be largerthan only the exposing compartment area and will serve to lower the overall HGL temperature. Asan exception to the HGL formation assumption, a unique scenario exists in which a HGL may notform in the exposing compartment, but may form in an adjacent (exposed) compartment due to thesmaller relative volume of the adjacent compartment. Essentially, a fire that initiates in the exposingcompartment could funnel hot gases into a smaller compartment where the HGL would be moresevere than in the original exposing compartment. This unique scenario was considered during theHGL analysis, but no such scenarios were determined to exist.FM RAI S01 j.iiiBarrier failure from the second to a third compartment was not considered based on guidance fromNUREG/CR-6850 that recommends limiting cases to one barrier failure. Additionally, inclusion of anadditional barrier failure probability would likely cause any MCA scenario of failure into a thirdcompartment to be screened based on the quantitative screening of retained plant areas in Section3.3 of PRA-W3-05-005. toW3F1-2015-0015Page 15 of 44FM RAI S02American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) Standard RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessments forNuclear Power Plant Applications," Part 4, requires damage thresholds be established to support theFire PRA. Thermal impact(s) must be considered in determining the potential for thermal damage ofstructures, systems, and components and appropriate temperature and critical heat flux criteria mustbe used in the analysis.In the updated response by letter dated June 11, 2014, to previous FM RAI 02.a, the licensee stated,in part, "[t]he design specifications for Waterford 3 cables required IEEE [Institute of Electrical andElectronics Engineers]-383 qualification. The materials of construction of the cables are consistentwith thermoset performance which was the basis for the determination for the Fire PRA."However, it appears that a damage threshold of 380 ºCentigrade (C) was used for thermoset cable,as opposed to 330 ºC, which is the NUREG/CR-6850-recommended bounding value for thermosetcable.a.120 day response b.Explain how the damage thresholds for non-cable components (i.e., pumps, valves,electrical cabinets, etc.) were determined. Identify any non-cable components that wereassigned damage thresholds different from those for thermoset and thermoplastic cables,and provide a technical justification for these damage thresholds.c. Explain how exposed temperature-sensitive equipment was treated, and provide thetechnical justification for the damage criteria that were used.Waterford 3 ResponseFM RAI S02 b.The majority of active non-cable components such as pumps, valves, and electrical panels(cabinets) have damage thresholds based on the cable type powering the component, with thefailure criteria being taken from NUREG/CR-6850 as appropriate for the particular cable type(thermoset or thermoplastic). Thermoplastic cable material is assumed if no other knowledge of thepower cable could be determined. For Waterford 3 no selection of thermoplastic cable performancewas found to be necessary based on plant information.Select electrical panels (cabinets) were assigned damage thresholds based on sensitive electronics'failure criteria taken from NUREG/CR-6850 as 65° C for temperature impacts or 3 kW/m 2 for heatflux impacts.One exception to the NUREG/CR-6850 failure criteria for electrical panels (cabinets) based onsensitive electronics is taken in the multi-compartment analysis (MCA) fire modeling for physicalanalysis unit (PAU) RAB7 using CFAST as documented in PRA-W3-05-030. In the RAB7 MCA firemodeling, a temperature failure criterion of 60° C rather than 65° C is used based on previouslydetermined failure criteria for the components housed within the electrical panels (cabinets) of PAURAB7. The heat flux failure criterion of 3 kW/m 2 is also used in PAU RAB7 for the MCA. The toW3F1-2015-0015Page 16 of 44scenario impacts of the MCA fire modeling of PAU RAB7 are used to supplement the fire sourceFDT analyses of PAU RAB7 for quantification in the fire PRA results.Passive, non-combustible components such as tanks, pipes, and check valves were deemed to notbe impacted by potential fire scenarios and as such are not assigned damage thresholds based onguidance in NUREG/CR-6850.No credit for exposure duration or shielding is taken during the various analyses described above.The methodology for these damage threshold assignments is presented in PRA-W3-05-006F for usein potential fixed ignition source analyses, in PRA-W3-05-013 for use in potential transient ignitionsource analyses, and in PRA-W3-05-029 for oil source fire scenarios.FM RAI S02 c.Temperature-sensitive equipment (sensitive electronics) at the Waterford 3 plant are typicallyinstalled and housed in electrical panels (cabinets) rather than being mounted in an open or exposedlocation. No instances of directly exposed temperature-sensitive equipment (sensitive electronics)were found in the fire PRA scenario analyses. Most panels (cabinets) installed at the Waterford 3plant are ventilated and as such no credit is taken for shielding of the enclosed sensitive electronicsduring fire scenario analyses.Sensitive electronics as targets in potential fire source analyses are treated with a zone of influence(ZOI) investigation, which is discussed in greater detail in RAI PRA S08.FM RAI S03Regarding the V&V of fire models, NFPA 805, Section 2.7.3.2, states that each calculational modelor numerical method used shall be verified and validated through comparison to test results orcomparison to other acceptable models.The LAR Section 4.5.1.2 states that fire modeling was performed as part of the FirePRA development (NFPA 805, Section 4.2.4.2). Reference is made to LAR Attachment J, for adiscussion of the V&V of the fire models that were used. Furthermore LAR Section 4.7.3,"Compliance with Quality Requirements in Section 2.7.3 of NFPA 805," states, in part, that"Calculational models and numerical methods used in support of compliance with 10 CFR 50.48(c)were verified and validated as required by Section 2.7.3.2 of NFPA 805."For any tool or method identified in the response to FM RAI S01.a above, provide the V&V basis ifnot already explicitly provided in the LAR (for example in LAR Attachment J). Provide technicaldetails to demonstrate that these models were applied within the validated range of inputparameters, or justify the application of the model outside the validated range in the V&V basisdocuments.Waterford 3 ResponseAs discussed in RAI FM S01.a, the use of NUREG-1805 Fire Dynamics Tools (FDTs) spreadsheet05.1_Heat_Flux_Calculations_Wind_Free.xls (v. 1805.0) Solid Flame 1 tool which calculates theheat flux at a specific distance from a fire source is not discussed in Attachment J of the Waterford 3licensing amendment request (LAR) Supplement (Waterford Letter W3F1-2013-0048). This tool andits methods are used in the fire PRA to generate a comparison plot of various heat flux calculationalmethods in the fixed ignition source and transient ignition source methodology reports PRA-W3-05-006F and PRA-W3-05-013, respectively. toW3F1-2015-0015Page 17 of 44The V&V basis of the Solid Flame 1 tool is the same as the existing information in Attachment J ofthe LAR Supplement for the NUREG-1805 FDTs spreadsheet05.1_Heat_Flux_Calculations_Wind_Free.xls (v. 1805.0) Point Source Method tool which is alsoused to generate the comparison plot of various heat flux calculational methods in the fixed ignitionsource and transient ignition source methodology reports PRA-W3-05-006F and PRA-W3-05-013respectively. The V&V basis from the Point Source Method tool is repeated for the Solid Flame 1tool.This is appropriate as the two methods are estimating the same parameter (heat flux) and NUREG-1805 lists the point source method as being the simplest (i.e. more conservative estimate) and thesolid flame method being somewhat more refined with both tools being taken from the sameoriginating source, the SFPE Handbook of Fire Protection Engineering, 3 rd Edition.A revised Attachment J of the LAR Supplement will be provided as part of the 120 day RAI responses.FM RAI S04 a.120 day response b.90 day response.FM RAI S0590 day response. toW3F1-2015-0015Page 18 of 44PRA RAI S0190 day response.PRA RAI S02By letter dated June 11, 2014, the response to PRA RAI 04, in which the NRC staff asks about howdefense-in-depth was evaluated, did not describe the criteria used to determine when there was a"substantial imbalance between echelons" and did not describe the types of improvements made inresponse to the evaluation. Also, though the licensee presents criteria used to address safetymargin, description of how that criteria was applied to specific parts of Fire Risk Evaluation (FRE)was not provided. Please explain the method used in the FREs to determine when a substantialimbalance between echelons was determined to exist and describe the improvements that weremade as a result of the defense-in-depth evaluation. Also provide a description of how criteria usedin the FREs to evaluate safety margin criteria was applied to specific elements of the FREsconsistent with guidelines/criteria in NEI 04-02.Waterford 3 ResponseThis response will address the key elements of PRA RAI S02:

A.Provide defense-in-depth (DID) criteria use to determine when there was a "substantialimbalance between echelons".

B.Describe the method(s) used in the FREs to implement the stated criteria.

C.Describe the types of improvements prescribed as a result of the DID evaluation.

