Joseph Pacher Fax: CENGOS - NRC: Home Page · 2014. 10. 3. · Joseph Pacher Site Vice President...

Joseph PacherSite Vice President

CENGOSa joint venture of

Constellation eDFEnergy, 10

Office: 585-771-5200Fax: 585-771-3943

Email: [email protected]

December 17, 2013

U.S. Nuclear Regulatory CommissionWashington, DC 20555-0001

ATTENTION:

SUBJECT:

Document Control Desk

R.E. Ginna Nuclear Power PlantRenewed Facility Operating License No. DPR-18Docket No. 50-244

Response to Request for Additional Information RE: License Amendment totransition to NFPA 805

REFERENCES: (a) Letter from Mr. Joseph E. Pacher (Ginna LLC) to Document Control Desk(NRC) dated March 28, 2013, Subject: License Amendment RequestPursuant to 10 CFR 50.90: Adoption of NFPA 805, Perfornmance-BasedStandard for Fire Protection for Light Water Reactor Electric GeneratingPlants (ML 1 3093A064)

By Reference (a), R.E. Ginna Nuclear Plant, LLC (REG) submitted a request for the adoption of NFPA805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants.On October 9, 2013, the NRC requested additional information regarding this submittal. Attached pleasefind the first set of three responses to the staffs questions and revised documentation. There are noregulatory commitments identified in this letter.

Should you have any questions regarding this submittal, please contact Thomas Harding at 585-771-5219.

I declare under penalty of perjury that the foregoing is true and correct. Executed on December 17, 2013.

S' lcerely,

JP/KCAttachument: (1) 60-Day Responses to Request for Additional Information for NFPA 805

R.E. Ginna Nuclear Power Plant, LLC1503 Lake Road, Ontario, New York 14519-9364

Document Control DeskDecember 17, 2013

cc: NRC Regional Administrator, Region INRC Project Manager, GinnaNRC Resident Inspector, GinnaA.L. Peterson, NYSERDA

Attachment (1)

60-Day Responses to Request for Additional Information for NFPA 805


FM RAI 02

American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) StandardRA-S-2008, "Standard for Level 1/Large Early Release Frequency Probabilistic RiskAssessments for Nuclear Power Plant Applications.", Part 4, requires damage thresholds beestablished to support the FPRA. Thermal impact(s) must be considered in determining thepotential for thermal damage of SSCs. Appropriate temperature and critical heat flux criteriamust be used in the analysis.

1. Describe how the installed cabling in the power block was characterized, specifically withregard to the critical damage threshold temperatures and critical heat flux for thermoset andthermoplastic cables as described in NUREG/CR-6850.

2. During the audit, it was discussed that four cables, which run through one of the batteryrooms (BRIA) are credited with thermoset damage thresholds for the purposes of firemodeling and PRA calculations. Provide justification for treating these cables differentlythan the rest of the cables at the plant.

Response

1. All targets are assumed to be of thermoplastic material for damage criteria (i.e., 205 C and 6kW.m 2, NUREG/CR-6850 Table H-i) with one exception, which is discussed later in theresponse to this RAI. The assumption of thermoplastic damage criteria is documented inG1-FSS-FOO1, section B.1.

2. The exception to thermoplastic damage criteria applies to cables in 5 conduits in BatteryRoom A These 5 conduits were verified to be of Thermoset insulation material (Ginna KeyInput 83, EWR-1444, and PCR-98-015) and the damage criteria of 330 C and 11 kW/m 2

described in NUREG/CR-6850 Table H-1 was assigned to them. The treatment of the 5conduits in Battery Room A is documented in G1-FSS-FOO1, Appendix B.

FM RAI 021

Page 1 of 114


FM RAI 05

Section 4.5.1.2, "Fire PRA" of the Transition Report states that fire modeling was performed aspart of the Fire PRA development (NFPA 805, Section 4.2.4.2). This requires that qualified firemodeling and PRA personnel work together. Furthermore, Section 4.7.3, "Compliance withQuality Requirements in Section 2.7.3 of NFPA 805," of the Transition Report states that"Cognizant personnel who use and apply engineering analysis and numerical methods insupport of compliance with 10 CFR 50.48(c) are competent and experienced as required bySection 2.7.3.4 of NFPA 805."

Regarding qualifications of users of engineering analyses and numerical models (i.e., firemodeling techniques):

1. What are the licensee's requirements to qualify personnel for performing fire modelingcalculations in the NFPA 805 transition?