D.Describe how safety margin criteria were applied to specific elements in the FRE process.(A, B)The process used for evaluating/considering DID for the Waterford 3 NFPA 805 project is describedbelow. The process was applied to each fire area/fire scenario. DID considerations were part of theFRE process (determining credit and if elements needed to be strengthened to offset weakness) aswell as in qualitative DID expert panel evaluations. Balance between the DID echelons is examinedin the process. If the criteria for balance is not met (e.g. substantial imbalance between echelons),the process provides guidance to mitigate as much as possible/feasible.The following methods and criteria for evaluating DID and examples of these improvements aresourced from WF3-FP-13-00004, Rev. 0, "Waterford 3 Defense in Depth Report for NFPA 805" andNEI 04-02 and summarized below:A review of the impact of the change on DID was performed using the guidance below from NEI 04-02. NFPA 805 defines DID as:Preventing fires from startingRapidly detecting fires and controlling and extinguishing promptly those fires that do occur,thereby limiting damageProviding adequate level of fire protection for structures, systems and components importantto safety; so that a fire that is not promptly extinguished will not prevent essential plant safetyfunctions from being performed.In general, the DID requirement was considered satisfied if the proposed change does not result in asubstantial imbalance among these elements (or echelons). The review of DID was qualitative andaddressed each of the elements with respect to the proposed change. Fire protection features and toW3F1-2015-0015Page 19 of 44systems relied upon to ensure DID were identified in the assessment (e.g., detection, suppression system).Consistency with the DID philosophy is maintained if the following acceptance guidelines, or theirequivalent, are met:A reasonable balance is preserved among 10 CFR 50.48(c) DID elements.Over-reliance and increased length of time or risk on performing programmatic activities tocompensate for weaknesses in plant design is avoided.Pre-fire nuclear safety system redundancy, independence, and diversity are preservedcommensurate with the expected frequency and consequences of challenges to the systemand uncertainties (e.g., no risk outliers). (This should not be construed to mean that morethan one NSCA train must be maintained free of fire damage.)Independence of DID elements is not degraded.Defenses against human errors are preserved.The intent of the General Design Criteria in Appendix A to 10 CFR Part 50 is maintained.The process for evaluation of DID begins with the risk assessment. The associated fire area risk(CDF) and consequences (CCDP) were reviewed by the site DID expert panel to address generalDID echelon imbalances. The following techniques were implemented to mitigate potentialimbalance:Areas with high ignition frequencies were discussed as areas where additional fire brigade orfire prevention methods may be neededAreas with high CCDPs were discussed for additional fire prevention elements.Beyond the risk assessment, a discussion of firefighting strategies and operations impactswas performed to determine if additional DID methods could enhance fire brigade responseThe following list of items was discussed for each fire area as applicable:Review DID echelons for support of specific assumptions in risk evaluations (transientcombustible limits, hot work limits, fire barrier considerations).Consider CCDP sensitive areas for improvement in fire prevention.Develop Operation's procedure enhancement to improve response to fire scenarios.Consider the need for any high-risk, combustible sensitive areas where habitability is limited.Consider any active or passive fire protection features retained going forward.For high-risk affected areas consider changes to operations procedures consideringadvanced response or alarm conditions.

oA review of possible recovery actions not modeled in the PRA was performed usingexisting strategies to determine if they could provide needed enhancements in DID.

oThis review included a review of the following procedures:OP-901-502---Rev 027---Evacuation of Control Room and Subsequent PlantShutdown.OP-901-503---Rev 308---Isolation Panel Fire.OP-901-524---Rev 012---Fire in Areas Affecting Safe Shutdown. toW3F1-2015-0015Page 20 of 44Consider possible modifications that could enhance DID.

oA review of possible modifications was performed to determine if additionalmodifications not modeled in PRA could provide needed enhancements in DID.Review of possible DID enhancements based on the 92-18 valves found to be modeled in thePRA inconsistent with their NSCA function.(C)Examples of DID enhancements include, but not limited to the following: Hot work and combustiblecontrol enhancements, requiring suppression and/or detection to meet DID criteria, installation ofradiant fire barriers, etc. These examples are derived from panel discussions as documented in theFRE and DID source reference(s). The primary enhancements chosen as a result of this review werehot work and combustible control enhancements and credit for partial height walls.Table 1 (next page) was used to aid in the consistency of the review of DID. toW3F1-2015-0015Page 21 of 44Table 1-Considerations for Defense

-in-Depth DeterminationMethod of Providing DIDConsiderationsEchelon 1:Prevent fires from starting Combustible Control Hot Work ControlCombustible and hot work controls are fundamental elements ofDID and as such are always in place. The issue to be consideredduring the FREs is whether this element needs to bestrengthened to offset a weakness in another echelon therebyproviding a reasonable balance. Considerations include: Creating a new Transient Free Areas Modifying an existing Transient Free AreaThe fire scenarios involved in the FRE quantitative calculationshould be reviewed to determine if additional controls should beadded.Review the remaining elements of DID to ensure an over-relianceis not placed on programmatic activities to compensate forweaknesses on plant design.Echelon 2:Rapidly detect, control and extinguish promptly those fires that do occur thereby limitingfire damage Detection system Automatic fire suppression Portable fire extinguishers providedfor the area Hose stations and hydrantsprovided for the area Fire Pre-Fire PlanAutomatic suppression and detection may or may not exist in theFire Area of concern. The issue to be considered during the FREis whether installed suppression and or detection is required forDID or whether suppression/detection needs to be strengthenedto offset a weakness in another echelon thereby providing areasonable balance. Considerations include: If a Fire Area contains both suppression and detection andfirefighting activities would be challenging, both detection andsuppression may be required If a Fire Area contains both suppression and detection andfirefighting activities would not be challenging, requiredetection and manual firefighting (consider enhancing thepre-plans) If a Fire Area contains detection and a recovery action isrequired, the detection system may be required. If a Fire Area contains neither suppression nor detection anda recovery action is required, consider adding detection orsuppression.The fire scenarios involved in the FRE quantitative calculationshould be reviewed to determine the types of fires and relianceon suppression should be evaluated in the area to best determineoptions for this element of DID. toW3F1-2015-0015Page 22 of 44Echelon 3:Provide adequate level of fire protection for systems and structures so that a fire will notprevent essential safety functions from being performed Walls, floors ceilings and structuralelements are rated or have beenevaluated as adequate for thehazard. Penetrations in the Fire Area barrierare rated or have been evaluated asadequate for the hazard. Supplemental barriers (e.g.,ERFBS, cable tray covers,combustible liquid dikes/drains, etc.) Fire rated cable Reactor coolant pump oil collectionsystem (as applicable) Guidance provided to operationspersonnel detailing the requiredsuccess path(s) including recoveryactions to achieve nuclear safetyperformance criteria.If fires occur and they are not rapidly detected and promptlyextinguished, the third echelon of DID would be relied upon. Theissue to be considered during the FRE is whether existingseparation is adequate or whether additional measures (e.g.,supplemental barriers, fire rated cable, or recovery actions) arerequired offset a weakness in another echelon thereby providinga reasonable balance. Considerations include: If the VFDR is never affected in the same fire scenario,internal Fire Area separation may be adequate and noadditional reliance on recovery actions is necessary. If the VFDR is affected in the same fire scenario, internal FireArea separation may not be adequate and reliance on arecovery action may be necessary. If the consequence associated with the VFDRs is highregardless of whether it is in the same scenario, a recoveryaction and / or reliance on supplemental barriers should beconsidered. There are known modeling differences between a Fire PRAand nuclear safety capability assessment due to differentsuccess criteria, end states, etc. Although a VFDR may beassociated with a function that is not considered a significantcontribution to CDF, the VFDR may be considered importantenough to the NSCA to retain as a recovery action.The fire scenarios involved in the FRE quantitative calculationshould be reviewed to determine the fires evaluated and theconsequence in the area to best determine options for thiselement of DID.(D)The fire risk evaluation incorporated the re-examination of plant system performance given thespecific demands associated with postulated fire events (by means of the analytical fire PRA modeland qualitative DID panel). The methods, input parameters, and acceptance criteria used in theseanalyses have been reviewed against that used for the plant designbasis events.This is the processby which the safety margin inherent in the analyses for the plant design basis events have beenpreserved in the analysis for the fire event.An example of a practical application of maintaining safety margin is the evaluation of compensatorymeasures that provide an equivalent level of protection of a suppression system being impaired. Thisexample would be identified and evaluated in the FRE process and consideration of equivalency (ornot) provides the basis of adequately evaluating and maintaining safety margin. Each FRE providesthe specific details of how safety margins are maintained for each fire area. For example, theWaterford 3 FRE for RAB 1 (EC-F10-002) lists four specific areas where safety margins aremaintained; fire modeling, plant system performance, PRA logic model, and success pathconfirmation.PRA RAI S03120 day response. toW3F1-2015-0015Page 23 of 44PRA RAI S04120 day response.PRA RAI S05By letter dated June 11, 2014, the response to PRA RAI 08 explains that Attachment W of the LARwas entirely revised as part of the LAR supplement, and that calculation of CDF and LERF isdiscussed in the updated Section W.2.1 of the LAR. The discussion in the updated Section W.2.1 ofthe LAR primarily references an internal Waterford document instead of providing the requesteddescription of the calculations for CDF and LERF used at Waterford.The response to PRA RAI 57a is also related to the calculation of CDF and LERF because itstates, in part, "LAR Supplement Attachment C (Table B-3) provides detailed dispositions for non-modeled VFDRs." The NRC staff reviewed the updated Attachment C of the LAR and notes that in187 instances VFDR dispositions state: "[t]his condition has no corresponding PRA basic event andby definition has insignificant risk" and references an internal Waterford document. Review of thereferenced internal Waterford document indicates that some component failure modes were selectedfor exclusion from the PRA (e.g., control indication, component cooling water makeup to emergencydiesel generators, HVAC and Feed Tank level), but the rationale and justification for exclusion is notalways clear. Provide the following information:

a.Summary of the model adjustments made to remove VFDRs from the compliant plantmodel, such as adding events or logic, or use of surrogate events.

b.Identify any risk-reduction modifications credited in the change-in-risk calculations andexplain how these are included in the post-transition and compliant plant models.

c.Discussion of the rationale used for excluding VFDRs from the calculation of CDF andLERF. Include justification of why so many VFDRs were excluded from the change-in-risk calculations.

d.Summary of how the change in risk was determined for MCR abandonment scenarios,including a summary of how the CCDP was determined for the compliant and for thevariant plant models.

e.Provide separately the total risk increase associated with retained VFDRs and the totalrisk decrease associated with risk reduction modifications.Waterford 3 ResponseA) The baseline fire PRA represents the non-compliant plant since it addresses the impact of theVFDRs on plant components. To derive the compliant plant model, the non-compliant plant isadjusted to represent a condition where the VFDR no longer exists in the specific PAU. In somecases, this required additional logic to address spurious impacts that were not considered asplausible in the original fire PRA model.To simulate the compliant condition, the impacts to the source of the VFDR were removed from thePAU cable and component impact listing such that components associated with the VFDR were notfailed by any fire in the PAU. This results in no contribution to the CDF or LERF results from thesource of the VFDR and a compliant plant is represented. For example, if a specific VFDR toW3F1-2015-0015Page 24 of 44addressed a power cable to a valve that was present in a specific PAU, the non-compliant casewould assume a fire could potentially fail the cable and the valve would be failed due to a loss ofpower. For the compliant plant the cable is excluded (removed) from the PAU inventory such that afire cannot fail that specific cable and the valve will not be failed by a fire in that PAU.B) The risk reduction modifications (plant modifications, recovery actions and program changes) aredescribed in PSA-WF3-03-02, "Waterford Steam Electric Station 3 Summary of Fire PRA DrivenPlant Improvements to Waterford 3 to Support Risk Optimization" are summarized below:

o Removal of secondary combustible material from RAB 27. This involved the removalof combustible materials from RAB 27 to preclude fire growth due to secondaryignition.o Inclusion of local manual trip of reactor coolant pumps given a loss of remote tripswitch function. This involves locally tripping power for the reactor coolant pumpsfrom the turbine building switchgear for scenarios where the main control boardswitches may not be available due to fire induced effects.

o Inclusion of ERFBS wrap in RAB 6. The walkdown confirmed the existence of theERFBS wrap in RAB 6. Therefore, the fire PRA included it as an "as found" condition.