2. What is the process for ensuring that the fire modeling personnel meet those qualifications,not only before the transition but also during and following the transition?

3. When fire modeling is performed in support of FPRA, how is proper communication betweenthe fire modeling and Fire PRA personnel ensured?

Response

1. Fire modeling calculations will be performed by a Fire Protection Engineer who meets thequalification requirements of Section 2.7.3.4 of NFPA 805. The qualification process willfollow the guidance of ACAD 98-004 and CENG procedure CNG-TR-1.01-1000, "Conduct ofTraining." All those performing Fire Modeling for the Fire PRA require qualification to FirePRA (ESP-FIQ-FPRA) and are documented per CNG-TR-1.01-1014. This qualificationincludes basic fire modeling techniques as part of the qualification as well as Fire PRAtechniques. The CENG PRA Engineering Supervisor reviews experience and education forall Fire Modeling work. Those performing detailed fire modeling analysis using tools suchas CFAST (Consolidated Model of Fire and Transport) or FDS (Fire Dynamics Simulator)will be required to have the relevant education and experience in fire modeling to performthe analysis.

2. In the case of the initial fire modeling, the vendor provided the credentials of the firemodelers, which were reviewed and approved by CENG supervision per procedure CNG-TR-1.01-1000. During and following transition, the existing engineering staff will continue tobe knowledgeable in fire modeling techniques, including interpreting and maintaining the firemodeling database. If new fire modeling personnel are needed in the future, their credentialswill also be reviewed and approved by CENG supervision. Also, per CNG-TR-1.01-1014,Section 4.6.0 requires that Engineering Supervisors have responsibility for "Verifyingqualifications prior to assigning personnel to perform job performance requirementindependently". This requires verifying the qualifications in the training server to verify thequalification is current.

FM RAI 051

Page 2 of 114


3. Throughout the Fire PRA process, the Fire Protection Engineers (FPE) who conducted thefire modeling and the PRA engineers maintained frequent communications. During thedevelopment phase of the Fire PRA, the fire modeling personnel populated the fire modelingdatabase (FMDB), in which all the relevant fire modeling inputs are maintained. Thescenario frequencies, which are produced by the FMDB, are electronically sent to the PRAengineers, who perform the quantification. Both the FPEs and the PRA engineersparticipated in the cutset review meetings during the development of the Fire PRA. The firemodeling database will be maintained under the responsibility of the FPE and the PRAengineers.

FM RAI 052

Page 3 of 114


FPE RAI 01

The compliance statement for License Amendment Request (LAR) (Agency wide DocumentsAccess and Management System (ADAMS) Accession Number ML13093A065), Attachment A,Table B-i, Section 3.3.6 is "complies with use of evaluation". National Fire ProtectionAssociation Standard 805, "Performance-Based Standard for Fire Protection for Light-WaterReactor Electric Generating Plants" (NFPA 805), Section 3.3.6, requires metal roof coverings beClass A as determined by tests described in NFPA Standard 256, "Standard Methods of FireTests of Roof Coverings". The compliance basis states that "all metal roofs were specified to beFM class I.. The basis further describes the difference between the scope of testing for aClass A rating versus the scope of testing for a Factory Mutual (FM) Class 1 rating as detailed inFM Approval Standard 4470, "Approval Standard for Single-Ply, Polymer-Modified BitumenSheet, Built-Up Roof (BUR) and Liquid Applied Roof Assemblies for use in Class 1 andNoncombustible Roof Deck Construction." Section 4 of FM 4470 indicates that Class A, B, andC materials may achieve an FM Class I rating provided the conditions of acceptance for spreadof flame, intermittent flame and burning brand are met. Provide a justification demonstratinghow the requirements of Class A are met using the alternate classification.

Response

The following is a description of the changes to LAR Attachment A, Table B-i, Section 3.3.6:

Change compliance statement associated with the B-1 table section 3.3.6 to:"Complies via Previous Approval" and"Complies with use of Evaluation"

Delete compliance basis in its entirety and replace with:

Compliance basis statement associated with "Complies via Previous Approval" to:For the Fire Protection Evaluation documented via, letter From L.D. White, RG&E, To A.Schwencer, NRC "Fire Protection Evaluation (Gilbert Associates Report No. 1936) dated02/24/1977 Accession # 4006003751 [RG010540] [Sec. 5.0, Guideline D.1.e, pg. 5-23] theresponse to BTP APCSB 9.5-1, Appendix A, Guideline D.1.e stated, "All metal roofs werespecified to be FM Class I except the Screen House. The FHA describes the roof constructionfor this area and outlines fire protection requirements."