o An implementation item (S2-21) was added to update plant procedures to satisfy theFPRA credited mission times for Nitrogen Accumulators.These four risk reduction modifications were assumed to be installed when the VFDRs werequantified for both post-transition and compliant plant models. By their inclusion the risk reductionobtained by addressing VFDRs is captured and not masked by the risk reductions of thesemodifications.C). PSA-WF3-03-01,"Waterford Steam Electric Station 3 Methodology for Addressing VFDRs in theFire PRA and NFPA-805, November 2013" provides a detailed evaluation of each VFDR. In somecases the component associated with the VFDR is not contained in the fire PRA model. This istypically due to differences in the deterministic requirements and the PRA success criteria boundaryconditions. All VFDRs that would impact the risk model (component included in the model) wereaddressed in the FRE calculations. Additionally, if the failure would impact the PRA model andshould be included, it was added to the model to reflect the postulated failure modes. The remainingVFDRs have no impact on the PRA model and were screened using one of several documentedcriteria. A VFDR was excluded from the fire risk evaluation calculation if it met any of the following:Disposition 1: Excluded based on Refined Room Cooling AssessmentThe internal events PRA model serves as the starting point for the fire PRA and the system successcriteria are derived from information contained in PSA-WF3-01-SC. The internal events PRA wasupdated in March 2013. One area updated was the room cooling requirements.The internal events system notebook (PSA-WF3-01-SY) indicates that the steady state peaktemperature following a loss of HVAC is below the realistic failure temperatures for equipment in thesafeguards, CCW pump, CCW heat exchanger, charging pump, battery, +46 foot heating andventilation equipment, control room HVAC equipment room, EFW pump, and electrical switchgearrooms. The loss of HVAC to any of these rooms would not result in component failure in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.The fault tree logic for these HVAC models was removed in Revision 5 of the Waterford 3 internalevents model (PSA-WF3-01-QU) and there is no longer any basic event to be associated with thecable fault.Disposition 2: Excluded based on PRA Success CriterionPSA-WF3-01-SC provides the basis for the internal events PRA and therefore the fire PRA. Itdefines the minimum equipment and systems necessary to maintain a safe stable state and no core toW3F1-2015-0015Page 25 of 44damage. This must be maintained for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. In some cases this includes specificequipment that is useful in regard to safe and stable, but is not strictly necessary for successfulshutdown operation. This includes makeup to systems which have been shown to have adequateinventory to function for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. More specific descriptions for the use of Disposition 2 in excludingVFDR components are provided below.Disposition 2a: BAM ILT 206 and BAM ILT 208 Boron Acid Makeup Tank Level SensorLevel sensors BAM ILT 206 and BAM ILT 208 are associated with the Boric Acid MakeupTank which is only required for an Anticipated Transient Without Scram (ATWS) event in theWaterford 3 internal events PRA. NUREG/CR-6850 allows exclusion for sequencesassociated with events that a low-frequency argument can be justified. In the providedexample, the text states on page 2-7 that:"it can often be easily demonstrated that anticipated transient without scram (ATWS)sequences do not need to be treated in the Fire PRA because fire-induced failures will almostcertainly remove power from the control rods (resulting in a trip), rather than cause a "failure-to-scram" condition. Additionally, fire frequencies multiplied by the independent failure-to-scram probability can usually be argued to be small contributors to fire risk due to theassociated low frequency of core damage and LERF scenarios."On this basis the level transmitters are excluded from the fire PRA.Disposition 2b: Excluded Component Cooling Water Valves for Containment Fan CoolersVFDRs identify valves CC-807B and CC-823B which are normally open valves associatedwith the B containment fan cooler. The failure of the valves will result in a loss ofcontainment fan cooler water supply. The internal events PRA combined these componentswith the fan cooler itself based on guidance from NUREG/CR-6928.A sensitivity study (PSA-WF3-03-01, Att. 1," Mapping of CCS Isolation Valves toContainment Fan Cooler Failures") was performed to determine the impact of not mappingthe valves. The valve impact was mapped directly to the containment fan cooler. Thesensitivity results indicated that the inclusion of these impacts had no measurable effect oncore damage or large early release frequency. The removal of the valves from the mappinghas no impact on the fire PRA.Disposition 2c: Excluded Level Indication on Basis of Low Frequency of OccurrenceLevel sensors CC-ILT-7010A and CC-ILT-7010B provide remote indication to CP-8 andmaintain normal makeup to the CC surge tank. The PRA credits the secondary flow path thatresponds on low level in the tank and is typically associated with leakage events inaccordance with the PRA modeled failures that deal with divergent flow events.Although the supply would be present for most postulated fires, components for this path arenot powered by onsite emergency power and would be unavailable following a loss of offsitepower.The level switches CC-ILS-7012 (LO) and CC-ILS-70l0 (HI) control CMU-226 operation inAUTO. The automatic controller is mapped to address the need to provide makeup if a flowdivergence occurs. The contribution to CCW failure due to divergence has a low frequencycontribution.Since the modeled function is addressed by the automatic makeup and the contribution is lowfrequency the control panel indication is not required in the PRA. toW3F1-2015-0015Page 26 of 44Disposition 2d: Main Steam Valves MS-119A and MS-119BMain steam valves MS-119A and MS-119B are drain lines that branch off of the steam lines.The size of the line is 2 inch diameter and the main line is a 40 inch diameter pipe. Based ongeneral PRA guidance branch lines less than one-third the diameter of the main line can beexcluded. Therefore, the valves are not included in the fire PRA.Disposition 2e: HVR-502B Exhaust Fan Blade Pitch ControllersThe internal events system notebook for HVAC indicates that HVR-502B (EDG B exhaust fanblade pitch controller) is not a damper, but a hydro-motor attached to the EDG ventilation fan(E28 3B) and is used to adjust the fan blade pitch to control air flow and room temperature.The internal events PRA combined this component with the fan itself based on guidance fromNUREG/CR-6928.In evaluating the effect of the hydro-motor failing and the pitch going to minimum (minimumflow) the impact would be to make EDG room temperature higher than it would be if thehydro-motors were functioning and controlling temperature.However, plant experience over several years has demonstrated that there is sufficientmargin that even if the fan blades are kept at a minimum pitch that the EDG roomtemperature was maintained within the design temperature of 120F. This is supported by theinherent margin in the fan design which is greater than 30% capacity.Therefore, operability of the pitch control is not required for successful fan performance andcan be excluded from the PRA model because it does not affect success of the EDGventilation function. The failure mode is not mapped.Disposition 2f: CHW-603 and CHW-919 Main Control Room Flow Control ValveVFDRs identify valves CHW-603 and CHW-919 which are flow control valves associated withthe main control room air handling units. The failure of the valves will result in a loss ofcooler water supply for the air handling units. The internal events PRA combined thesecomponents with the fan cooler itself based on guidance from NUREG/CR-6928.A sensitivity study was performed to determine the impact of not mapping the valves (PSA-WF3-03-01 Rev 1, Att. 2). The valve impact was mapped directly to the air handling unit.The sensitivity results indicated that the inclusion of these impacts had no measurable effecton core damage or large early release frequency. The removal of the valves from themapping has no impact on the fire PRA.Disposition 2g: HVC-ITE-5026A(B) and HVC-ITE-5028A(B) Main Control Room TemperatureInstrumentThese temperature elements provide temperature indication in the control room. For the firePRA, habitability is not based on a strict temperature criterion but rather a realistic modelingof the effects of a fire in the main control room. The analysis predicts the time until the areawould be considered uninhabitable and evacuation would be required. Since the evacuationis based on physical criteria such as smoke density, room heating, and degree of controlpanel failure, the instruments are not mapped in the fire PRA.Disposition 2h: HVC-MAHU-0013A and HVC-MAHU-0013B Air Handling UnitsThese units are not associated with the main control room. These AHUs are associated withthe main control room HVAC room (AHU-26s). AHU-26s are not realistically required due tothe load heat load in normal operation (post-LOCA is not applicable to the FPRA) and isestimated to be roughly 5% of capacity. They are excluded based on refined HVAC analysisdiscussed under Disposition 1. toW3F1-2015-0015Page 27 of 44Disposition 2i: Reactor Pressure and Temperature InstrumentsThe listed instruments are associated with the ability to monitor subcooling which is mosttypically associated with cooldown. The end state is hot standby which does not requirethese instruments for maintaining hot standby; i.e., inability to measure subcooling will notaffect mitigation of core damage scenarios at hot standby. Since this is the case, it is notmapped in the fire PRA.Disposition 2j Steam Generator Pressure Transmitters SG-IPT-1115A and SG-IPT-1125BThe pressure transmitters are associated with the diverse emergency feedwater actuationsystem (DEFAS) and main feedwater isolation. Main feedwater isolation is not a concernsince feedwater is not required or credited. DEFAS would provide an additional EFW startsignal and would not provide an adverse response since EFW is the desired fire PRAresponse for decay heat removal. Since no credit is given for these signals they are notrequired in the PRA and are not mapped. They do not have any adverse impacts on the firePRA response model.Disposition 2k: CVC ILT 227 - Level TransmitterThe level transmitter listed is associated with the volume control tank (VCT). A failure of thetransmitter may isolate valves from VCT (CVC-183) and the chemical addition tank andRWSP (CVC-507) to the charging pumps. It would have no effect on the safety injectionpumps.The internal events PRA model success criteria indicates that a loss of charging for a 24 hourperiod (PRA mission time) combined with normal leakage is insufficient to result in a loss ofcore cooling based on the accident sequence supporting success criteria documentation .Further, RCS inventory shrinkage is also minimized when the plant state is maintained at hotstandby. Therefore, charging is not required for the PRA model and this failure has noimpact on the PRA success criteria. Should additional inventory be required due to additionalleakage, the PRA credits the safety injection systems to restore level.Since charging is not required the status of the level transmitter has no impact and it is notmodeled based on low probability.Disposition 2l: Condensate Storage Pool Level Transmitter EFW-ILT-9013BThe instrument provides indication of condensate storage pool (CSP) in the main controlroom. The time required to deplete the CSP is over nine (9) hours as defined in the PRAsuccess criterion. The long delay time between EFW actuation and low CSP level wouldprovide the operators ample time to deal with any fire-induced failure of CSP levelinstrumentation and utilize alternative measures. The fire PRA increases the human failureprobability by a factor of 10x to address the potential for having to address adverse effectsand the level indication is not included in the model.Disposition 2m: Diesel Generator B Day Tank Level Transmitter EGF-ILT-6903BThe instrument transmitter provides day tank level indication in the control room. Indication isnot relied upon to refill the tank. Automatic refill is provided by a separate set of instruments.The level instrument that controls the EGF fuel oil feed pump is EGF-ILS-6907B and thiscomponent is modeled in the PRA. No manual actions are credited based on control roominstrumentation. The component is not mapped. toW3F1-2015-0015Page 28 of 44Disposition 2n: Makeup System Valves CMU-ISV-0524B, CMU-ISV-0532A/B, CMU-ISV-0532BA sensitivity study (PRA-W3-05-040 Att. 2, "Documentation of CCW Surge Tank InventoryDepletion Timing for Line Opening"), was performed to assess the impact of flow divergenceon CSP depletion time. The assessment indicated that prolonged operation, several hours,would not significantly impact the time available for the operator to accomplish alignment ofadditional water sources to maintain EFW flow. Since the impact of the postulateddivergence of CSP inventory does not have a measureable impact it is screened based onlow probability.Disposition 3: Selected Exclusion based on Postulated Failure ModesThe exclusion involves one or more steps depending on the type of exclusion being considered. Thefailure mode specified in the VFDR is used to define the component failure modes to be excluded in the analysis.The interpretation of the failure mode presented in the VFDR is based on the component failuremodes addressed in the fire PRA. Another consideration is the type of cable involved. Power andcontrol cables have different potential impacts and, depending on the component, could have noimpact at all.Since the fire-induced failure mode will not result in a credible failure mode for the fire PRA model, itis removed from both the non-compliant and the compliant cases so the net impact on the delta CDFand LERF calculation is zero.Note that for some components, conflicting failure modes may exist (fails to open and fails to close,for example). The Waterford 3 FPRA includes mutually exclusive logic to exclude both failure modesappearing in a single cut set such that a single impact is defined for each cut set assessment in mostall cases.A few scenarios exist where it is possible to have conflicting failure modes leading to different failuresequences. In that case a validation of the inclusions of all failure modes was performed. Note thatthe actual failure mode (blocks flow, diverts flow, precludes flow) may make multiple failure modesvalid which have the same impact. An example is a failure to close having the same impacts as atransfers open.Disposition 4: Excluded based on Conservative Instrumentation ModelingThe original fire PRA mapped the instruments for steam generator level (and others) to the operatoraction that relies on the instrument and not to the instrument itself. As a result the original fire PRAmodel fails the operator's ability to control SG level if only one of the sensors failed. This is clearlynot the case based on discussions with Waterford 3 project staff with operational experience or othersimilar designed Combustion Engineering plants.There are five operator actions in the internal events PRA that are related to the level instruments(SG-ILT-1115B and SG-ILT-1125B) of which only one is pertinent to the fire PRA. For that action,the procedural step related to these instruments is met regardless of the loss of a single indicator.Instruments SG-ILT-1113A/B/C/D and SG-ILT-1123A/B/C/D are currently mapped to componentsand the impact is reflected in the assessment of the operator action probability of success giveninstrumentation fault.Disposition 5: Excluded based on Status at Power OperationSI 401A and B are used only during shutdown cooling operations and are not required in the PRA.Spurious operation of both valves is postulated to initiate an interfacing systems LOCA. However,during power ascension and while at power operation the motive power to these valves is removed.Since power is removed at the breaker there is no motive power to cause the valve to change toW3F1-2015-0015Page 29 of 44position so a fire-induced cable fault would not cause the valves to change position. Modeling ofcable faults for these valves is not required. This forms the basis of Disposition 5.The summary of the VFDR categorization as a result of the disposition approach is provided in Table1 below.Table 1. VFDR DispositionInitial VFDRs to Disposition231100%Disposition 1 classes 4519.5%Disposition 2 classes 3716.0%Disposition 3 1 83.5%Disposition 4 2<1%Disposition 5 2<1%Retained for FRE Assessment13759.3%1. Retained for FRE Assessment, reduced modes.As the table indicates, 62.8% of the VFDRs (Disposition 3 and those not dispositioned) wereretained and assessed using the FRE process. Table 1 provides insight into the rationale forexcluding VFDRs from the PRA model. Examination shows that almost all exclusions were due toeither disposition 1 or disposition 2. For disposition 1 detailed analysis supporting the PRA for roomheatup substantially reduced the need for HVAC. The HVAC-associated VFDRs were retained butare not necessary to meet PRA success criteria. Disposition 2 involved similar considerations thatinvolved PRA success criterion. Only 37% of the VFDRs were excluded with over half of thosebeing associated with improved internal events success criteria for HVAC and not on the actualVFDR consideration. The remaining were based on PRA success criteria requirements. Thisconclusion is drawn that over two thirds were retained and those not addressed explicitly in the PRAwere due to the aggregate effects of applying improvements to the PRA modeling.D). The MCR abandonment scenarios were assessed based on the human performance to assessthe implementation of the remote shutdown panel (LCP-43). A conservative criterion was appliedthat a single operator failure to implement a procedural step was assumed to lead to core damage.This limiting factor dominated the risk contribution but did not result in unacceptable riskconsequences. Due to the relatively high contribution derived from the operator action assessmentthe supporting equipment was not explicitly modeled. Therefore, the model was insensitive to anyVFDRs when examining MCR abandonment. Note that the MCR abandonment is being re-evaluated and results will be included in response to RAI PRA S18.Other MCR scenarios that did not result in abandonment were evaluated and estimates for the riskimpact of the variances was determined. This included the main control board and back panel firesthat did not cause a significant loss of control and did not require evacuation. In these scenarios theVFDR impacts were excluded in the manner described above and the change in CDF and LERFnoted. In some cases the CCDP was calculated to be 1.0 prior to the VFDR being excluded and theexclusion may not have impacted the failures such that the CCDP remained at 1.0. In those cases itwas not possible to determine a change in risk associated with the VFDR.E). The increase associated with retained VFDRs and the net risk reduction due to plant riskreduction actions is provided in Table 2. Note that the removal of combustible materials in RAB 27 toW3F1-2015-0015Page 30 of 44and the extension of accumulator mission time were not driven by VFDRs but rather an overall riskreduction strategy and this was based on CDF reduction. In these cases a similar process wasperformed as for other VFDRs but the risk impact was developed on a total change in CDF. If thetotal change exceeded the guidance used for the VFDRs a refinement step was performed todetermine if any PAU contribution exceeded the change in risk criterion for the VFDRs. If this wasfound, then the same basis was used to support the plant change.Table 2 presents the risk impact of the retained, (not brought into compliance) VFDRs to the totalCDF and LERF. The first row represents the total risk attributed to the retained VFDRs. TheseVFDRs were retained because the FRE risk evaluation found their impact was sufficiently small to beacceptable. The next row indicates the total CDF and LERF reduction by implementation of actionsto bring VFDRs into compliance. The third and fourth rows are the risk reduction associated withCDF reduction. This includes the removal of combustibles in RAB 27 and the accumulator missiontime extension. The risk reduction due to implementing the ERFBS modification was found to benegligible and therefore not necessary to include in this response (RSC-CALKNX-2013-1004).The total risk decrease from risk reduction modifications is given by the second to last row of Table 2below. The net change represents the sum of the increases and reductions.Table 2. Retained VFDR CDF and LERF Risk ContributionAttributeChange in CDF (/yr)Change in LERF (/yr)Total Risk Increase FromRetained VFDRs2.31E-6 Increase1.29E-7 IncreaseRisk Reduction Actions3.42E-6 Reduction8.70E-7 ReductionRemoval of CombustibleMaterials in RAB 272.09E-4 ReductionNot calculated (Note 1)Extension of AccumulatorMission Time7.30E-06 ReductionNot calculated (Note 1)Total Risk Decrease from RiskReduction Modifications2.20E-4 Reduction8.70E-7 ReductionNote 1: CDF reduction was sufficient to justify implementation of change; no calculation of LERF wasrequired.Based on the results of Table 2 the net decrease far exceeds the retained risk from identified VFDRsand other identified improvements. The total retained risk is considerably lower than the total risklimits identified in Regulatory Guide 1.174 which indicates if the total risk including the sum of theincreases due to VFDRs does not exceed the criteria (delta CDF <1E-5 and delta LERF <1E-6), thenthe risk is deemed acceptable.PRA RAI S06By letter dated June 11, 2014, the response to PRA RAI 09 explains that the peer reviews of theInternal Events and Fire PRAs were performed in accordance with the guidelines in NEI 05-04 andNEI 07-12, which reference the clarifications and qualifications of RG 1.200, "An Approach forDetermining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-InformedActivities." The response does not explicitly state which revision of RG 1.200 (i.e., Revision 1 or 2) toW3F1-2015-0015Page 31 of 44was used in the peer reviews. The licensee's Internal Events PRA peer review appears to indicatethat it was performed in 2009 against PRA Standard ASME RA-Sb-2005 as clarified by RG 1.200Revision 1, dated January 2007. The current guidance for using PRA for risk-informed applicationsis RG 1.200 Revision 2, dated March 2009 (ADAMS Accession No. ML090410014), which is basedon PRA Standard ASME/ANS RA-Sa-2009.Explain whether the gap assessment was performed for the Internal Events PRA to ASME/ANS RA-Sa-2009 and RG 1.200 Revision 2; if not, determine whether the differences between RG 1.200Revision 1 and RG 1.200 Revision 2 have any impact on the LAR, and provide a summary of theevaluation.Waterford 3 ResponseThe Internal Events peer review (LTR-RAM-II-09-039) was performed using PRA Standard ASMERA-Sb-2005 as clarified by RG 1.200 Revision 1. Both of these documents were the latest availableat the time but have since been revised and current PRA quality is judged against Regulatory Guide1.200 Revision 2 and ASME/ANS RA-Sa-2009.The Internal Events PRA model at Waterford 3 was updated in 2013 to Revision 5. The InternalEvents model documents for this revision (documented in EC 42382) contain individual self-assessments documenting how each relevant High Level Requirement (HLR) and SupportingRequirement are met in the PRA product. These self-assessments were completed usingASME/ANS RA-Sa-2009 clarified by RG 1.200 Revision 2. The Supporting Requirements (SR) forall technical elements, except those associated with internal flooding (which have no impact on theapplicability of the PRA model for the NFPA 805 application) are covered in these documents.SR CategorySource of Self Assessment to ASME/ANS RA