Compliance basis statement associated with complies with use of Evaluation to:"As part of the NFPA 805 transition process, Design Analysis DA-ME-08-013 was issued todocument the acceptability of the Screen House roof assembly. Based on the low combustibleloading inside the Screen House and the passive and active fire suppression features tomitigate a fire, the lack of a Class A roof was evaluated in the design analysis as beingacceptable. As stated in the analysis, "There are robust fire mitigating design features withinthe Screen House which include: curbing around the diesel fire pump and oil storage tank, adrainage system within the curbed area, an automatic deluge sprinkler system S17 and smokedetectors installed over the cable trays in the SH basement, an automatic wet-pipe suppressionsystem S18 and detection system Z26 installed above the fire pumps and service water pumpsin the SH Operating floor, inside and outside hose reel coverage, and a large overhead doorlocated in the south wall and roof exhaust fans to be used for potential smoke removal." Thecredited bases for acceptance are valid and meet applicable quality requirements."

FPE RAI 011

Page 4 of 114


"For new construction, A-202 [3.3.13] is the administrative control to utilize Class A roofcoverings as determined by tests described in NFPA 256-2003."

FPE RAI 012

Page 5 of 114

60-Day Responses to Request for Additional Information for NFPA 805Attachment A - NEI 04-02 Table B-1 Transition of Fundament Fire Protection Program & Design ElementsConstellation Energy Nuclear Group

NFPA 805 Requirements/Guidance Compliance Compliance Basis ReferencesChapter 3 StatementReference

3.3.6Roofs

Metal roof deck construction shall be designed andinstalled so the roofing system will not sustain aself-propagating fire on the underside of the deckwhen the deck is heated by a fire inside thebuilding. Roof coverings shall be Class A asdetermined by tests described in NFPA 256,Standard Methods of Fire Tests of Roof Coverings. (

use-ef

Complies viaPreviousApproval

Complies withuse of

Evaluation

For the Fire Protection Evaluation documented via,letter From L.D. White, RG&E, To A. Schwencer,NRC "Fire Protection Evaluation (GilbertAssociates Report No. 1936) dated 02/24/1977)Accession # 4006003751 (RG010540] [Sec. 5.0,Guideline DAle, pg. 5-23], the response to BTPAPCSB 9.5-1, Appendix A, Guideline D.1.e stated,"All metal roofs were specified to be FM Class Iexcept the Screen House. The FHA describes theroof construction for this area and outlines fireprotection requirements.' A Class. A rating rolatcs

RG010540, RG&E FireProtection Evaluation,2/24/77DA-ME-08-013,rev. 0,Evaluation of ScreenHouse Roof AssemblyA-202,rev. 03100, TheFire Protection Programand Ginna Station StaffResponsibilities for FireProtection

to oxternar fire ponornmanco w ers an ,R-,-- -1aG1 roqui-rs that a roof dock acccmbly is subje6t4e

te aR corice of toctcInton fire,eoxtoFRral fire.-windUplift roSiratanco, foo9t traffic, cForroion FrozictaRG9,im..pat r•tAn., and suceopt~bility te heatdamage as doscr.ibod in FM 4479 ApprovalStandward- for Clacs 1 Roof CoVors. A rootassembly must pass all those tests in erdor o ga!9a Clace 1 d6signation, FMV Clao 1. inncludoesFIa.447•n•.d FIVI440 Approval Standard for CGlse1. n'cuatod Stool Docks Roofs. A Class 1arefoAmbly cn b- ,ube 4tuo ! for a Class A, B or- Croof Annomblv foi~nc it C inmo AAocR.'AtI'.'l.