-Sa-2009(AS) - AccidentSequencePSA-WF3-01-AS, R0 - WF3 PSA At-Power Level 1 Accident SequenceAnalysisAPPENDIX A: Assessment of WF3 Accident Sequence CapabilityCategory (ASME RA-Sa-2009)(DA) - DataAnalysisPSA-WF3-01-DA-01, R0 - At-Power Level 1 PRA Plant Specific FailureData DevelopmentTable A.2 ASME Requirements for Data Analysis(HR) - HumanReliability AnalysisPSA-WF3-01-HR, R0 - WF3 At-Power Level 1 Human ReliabilityAnalysis NotebookTable 1. W3 ASME Capability Category Assessment(IE) - Initiating EventAnalysisPSA-WF3-01-IE-02, R0 - WF3 PSA At-Power Level 1 Initiating EventsAnalysisAPPENDIX K: Assessment of WF3 Initiating Events Capability(LE) - Large EarlyRelease FrequencyAnalysisPSA-WF3-01-LE, R0 - Large Early Release Frequency (LERF) ModelTable B-2: Comparison to ASME PRA Standard SupportingRequirements(QU) - QuantificationPSA-WF3-01-QU, R0 - WF3 PSA At-Power Level 1 Integration andQuantification AnalysisTable G-1. ASME/ANS Requirements for Integration and Quantification(SC) - SuccessCriteriaPSA-WF3-01-SC, R0 - WF3 PSA At-Power Level 1 Success CriteriaAnalysisAppendix C: Assessment of W3 Success Criteria Capability Category(SY) - SystemsAnalysisPSA-WF3-01-SY - PSA Model System Analysis Work PackageAppendix 21: Comparison of WF3 System Analysis to ASME PRAStandard toW3F1-2015-0015Page 32 of 44A review of the documents and attachments listed in the above table concluded that with fewexceptions all ASME/ANS RA-Sa-2009 Supporting Requirements associated with the internal eventsPRA model are met at Capability Category II or greater.As part of the NFPA 805 License Amendment Request application, a self-assessment/gapassessment was completed to document the status of the 2009 peer review findings. Waterford 3calculation PRA-W3-05-051 "Documentation of PRA Model Quality for the Waterford 3 NFPA 805LAR" documents this assessment. PRA-W3-05-051 is the site reference for the contents ofAttachment U in the LAR Supplement (Waterford 3 Letter W3F1-2013-0048) on Internal Eventsmodel quality. The PRA-W3-05-051 analysis was a snapshot of the resolutions to the 2009 peerreview, but the assessment was made using the updated ANS/ASME Standard and RegulatoryGuide 1.200 Revision 2. The results of that review noted a few supporting requirements (SRs) asnot meeting Category II. These SRs are summarized in Attachment U of the LAR with anassessment of the impact for the NFPA 805 application.Though the last full scope internal events peer review used the previous Standard and RegulatoryGuide, the current model has been constructed to meet the updated guidance and no significantgaps exist. Any remaining gaps are documented in the Waterford 3 Model Change Request (MCR)database so that they can be tracked and their potential impacts accounted for in applications whereappropriate. All currently open Findings are judged to have a low impact on the PRA model or itsability to support a full range of PRA applications.For completeness, Waterford 3 will include a Gap Assessment as part of the 120 day RAI Submittal.PRA RAI S0790 day response.PRA RAI S08By letter dated June 11, 2014, the response to PRA RAI 16 explains that the updated Fire PRA wasrevised to address sensitive electronics rather than screening them from the analysis. The responseto PRA RAI 16 indicates that expanded ZOIs were used to account for sensitive electronics, but theapproach was not described in the response. Although the treatment of sensitive electronics may beconsistent with recent guidance on modeling sensitive electronics, the licensee's analysis does notcite FAQ 13-0004 (Clarifications on Treatment of Sensitive Electronics, issued December 3, 2013(ADAMS Accession No. ML13322A085), as one of the FAQs used in the Fire PRA. Explain theupdated treatment of sensitive electronics used in the Fire PRA. Include explanation of whether thetreatment of sensitive electronics performed for the Fire PRA is consistent with the guidance inFAQ 13-0004, including the caveats about configurations that can invalidate the approach (i.e.,sensitive electronics mounted on the surface of cabinets or in the presence of louvers or vents). Ifthe approach cannot be justified using available NRC guidance, then replace the current approachwith an acceptable approach as part of the integrated analysis performed in response to PRA RAIS18.Waterford 3 ResponseSensitive electronics as targets in a potential fire ignition source analysis are treated with a zone ofinfluence (ZOI) investigation. The method for this ZOI analysis uses the Fire Dynamics Tools (FDTs)in NUREG-1805 for fire plume temperature as well as radiant heat flux to calculate temperature and toW3F1-2015-0015Page 33 of 44heat flux level at various distances away from various fire sources to arrive at a separation distanceaway from the fire source at which damage is no longer expected based on the target's (sensitiveelectronics in this instance) failure criteria. Bounding heat release rates (HRRs) are used as inputsin the FDT analyses and are taken from the presented 98 th percentile HRRs in NUREG/CR-6850 forvarious ignition sources.The ZOI analysis is carried out on a physical analysis unit (PAU) basis for each counted fixedignition source as well as for transient ignition sources. The methodology and resultant ZOIs for thevarious fixed ignition source bins are presented in PRA-W3-05-006F. The methodology andresultant ZOI for the transient ignition source bin is presented in PRA-W3-05-013. Each potentialignition source's ZOI for sensitive electronics is investigated on a case-by-case basis to determine ifany sensitive electronics are located within that ZOI for the particular ignition source and if anysensitive electronics are located within that particular ZOI the sensitive electronics are listed asfailures for that particular scenario for quantification in the fire PRA results. Similar analyses for oilsource fire scenarios are documented in PRA-W3-05-029.In all of the above analyses, impacts to sensitive electronics use failure criteria taken fromNUREG/CR-6850 as 65° C for temperature impacts or 3 kW/m 2 for heat flux impacts. No credit forexposure duration or shielding is taken during the various ZOI analyses for sensitive electronicsdescribed above.One exception to the NUREG/CR-6850 failure criteria for sensitive electronics is taken in the multi-compartment analysis (MCA) fire modeling for PAU RAB7 using CFAST as documented in PRA-W3-05-030. In the RAB7 MCA fire modeling, a temperature failure criterion of 60° C rather than 65° C isused based on previously determined failure criteria for the components housed within the electricalpanels (cabinets) of PAU RAB7. The heat flux failure criterion of 3 kW/m 2 is also used in PAU RAB7for the MCA. The scenario impacts of the MCA fire modeling of PAU RAB7 using CFAST are usedto supplement the fixed ignition fire source FDT ZOI analyses of PAU RAB7 for scenarioquantification in the fire PRA results.While FAQ 13-0004 was published after the fire PRA work with regard to sensitive electronics wascompleted, the work performed is consistent with or bounding of the methodology presented in FAQ13-0004. Most panels (cabinets) installed at the Waterford 3 plant are ventilated and as such couldnot make use of the enhanced failure criteria proposed in FAQ 13-0004 but rather would be subjectto the lower NUREG/CR-6850 failure criteria which are employed in the current fire PRA analyses forsensitive electronics. The panels (cabinets) within PAU RAB7 could potentially be classified assealed and robust enclosures and make use of the enhanced failure criteria proposed for use in FAQ13-0004; however, the panels (cabinets) evaluated in PAU RAB7 using the FDT analyses for fixedsource ZOI impacts follow the same treatment as other plant locations and make use of the lowerNUREG/CR-6850 failure criteria. Evaluations of the panels (cabinets) within PAU RAB7 for the MCAmade use of the previously listed temperature criterion of 60°C rather than 65°C based on previouslydetermined failure criteria for the components housed within the electrical panels (cabinets) of PAURAB7, and the heat flux failure criterion of 3 kW/m 2 is also used in PAU RAB7 for the MCA. Thesefailure criteria were based on predictions taken on the exterior of the upper surface of the panel(cabinet) as the failure mechanism of interest in the MCA are impacts from a potential hot gas layer,which would be expected to form in the ceiling area above the panels (cabinets) and descend to thetop of the panels (cabinets), thus impacting the top surface of the panel (cabinet) first. No credit forshielding is taken during the MCA analysis of PAU RAB7 for sensitive electronics' failure. toW3F1-2015-0015Page 34 of 44As the current treatment of sensitive electronics is consistent with the guidance of NUREG/CR-6850and consistent with the subsequently published FAQ 13-0004 with the noted more bounding analysisof PAU RAB7 over the enhanced failure criteria proposed in FAQ 13-0004, no additional updates tothe current methods, analysis, or results are deemed necessary.PRA RAI S09120 day response.PRA RAI S10By letter dated June 11, 2014 the response to PRA RAI 28 states, in part, that "the updated analysisno longer uses a floor for joint HEP values," because a simple multiplier approach is used todetermine the HEP of fire event HFEs based on the HEP of Internal Events HFEs. The responseindicates that joint HEPs with probabilities lower than 1E-05 are used. NUREG-1921 discusses theneed to consider a minimum value for the joint probability of multiple HFEs, and NUREG-1792,"Good Practices for Implementing Human Reliability Analysis (HRA)," issued April 2005 (ADAMSAccession No. ML051160213), Table 2-1 recommends that joint HEP values should not be below1E-5. Table 4-3 of EPRI 1021081, "Establishing Minimum Acceptable Values for Probabilities ofHuman Failure Events," provides a lower limiting value of 1E-6 for sequences with a very low level ofdependence.

a.Confirm that each joint HEP value used in the Fire PRA below 1E-5 includes its ownjustification that demonstrates the inapplicability of the NUREG-1792 lower valueguideline.

b.Provide an estimate of the number of these joint HEPs below 1E-5 and at least twodifferent examples of the justification.Waterford 3 ResponseA review of the eight actions is provided below with their inapplicability to NUREG-1792. Theserepresent the eight combinations (JHEPs) utilized in the fire PRA model with probabilities valuesbelow 1E-5. From the aggregate cut set file provided in PRA-W3-05-032, the total Fussell-Vesely core damage importance for the JHEPs is 2.42E-4. Based on the reported CDF of1.55E-5/yr this equates to a CDF contribution from cut sets containing JHEPs of 3.75E-9/yr.Therefore, even a ten-fold increase would have little importance to the current results.No JHEP utilized in the fire PRA is below 1E-6 using the 10x multiplier suggested whentransitioning human actions from internal events to the fire PRA and is consistent with theguidance for screening values given in NUREG-1921. Other JHEPs exist at levels below 1E-5for the internal events PRA but involve actions or systems not credited in the fire PRA, e.g.,auxiliary feedwater.The EPRI guidance ("Establishing Minimum Acceptable Values for Probabilities of HumanFailure Events Practical Guidance for Probabilistic Risk Assessment", Electric Power ResearchInstitute, EPRI 1021081, October 2010) indicates that lower truncation values can be utilized ifjustification is provided to determine very low or no dependence. toW3F1-2015-0015Page 35 of 44JHEP Name JHEP ValueElementsComparison to NUREG-1792, Section 5.3.3.6ZHF-C2-0161.00E-06OHFRCPTRIPQHFCSPEMPThe first action in time occurs early when theCCW is lost and the RCPs are tripped. Thesecond involves a much later time when CSPmakeup is required for EFW and would utilize adifferent operator. They address different safetyfunctions and have different cues. Thisdemonstrates that strict adherence to NUREG-1792 is not necessary and application of EPRI1021081 methodology is reasonable.ZHF-C2-0191.00E-06OHFRCPTRIPHHFISOMINPThe first is associated with RCP trip and thesecond is an action within the RAB to isolaterecirculation lines for HPSI upon receipt ofRecirculation Actuation Signal (RAS). Theseare separated in time, two different operationsstaff and associated with different functions andcues. Very low or no dependence isanticipated. This demonstrates that strictadherence to NUREG-1792 is not necessaryand application of EPRI 1021081 methodologyis reasonable.ZHF-C2-0301.00E-06EHFMANTRNPQHFCSPEMPThe first action is associated with transferringloads to the startup transformers following anABT failure. This would be very close to thetime of trip. The second event is associatedwith CSP refill occurring later in the event and isneeded to support continued EFW operation. .They address different safety functions andhave different cues. Very low or no dependenceis anticipated. This demonstrates that strictadherence to NUREG-1792 is not necessaryand application of EPRI 1021081 methodologyis reasonable.ZHF-C3-0011.00E-06EHFMANTRNPQHFCSPEMPQHFCSPWCTPThe first action is associated with transferringloads to the startup transformers following anABT failure. This would be very close to thetime of trip. The second and third event aresimilar actions associated with CSP refill later inthe event and is needed to support continuedEFW operation which occurs later in the event.The last two would be judged to be completelydependent such that they are a single eventsince performing same action, same indication,same limiting timing. So can be considered thesame as the prior event and the same basisapplies. This demonstrates that strictadherence to NUREG-1792 is not necessaryand application of EPRI 1021081 methodologyis reasonable. toW3F1-2015-0015Page 36 of 44JHEP Name JHEP ValueElementsComparison to NUREG-1792, Section 5.3.3.6ZHF-C3-0031.00E-06OHFRCPTRIPQHFCSPEMPQHFCSPWCTPThe actions associated with the CSP refill tosupport continued EFW operation are similarand can be considered as a single action aswas discussed in the prior event (ZHF-C3-001).Given this assessment then the case is similarto ZHF-C2-016 and the same basis applies.This demonstrates that strict adherence toNUREG-1792 is not necessary and applicationof EPRI 1021081 methodology is reasonable.ZHF-C2-0141.40E-06EHFMANTRNPQHFCSPWCTPThe first action is associated with transferringloads to the startup transformers following anABT failure. This would be very close to thetime of trip. The second occurs when CSP refillis needed to support continued EFW operationwhich occurs later in the event. These areseparated in time, two different operations staffand associated with different functions andcues. Very low or no dependence isanticipated. This demonstrates that strictadherence to NUREG-1792 is not necessaryand application of EPRI 1021081 methodologyis reasonable.ZHF-C2-0372.40E-06OHFPZRPRCPHHFISOMINPThe first is associated with RCP trip and thesecond is an action within the RAB to isolaterecirculation lines for HPSI at recirculation.These are separated in time, two differentoperations staff and associated with differentfunctions and cues. Very low or no dependenceis anticipated. This demonstrates that strictadherence to NUREG-1792 is not necessaryand application of EPRI 1021081 methodologyis reasonable.ZHF-C2-0394.00E-06EHFMANTRNPQHFEFWFLOPThe first action is associated with transferringloads to the startup transformers following anABT failure. This would be very close to thetime of trip. The second occurs when CSP refillis needed to support continued EFW operationwhich occurs later in the event. These areseparated in time, two different operations staffand associated with different functions andcues. Very low or no dependence isanticipated. This demonstrates that strictadherence to NUREG-1792 is not necessaryand application of EPRI 1021081 methodologyis reasonable. toW3F1-2015-0015Page 37 of 44As demonstrated the identified combinations all exhibit at least two of the following characteristicsthat reduce the dependence to very low or independent: Considerable separation in time. At least one action would not occur until a substantial period after the fire would beexpected to be suppressed. Utilize different staff to perform the function. Are based on different cues. Support different safety functions.Therefore the use of a value a decade lower than the NUREG-1792 minimum value is deemedappropriate and consistent with the guidance discussed in EPRI 1021081 that the cut off should notbe driven by a strict limit but should be based on reasonable assessment of the dependence incases where the events exhibit weak or negligible dependence.PRA RAI S11By letter dated June 11, 2014 the response to PRA RAI 30 explains that as a result of the revisedFire PRA the "generic" non-suppression factors originally used were removed and non-suppressionprobability values from NUREG/CR-6850, Supplement 1, were incorporated. The updated approachis not described in the response; however, the NRC staff reviewed the description in the licensee'sanalysis and notes potential inconsistencies. The licensee's analysis identifies several PAUs (i.e.,FPH, RAB 15A, RAB 16A, RAB 20, RAB 40, and RAB 41) for which automatic suppression systemsare used as the bases for non-suppression probabilities credited in the Fire PRA. However, thesesuppression systems appear not to be required in Table C-2 of the LAR supplement for riskreduction. Please explain this apparent inconsistency.Waterford 3 ResponseThe fire areas listed (FPH, RAB 15A, RAB 16A, RAB 20, RAB 40, and RAB 41) are all remainingDeterministic and will be maintained under NFPA 805 chapter 4.2.3.2 and not under thePerformance Based (i.e. risk informed) approach of NFPA 805 chapter 4.2.4.2. No credit is taken inthe Fire PRA for automatic suppression in the FPH fire area. Calculation PRA-W3-05-031"Development of Fire Non-suppression Factors for WF3 FPRA Scenarios" only notes that this areahas installed and functional automatic suppression and that credit would be permissible.Automatic suppression was credited in the other listed areas (RAB 15A, RAB 16A, RAB 20, RAB 40,and RAB 41) for the base risk numbers calculated for these deterministic areas. All rooms wereverified to have installed and functioning automatic suppression systems prior to being credited(PRA-W3-05-031). The credit was applied in an effort to accurately model the as-built, as-operatedplant. These suppression features were not credited in fire risk evaluations as these areas have novariances from deterministic requirements (VFDRs). There is no need to credit these suppressionfeatures for risk reduction.These areas all have reasonably low CDF and LERF values and no calculated values for deltaCDF/LERF. Moreover, the remaining credited areas (RAB 15A, RAB 16A, RAB 20, RAB 40, andRAB 41) will no longer have licensing actions to require suppression under NFPA 805. Because ofother requirements, the suppression systems in these areas will be maintained. In addition, these willbe included in the scope of the NFPA 805 Monitoring program. toW3F1-2015-0015Page 38 of 44PRA RAI S12120 day response.PRA RAI S1390 day response.PRA RAI S14By letter dated June 11, 2014 the responses to PRA RAI 48 and 43.f identifying modifications thatare credited in the PRA do not appear to be fully consistent with the updated Table S-1 in the LARsupplement. Please identify which modifications in the updated Table S-1 are credited in the PRA.Waterford 3 ResponseThe following modifications were originally included in LAR Attachment S but were deleted from theLAR Supplement Attachment S: S1-2, S1-3, S1-4, and S1-6. The response to PRA RAI 43.f insubmittal W3F1-2014-0025 indicated that S1-7 was no longer required and was withdrawn.S1-7 is not credited in the FPRA analysis.The following modifications are noted as "In FPRA" in the LAR Supplement Attachment S and arecredited for risk reduction: S1-5, S1-8, and S1-14.S1-5 is an EFRBS (3M fire wrap) in RAB 6. This is credited in the PRA to prevent fire damage to thewrapped cables.S1-8 is a fire/heat detector upgrade in several fire areas. The current detectors have incorrect (lowertemperature) factory set point trip range settings for the normal and accident design temperatures forthe applicable fire areas. This modification is not a result of fire risk evaluations, but is assumed inthe FPRA model for the appropriate Fire Area detection systems. The detection supports firesuppression and fire brigade response which is credited in the FPRA for several of the areasincluded in the modification.S1-14 involves the removal of combustible material from RAB 27. The modification is to remove thecombustible material from the area. This configuration change is credited in the fire PRA model. Themodification provides a decrease from the threat of the effects of a fire originating from a fire in RAB27. This change is not associated with a variance from deterministic requirement. The result of themodification is lower overall CDF (little to no impact on delta CDF). Significant combustible materialwas removed and the fire impacts are significantly reduced.PRA RAI S15By letter dated June 11, 2014 the responses to PRA RAI 49 and PRA RAI 45.e explain thatquantitative uncertainty analysis was developed for the updated Fire PRA. However, neither theseresponses nor the licensee's analysis discuss whether or how state of knowledge correlation(SOKC) is accounted for in the quantitative uncertainty analysis. Mean CDF and LERF values can toW3F1-2015-0015Page 39 of 44be affected by SOKC and should be accounted for as part of statistical parametric uncertaintyanalysis. It is not clear whether SOKC was taken into account in the quantitative uncertaintyanalysis performed for the updated Fire PRA. PRA Standard Supporting Requirement (SR) QU-A3(referenced by FQ-A4) and QU-E4 (referenced by FQ-E4) require that the CDF be estimatedaccounting for the SOKC between event probabilities. The SOKC should include consideration ofcorrelation between probability distributions for Internal Event and Fire PRA parameters (i.e.,component type failure mode probabilities, hot short probabilities, and non-suppressionprobabilities). Summarize the impact of the SOKC correlation on the results reported in Attachment W.Waterford 3 ResponseThe state of knowledge correlation (SOKC) addresses the uncertainty that exists when two or moreelements in a specific cut set are from the same data source such as two or more basic eventsassociated with the diesel generator failing to start. These are typically identified by the commontype code listing found in the CAFTA database file (.RR extension). The SOKC indicates that inthese instances the probability of the two events is not represented by the product of the meanvalues for the like events but rather the product of the mean plus the associated variance. Notincluding this consideration may lead to underestimation of the cut set contribution and in doing sounderestimate the core damage frequency (CDF) or large early release frequency (LERF)contribution. For the Waterford 3 analysis, the results indicate that the calculated estimate for CDFand LERF is not sensitive to the SOKC, as described below.The Waterford 3 uncertainty assessment was conducted using the UNCERT code which is based ona sampling approach to derive the overall distribution. The Waterford 3 uncertainty applies thecomponent uncertainty characteristics from the internal events model, and the uncertaintyparameters associated with fire ignition sources and suppression factors taken from NUREG/CR-6850 and Supplement 1. However, the contribution from SOKC was not explicitly included in theuncertainty analysis for the PRA model.No variance is specifically provided for the circuit failure probability so it was assumed that thevariance was bounded by 10% of the probability of failure or 0.06. This is based on assuming agamma distribution using=6,=10 ("NIST/SEMATECH e-Handbook of Statistical Methods",http://www.itl.nist.gov/div898/handbook/). Using these values the variance is defined as 0.06. Thesampling utilizes the associated type code parameters and selects the same value for all eventssuch that the results are completely correlated but this in itself does not address the SOKCcontribution.The estimate of the SOKC impact is developed by a review of the cut sets developed for theuncertainty assessment (PRA-W3-05-032 "Uncertainty Assessment for Waterford 3 Fire PRA,Revision 2"). This included a review of the top 80% of CDF and LERF cut sets to identify any cut setwith the conditions requiring a SOKC contribution, had multiple occurrences of the same componentfailure. From this assessment only twelve CDF cut sets were identified as meeting the condition.These are listed in Table 1. toW3F1-2015-0015Page 40 of 44Table 1. CDF Cut Sets with SOKC Conditions MetCS Frequency(/yr)% of CDFCut Set Elements (note pairs highlighted in yellow)26 8.02E-08 0.59%TGB-F43 AA_FAIL3AS AA_FAIL3BSEMPOILTRAAEMPOILTRBATGB-F43-NSF30 7.45E-08 0.55%TGB-F43 AA_FAIL3AS AA_FAIL3BSEDG0DG3ASFEEDG0DG3BSFETGB-F43-NSF111 1.40E-08 0.10%TGB-F43 AA_FAIL3AS AA_FAIL3BSECB12A3M4DECB312B8MDTGB-F43-NSF112 1.40E-08 0.10%TGB-F43 AA_FAIL3AS AA_FAIL3BSECB12A3M4DECB313B8MDTGB-F43-NSF113 1.40E-08 0.10%TGB-F43 AA_FAIL3AS AA_FAIL3BSECB12A3M4DECB314B2MDTGB-F43-NSF114 1.40E-08 0.10%TGB-F43 AA_FAIL3AS AA_FAIL3BSECB312A8MDECB312B8MDTGB-F43-NSF115 1.40E-08 0.10%TGB-F43 AA_FAIL3AS AA_FAIL3BSECB312A8MDECB313B8MDTGB-F43-NSF116 1.40E-08 0.10%TGB-F43 AA_FAIL3AS AA_FAIL3BSECB312A8MDECB314B2MDTGB-F43-NSF117 1.40E-08 0.10%TGB-F43 AA_FAIL3AS AA_FAIL3BSECB312B8MDECB313A8MDTGB-F43-NSF118 1.40E-08 0.10%TGB-F43 AA_FAIL3AS AA_FAIL3BSECB312B8MDECB314A2MDTGB-F43-NSF119 1.40E-08 0.10%TGB-F43 AA_FAIL3AS AA_FAIL3BSECB313A8MDECB313B8MDTGB-F43-NSF120 1.40E-08 0.10%TGB-F43 AA_FAIL3AS AA_FAIL3BSECB313A8MDECB314B2MDTGB-F43-NSF121 1.40E-08 0.10%TGB-F43 AA_FAIL3AS AA_FAIL3BSECB313B8MDECB314A2MDTGB-F43-NSF122 1.40E-08 0.10%TGB-F43 AA_FAIL3AS AA_FAIL3BSECB314A2MDECB314B2MDTGB-F43-NSFA similar search was performed for the LERF cut set results and is documented in Table 2. Notethat all SOKC contributions are associated with the same hot short event pair associated with CARSisolation valves leading to a containment isolation path being open. Paired events are identified inyellow. toW3F1-2015-0015Page 41 of 44Table 2. LERF Cut Sets with SOKC Conditions MetCS Frequency(/yr)% ofLERFCut Set Elements (note pairs highlighted in yellow) 18.64E-0718.1% CRA6T CROSSPROB1 CROSSPROB2 EF-CI-JPURGEOPEN CRA6T-SF CRA6T-NSF 42.55E-07 5.4CRA7T CROSSPROB1 CROSSPROB2 EF-CI-JPURGEOPEN CRA7T-SF CRA6T-NSF 72.03E-07 4.37BT05 AALOSPEVTB CROSSPROB1 CROSSPROB2 EF-CI-JPURGEO PEN05T01-NSF 81.85E-07 3.97DT01 CROSSPROB1 CROSSPROB2 EF-CI-JPURGEOPEN 05T01-NSF 91.58E-07 3.37-F07-MCA-01 CROSSPROB1 CROSSPROB2 EF-CI-JPURGEOPEN7-F01-MCA-01-NSF10 1.58E-07 3.37-F08-MCA-01 AALOSPEVTB CROSSPROB1 CROSSPROB2 EF-CI-JPURGEO PEN7-F01-MCA-01-NSF11 1.58E-07 3.3 CRA3M CROSSPROB1 CROSSPROB2 EF-CI-JPURGEOPEN CRA3M-SF CRA1M-NSF12 1.08E-07 2.3 CRA4M CROSSPROB1 CROSSPROB2 EF-CI-JPURGEOPEN CRA4M-SF CRA1M-NSF13 1.05E-07 2.27-F09-MCA-01 AALOSPEVTB CROSSPROB1 CROSSPROB2 EF-CI-JPURGEO PEN7-F01-MCA-01-NSF14 1.04E-07 2.2CRA3S CROSSPROB1 CROSSPROB2 EF-CI-JPURGEOPEN CRA3S-SF CRA1M-NSF15 7.85E-08 1.67-F06-MCA-01 CROSSPROB1 CROSSPROB2 EF-CI-JPURGEOPEN7-F01-MCA-01-NSF16 7.54E-08 1.6 CRA1M CROSSPROB1 CROSSPROB2 EF-CI-JPURGEOPEN CRA1M-SF CRA1M-NSF17 6.19E-08 1.3 CRA2M CROSSPROB1 CROSSPROB2 EF-CI-JPURGEOPEN CRA2M-SF CRA1M-NSF18 5.24E-08 1.17-F10-MCA-01 CROSSPROB1 CROSSPROB2 EF-CI-JPURGEOPEN7-F01-MCA-01-NSF19 5.08E-08 1.17DT03 CROSSPROB1 CROSSPROB2 EF-CI-JPURGEOPEN 05T01-NSF21 4.74E-08 1.0CRA1S CROSSPROB1 CROSSPROB2 EF-CI-JPURGEOPEN CRA1S-SF CRA1M-NSF22 4.31E-08 0.9CRA8T CROSSPROB1 CROSSPROB2 EF-CI-JPURGEOPEN CRA8T-SF CRA6T-NSF23 3.95E-08 0.87DT02 CROSSPROB1 CROSSPROB2 EF-CI-JPURGEOPEN 05T01-NSFSumming the LERF contributions of these cut sets (Table 2) yields a frequency value of 2.75E-6/yr. toW3F1-2015-0015Page 42 of 44To determine the SOKC contributions for CDF the identified cut sets were adjusted to include thevariance in the probability term for the component pairs. The impact of the state of knowledgecorrelation is to increase the contribution of the correlated basic events. The increase is equal to thevariance. The equation is represented by:P(SOKC) = P(A) x P(B) + VarThis equation is used along with the cut sets to define the increase. The first cut set (#26) is used asan example.The frequency of the cut sets is 8.02E-8/yr. The component probability is 5.69E-3 with a variance of4.62E-7. The first step is to remove the existing contribution P(A) x P(B) [5.69E-3 x 5.69E-3 =3.2376E-5]. The "partial" cut set value is 2.48E-3. Multiplying this value by the variance (4.62E-7)yields the SOKC increaseThe results were then summed to determine the additional CDF value. Table 3 summarizes theprocess for each identified CDF cut set in Table 1.Table 3. Evaluation of SOKC Contribution to CDF CS CSFrequency (/yr)SOKCComponentProbabilityType Code VarianceValuePartial CSFrequency (/yr)SOKC Contribution