As part of the NFPA 805 transition process, DesignAnalysis DA-ME-08-013 was issued to documentthe acceptability of the Screen House roofassembly. Based on the low combustible loadinginside the Screen House and 'the passive andactive fire suppression features to mitigate a fire,the lack of a Class A roof was evaluatihe-design analysis as beinq acceptable, As stated in

reare ro aus fir,•• mitigatingdesign features within the Screen House whichinclude: curbing around the diesel fire pump and oilstorage tank, a drainage system within the curbedarea, an automatic deluge sprinkler system S17and smoke detectors installed over the cable traysin the SH basement, an automatic wet-pipesuppression system S18 and detection system Z26

ý1_ JGinna LAR Rev 0 Page A-19

Page 6 of 114

60-Day Responses to Request for Additional Information for NFPA 805Attachment A - NEI 04-02 Table B-1 Transition of Fundament Fire ProtectionConstellation Energy Nuclear Group Program & Design Elements

Compliance Basis References

I

3.3.6Roofs

(Continued)

installed above the fire pumps and service waterpumps in the SH Operating floor, inside andoutside hose reel coverage, and a large overheaddoor located in the south wall and roof exhaust

to be used for ote.Lamoke remoal••The rfor acceptance are valid andmeet applicable quality requirements.

For new construction, A-202 [3.3.13] is theadministrative control to utilize Class A roofcoverings as determined by tests described inNFPA 256 - 2003.

U _____________________________ I _________ _____________________________ _________________

Ginna LAR Rev 0 Page A-20


Page 7 of 114


FPE RAI 02

The compliance basis for LAR Attachment A, Table B-i, Section 3.3.7 states that flammablegas storage in fire areas TB-1 and TB-2 is such that the total content is less than 400 standardcubic feet (scf). The basis further states that NFPA Standard 55, 2010 edition, "CompressedGases and Cryogenic Fluids Code," Section 10.1.1, indicates that the chapter "does not apply toindividual systems using containers having a total hydrogen content of less than 400 scf, if eachsystem is separated by a distance not less than 5ft." Describe the configuration of flammablegas storage in these fire areas and the administrative controls used to ensure the volume offlammable gas is maintained below 400 scf.

Response

The following description will be added to LAR Attachment A, Table B-I, Section 3.3.7:

The gas bottle racks are anchored to the concrete pedestal in the air ejector area located in theTurbine Building Mezzanine, and are also located greater than 5' away from the rack located inthe Turbine Building Basement near the elevator. The racks are not located in the vicinity ofany safety related equipment. The racks are located in a non-seismic building with no seismiccategory 1 or SR equipment; therefore, the rack is not required to meet seismic criteria. Signsare also provided to ensure total hydrogen content less than 400 scf.

FPE RAI 021

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3.3.7Bulk Flammable

Gas Storage

Bulk Flammable Gas Storage. Bulk compressed orcryogenic flammable gas storage shall not bepermitted inside structures housing systems,equipment, or components important to nuclearsafety.

Complies withuse of

Evaluation

K

Provisions are in place via A-202 [Sec. 3.7] andFPS-16 [Attach. 3] that prohibit bulk storage ofcompressed or cryogenic flammable gas withinstructures housing systems, equipment, orcomponents important to safety. With the exceptionof the Primary and Secondary Hydrogen StorageBuildings, FPS-16 requires that bulk compressedor cryogenic flammable gas be stored outdoors, aminimum of 50 feet away from buildings,structures, and equipment.

The Primary and Secondary Hydrogen StorageBuildings, which are designated storage areas, areseparated from adjacent plant structures by 3-hourrated fire barriers such that a fire or explosion willnot adversely impact systems, equipment, orcomponents important to nuclear safety.Assessment of the Hydrogen Storage Buildingswith the applicable requirements of NFPA 50A hasbeen documented via DA-ME-2002-005. Thecredited bases for acceptance are valid and meetapplicable quality requirements.

The gas bottle racks are anchored to the concretepedestal in the air ejector area located in theTurbine Building Mezzanine, and are also locatedgreater than 5' away from the rack located in theTurbine Building Basement near the elevator. Theracks are not located in the vicinity of any safetyrelated equipment. The racks are located in a non-seismic building with no seismic category 1 or SRequipment; therefore, the rack is not required tomeet seismic criteria. Signs are also provided toensure total hydrogen content less than 400 scf.

" DA-ME-2002-005,rev. 0,Primary and SecondaryHydrogen StorageBuildings NFPA 50A CodeReview.