(/yr)268.02E-085.69E-03 MP_A4.62E-072.48E-031.14E-09 307.45E-085.50E-03DG_FE8.07E-082.46E-031.99E-10 1111.40E-082.38E-03 CB_D5.95E-062.47E-031.47E-08 1141.40E-082.38E-03 CB_D5.95E-062.47E-031.47E-08 1151.40E-082.38E-03 CB_D5.95E-062.47E-031.47E-08 1161.40E-082.38E-03 CB_D5.95E-062.47E-031.47E-08 1171.40E-082.38E-03 CB_D5.95E-062.47E-031.47E-08 1181.40E-082.38E-03 CB_D5.95E-062.47E-031.47E-08 1191.40E-082.38E-03 CB_D5.95E-062.47E-031.47E-08 1201.40E-082.38E-03 CB_D5.95E-062.47E-031.47E-08 1211.40E-082.38E-03 CB_D5.95E-062.47E-031.47E-08 1221.40E-082.38E-03 CB_D5.95E-062.47E-031.47E-08TOTAL1.48E-07The SOKC contribution is obtained by taking the initial cut set and dividing out the identified pairprobability to arrive at a partial CS frequency. This is then multiplied by the variance to arrive at theSOKC contribution. Using the process an additional frequency of 1.48E-7/yr is calculated.A similar process can be used for the LERF. However, since in all cases the same pair is involved, itcan be simplified to a single calculation involving the sum of the contribution as presented in Table 4. toW3F1-2015-0015Page 43 of 44Table 4. Evaluation of SOKC Contribution to LERF LERFFrequency (/yr)SOKCComponentProbabilityBasicEvent(s)VarianceValuePartial CSFrequency (/yr) SOKC Contribution (/yr)2.75E-060.6Crossprob0.067.63E-064.58E-07Table 5 summarizes the assessment of the SOKC contributions for CDF and LERF.Table 5. Comparison of Baseline to Adjusted CDF and LERFFigure of MeritBaselineAdjusted for SOKCDifferenceCDF (/yr)1.32E-5/yr1.33E-5/yr1.1%LERF (/yr)4.72E-6/yr5.18E-6yr9.7%The results indicate that the calculated estimate for CDF and LERF is not sensitive to the SOKC.Based on the review, the basis for the low contribution is that, for the most part, the results aredominated by cut sets that do not represent combinations of component random failures. Fire-induced failures of components are completely dependent within a scenario with no severity factorincluded such that there is not a SOKC adjustment required.For the LERF scenarios the result is controlled by cut set contributions that are devoid of anycomponent failures not completely dependent on the presence of the fire. The 9.7% difference isassociated with cut sets including circuit failure probabilities related to the CARS system. Theimpact is strongly associated with the assumption on variance associated with the hot shortprobability.The approved analysis method for calculating the impacts of hot shorts is NUREG/CR-7150, Vol.2/EPRI 3002001989, "Joint Assessment of Cable Damage and Quantification of Effects from Fire(JACQUE-FIRE)", which provides updated estimates for the probability of hot short and also includesestimates for uncertainty. Considering information from this source indicates a variance level of 0.01is more likely which would significantly reduce the impact of SOKC and would further support theconclusion that the SOKC considerations do not impact the results and conclusions derived from thefire PRA.Waterford 3 will add to implementation item S2-22 to employ NUREG/CR 7150, Vol 2 methods.PRA RAI S1690 day response.PRA RAI S1790 day response. toW3F1-2015-0015Page 44 of 44PRA RAI S18120 day response.PRA RAI S19120 day response. toW3F1-2015-0015Revised Attachment ANFPA 805 Chapter 3 Requirement 3.11.3Waterford 3 NFPA 805 License Amendment Request toW3F1-2015-0015Page 1 of 1Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design ElementsNFPA 805 Ch. 3 ref.Requirements/GuidanceCompliance StatementCompliance BasisReference Documents3.11.3 Fire BarrierPenetrations. (2)3.11.3* (2) NFPA 90A, Standard for theInstallation of Air-Conditioning and VentilatingSystemsComplies via PreviousApprovalHVAC duct penetrations through fire areaboundaries separating the corridor/vestibule portionof Fire Area RAB 3A (Heating and VentilationRoom, Elevator Machine Room) from Fire ZonesRAB 8B (Switchgear Room B) and RAB 8C(Switchgear Room A/B) (Fire Area RAB 8) andFire Area RAB 25 (Equipment Access Area) fromFire Area RAB 32 (Pipe Penetrations: AuxiliaryComponent Cooling Water Pumps) do not haverated fire dampers (Reference LAR, Attachment K,Deviations 16 and 42). The lack of fire dampers inHVAC ductwork separating Fire Area RAB 3Afrom Fire Zones RAB 8B and RAB 8C and FireArea RAB 25 from RAB 32 were previouslyapproved by the NRC in letter dated January 17,1995 (ILN95-0017) and in NUREG 0787,Supplement No. 8, dated December, 1984respectively.ILN 95-0017 Rev. 000 [All] - SafetyEvaluation by the Office of NuclearReactor Reg. - Re-Evaluation of PreviousExemption from the Requirements ofAppendix R to 10 CFR Part 50, Sect.III.G.2 - Entergy Operations, Inc,Waterford Steam Electric Station, Unit 3- Docket No. 50-382NUREG-0787, Supplement No. 8 Rev.12/84 [Section 9.5.1.3(2)] - SafetyEvaluation Report - related to theoperation of Waterford Steam ElectricStation Unit No. 3 - Docket No. 50-382WF3-FP-13-00005 Rev. 000 [Attachment7.1, Deviations 16 & 42] - WF3 Reviewof Licensing Actions for NFPA 805TransitionComplies with use ofEEEE'sWaterford 3 compliance with the requirements inNFPA 90A, 1974 Edition, for air-conditioning andventilation systems credited in Chapter 4 to meetthe Nuclear Safety Performance Criteria, isdocumented in Engineering Report No. WF3-FP 00017 and engineering equivalency evaluationsdocumented in WF3-FP-11-00003 and in WF3-FP-13-00006. No other NFPA standards weredetermined to be applicable.WF3-FP-10-00017 Rev. 001 [All] - WF3Code Compliance Report for NFPA 90A,1974 Edition Standard for Installation ofAir Conditioning and VentilatingSystemsWF3-FP-11-00003 Rev. 000 [Section6.2.] - WF3 Review and Evaluation ofFire Protection Engineering Evaluationsfor NFPA 805WF3-FP-13-00006 Rev. 000[Attachments 8.1, 8.3 & 8.4] - Evaluationof Fire Area Boundaries