" A-202,rev. 03100, TheFire Protection Programand Ginna Station StaffResponsibilities for FireProtectionFPS-16,rev. 01700, BulkStorage of CombustibleMaterials and TransientFire Loads

I. .1 _____________ ± ________________________________________ a


Page 9 of 114


FPE RAI 03

The compliance statement for LAR Attachment A, Table B-i, Section 3.4.1(c) is "complies".NFPA 805, Section 3.4.1(c) specifically requires the fire brigade leader and two brigademembers to have sufficient training and knowledge of nuclear safety systems to understand theeffects of fire and fire suppressants on nuclear safety performance criteria. The compliancebasis for this element states that that the Brigade Captain and Backup Brigade Captain areAuxiliary Operators, but does not specify the details of the training and knowledge of thesemembers.

Describe how the requirements of NFPA 805 Section 3.4.1(c) are met with regard to trainingand knowledge of the brigade leader and at least two of the brigade members.

An approach acceptable to the staff for meeting this training and knowledge requirement isprovided in Regulatory Guide (RG) 1.189, "Fire Protection for Nuclear Power Plants," Rev. 2,Section 1.6.4.1, Qualifications:

"The brigade leader and at least two brigade members should have sufficient training inor knowledge of plant systems to understand the effects of fire and fire suppressants onsafe-shutdown capability. The brigade leader should be competent to assess thepotential safety consequences of a fire and advise control room personnel. Suchcompetence by the brigade leader may be evidenced by possession of an operator'slicense or equivalent knowledge of plant systems."

Response

The following is a description of the changes to LAR Attachment A, Table B-i, Section 3.4.1 (c):

Change compliance statement to: "Complies via Previous Approval".Change compliance basis to:Ginna's fire protection program is consistent with existing commitments and utilizes acompliance category of "Complies by previous NRC Approval" in accordance with NFPA 805Section 3.1. The fire brigade training is acceptable per letter From D.L. Ziemann, NRC To L.D.White, RG&E, "Fire Protection SER with Summary of MODs, Evaluation of Plant Features &Specific Plant Areas" dated 02/14/1979, Accession # 7903140202 [RG001680] Sections 3.1.31and 6.2, along with Ginna Station Fire Protection Program Report Table A-1 Appendix ASection B.5.b [EPM-FPPR].

FPE RAI 031

Page 10 of 114



3.4.1(c)Brigade Members

Training

During every shift, the brigade leader and at leasttwo brigade members shall have sufficient trainingand knowledge of nuclear safety systems tounderstand the effects of fire and fire suppressantson nuclear safety performance criteria. Exceptionto (c): Sufficient training and knowledge shall bepermitted to be provided by an operations advisordedicated to industrial fire brigade support.

(KCo~mplies

viaPreviousApproval

7.1] !rcFquiFrcs that the Fire Brigade Captain an~dBackup Fire Brigade Captain On; oach Shift b9trained plant Auxiliary OperatoFrs, have fire brigadmcmbor: qualification training, and- additionalFife

, -igad Captain training. Additionally, one (1)

additional FiPre Brigade m. mbr On each 6hift icrequired to be a trained plant Auxilia;.y Operatorand have fire brigade nmhembe qu alification tainig

Ginna's fire protection program is consistent with "existing commitments and utilizes a compliancecategory of "Complies by previous NRC Approval"in accordance with NFPA 805 Section 3.1. The firebrigade training is acceptable per NRC FP SER,2/14/79, sections 3.1.31 and 6.2 [RG001680],along with the Fire Protection Program ReportTable A-i, Appendix A section B.5.b [EPM-FPPRJ.

A '2A') Mcý, 32 A 151 a~.- the. Fi~ F2Et~l * EPM-FPPR ANL,rev. 8.0,Fire Protection ProgramReport (FPPR)

*A 202,rev. 03100, The-Fire Protection Program

and GiOnn Statil ta#Respensibilitfiec f(ar Fire

R RG0016 K , Safety 'Evaluation Report DocketNo. 50-244, 2/14/1979. )

± _________________________________________________ I ________________ L_________________________________________________


Page 11 of 114


FPE RAI 04

The compliance basis for LAR Attachment A, Table B-i, Section 3.11.5 states that Hemyc wrapis installed in Battery Room lB. LAR Attachment S, Table S-2, Item 7 is an implementation itemto evaluate the existing configuration of the wrap to determine if it is adequate to protect certaincables for 45 minutes and to modify the wrap to ensure 45 minutes is achieved. NFPA 805Section 3.11.5 states that "electrical raceway fire barrier systems (ERFBS) required by chapter4 shall be capable of resisting the fire effects of hazards in the area", and "ERFBS shall betested in accordance with and shall meet the acceptance criteria of NRC Generic Letter 86-10,Supplement 1." Provide additional detail regarding the capability of this ERFBS to meet therequirements of NFPA 805 Section 3.11.5. Include in the response a discussion of how theERFBS duration is based on fire testing of similar material and application.

Response

An analysis has been performed to address the commitment of LAR Attachment S, Table S-2,Item 7. The analysis (Ref. 0028-0018-000-001) compares the Ginna Hemyc wrap configurationin Battery Room B to the following test configurations and is representative of the wrapconfiguration in the field (air drops, terminations, support / interference protection, collars, etc.):

Omega Point Laboratories, Inc., Project Number 14790-123264, Date: April 18, 2005, Titled:HEMYC (1-Hour) Electrical Raceway Fire Barrier Systems Performance Testing: Cable Tray,Cable Air Drop and Junction Box Raceways.

Intertek Testing Services NA, Inc., Report Number 3106846, Revision Date: February 5, 2007,Titled: HEMYC 1-Hour Electrical Raceway Fire Barrier System (ERFBS), Fire ResistancePerformance

Furnace temperatures used in the testing follow the ASTM E-1 19 time-temperature curve. Thespecific combustible loading in BRIB is not an input in the analysis (Ref. 0028-0018-000-001).

The results of the evaluation (Ref. 0028-0018-000-001) indicate that the Hemyc wrapconfiguration will be able to provide 25 minutes of protection after the damage temperature (2050C) of the thermoplastic cables is reached. This is provided that the Unistrut supports insidethe steel cable chase are stuffed with ceramic fiber material to ensure a path for combustionproducts does not exist. 205 °C is the damage temperature of the thermoplastic cablesaccording to NUREG/CR-6850 and as referenced in the Fire PRA Notebook G1-FSS-F001.

The installation of the needed ceramic fiber will be tracked via ESR-12-0142.

The fire PRA model will credit an additional 25 minutes of protection beyond the point wherecables L0318 and C0687 in the back of Battery Room 1B would normally be damaged. This willbe reflected in the overall results and the delta risk calculations which will be part of ourAttachment W update. The Attachment W update will be transmitted as part of our to RAI PRA44 response. The Hemyc protection is not being credited to deterministically resolve any

FPE RAI 041

Page 12 of 114


VFDRs. The combustible loading in BR1B will be kept as low as reasonably achievable, suchthat the 1-hour structural steel fire proofing would not be compromised, as discussed in the B-1Table section 3.11.2.

FPE RAI 042

Page 13 of 114



3.11.5ERFBS

"ELECTRICAL RACEWAY FIRE BARRIERSYSTEMS (ERFBS). ERFBS required by Chapter4 shall be capable of resisting the fire effects of thehazards in the area. ERFBS shall be tested inaccordance with and shall meet the acceptancecriteria of NRC Generic Letter 86-10, Supplement1, "Fire Endurance Test Acceptance Criteria forFire Barrier Systems Used to Separate SafeShutdown Trains Within the Same Fire Area." TheERFBS needs to adequately address the designrequirements and limitations of supports andintervening items and their impact on the firebarrier system rating. The fire barrier system'sability to maintain the required nuclear safetycircuits free of fire damage for a specific thermalexposure, barrier design, raceway size and type,cable size, fill, and type shall be demonstrated.Exception No. 1: When the temperatures inside thefire barrier system exceed the maximumtemperature allowed by the acceptance criteria ofGeneric Letter 86-10, "Fire Endurance AcceptanceTest Criteria for Fire Barrier Systems Used toSeparate Redundant Safe Shutdown TrainingWithin the Same Fire Area," Supplement 1,functionality of the cable at these elevatedtemperatures shall be demonstrated. Qualificationdemonstration of these cables shall be performedin accordance with the electrical testingrequirements of Generic Letter 86-1 0, Supplement1, Attachment 1, "Attachment Methods forDemonstrating Functionality of Cables Protectedby Raceway Fire Barrier Systems During and AfterFire Endurance Test Exposure." Exception No. 2:ERFBS systems employed prior to the issuance ofGeneric Letter 86-10, Supplement I, are acceptableproviding that the system successfully met thelimiting endpoint temperature requirements asspecified by the AHJ at the time of acceptance."

Complies Ginna is not crediting the existing HEMYC wrapconfigurations within the plant as a fire ratedbarrier, with the exception of fire area BR1 B(Battery Room 1 B) as described below:

Engineering Service Request, ESR-12-0142, aslisted in Attachment S-2, item 7, of the LAR, willdetermine if the existing configuration of the Hemycwrap (HWCB03) in the Battery Room 1B isadequate to protect L0318, C0687, and a portionof E0053., for at least 4,5 minutcs. If the aRnalyicchowcV this- ;.6 not pG6cciblo, a mo`1dificatiOn Will 139

pei~emid Flald to the HamBYc to encuro 45

An analysis has been performed to address thecommitment of LAR Attachment S, Table S-2, Item7. The analysis (Ref. 0028-0018-000-001)compares the Ginna Hemyc wrap configuration in (Battery Room B to the following test configurationsand is representative of the wrap configuration inthe field (air drops, terminations, support /interference protection, collars, etc.):

(

P~AA

" A-52.12,rev. 06801,Nonfunctional EquipmentImportant to Safety

" A-202,rev. 03100, TheFire Protection Programand Ginna Station StaffResponsibilities for FireProtection

" ESR-12-0142, NFPA 805CDF Reduction Mod-014

HWC£3 iuae•-t4-

pratect L031 8, 00687,and Part of E0053-in,-th4eBatteiry Roam B for 45FR;;Ut6 gi~enBatteFyRoom B hac a major fire

rbcuh that 15 minu-to isdefe~dbe

,. c QW 2_ it( In 7* 0028-0018-000-001,

Qualification of HEMYCFire Barrier Wrap inBattery Room B of Ginna

LNuclear Station, Revision02.

Omega Point Laboratories, Inc., Project Number14790-123264, Date: April 18, 2005, Titled:HEMYC (1-Hour) Electrical Raceway Fire BarrierSystems Performance Testing: Cable Tray, CableAir Drop and Junction Box Raceways

Intertek Testing Services NA, Inc., Report Number3106846, Revision Date: February 5, 2007, Titled:HEMYC 1-Hour Electrical Raceway Fire BarrierSystem (ERFBS), Fire Resistance Performance

Furnace temperatures used in the testing follow theASTM E-1 19 time-temperature curve. The specificcombustible loading in BRIB is not an input in theanalysis (Ref. 0028-0018-000-001).

II

The results of the evaluation (Ref. 0028-0018-000- V



Page 14 of 114


NFPA 805 Requirements/GuidanceChapter 3Reference

3.11.5ERFBS

(Continued)


001) indicate that the Hemyc wrap configuration \will be able to provide 25 minutes of protection ¼after the damage temperature (205 °C) of thethermoplastic cables is reached. This is providedthat the Unistrut supports inside the steel cablechase are stuffed with ceramic fiber material toensure a path for combustion products does notexist. 205 0C is the damage temperature of thethermoplastic cables according to NUREG/CR-6850 and as referenced in the Fire PRA NotebookG1 -FSS-FOO1. The credited bases for acceptanceare valid and meet applicable quality requirements.

The installation of the needed ceramic fiber will betracked via ESR-12-0142.


Page 15 of 114


FPE RAI 05

LAR Attachment L, Approval Request 1, is for approval of the use of two portable gas fireheaters in the Screen House to keep the traveling screens from freezing during cold weather.The basis for the request states that the closest equipment credited for the Nuclear SafetyCapability Assessment (NSCA) in the Screen House, the motor fire pump and service waterpumps, are at a minimum 40 feet from the location of the portable gas heaters, which placesthem outside the zone of influence (ZOI) of the heaters. Describe the administrative controlsthat ensure the portable heaters are located at a minimum of 40 feet from the NSCA equipment,each time they are installed. In addition, describe how the administrative controls ofcombustibles account for the heaters when installed, including any changes to bulk storage ortransient fire load locations.

Response

The following is a description of the changes to LAR Attachment L, Approval Request 1 from therequirements of NFPA 805 Section 3.3.1.3.4:

Add this paragraph to the Basis for Request:

"The administrative controls that install the portable heaters are M- 115 (Installation andRemoval of Screen House Gas Space Heaters) and T-35P (Screen House Supplemental Heat,Placing in Service, and Removal from Service). The gas line for the gas fired heaters is hardpiped and therefore, the location is fixed, and maintained at a distance which is at a minimum of40 feet."

The administrative controls of combustibles is controlled by the use of FPS-16 (Bulk Storage ofCombustible Materials and Transient Fire Loads), EP-3P-0132 (Combustible loadingworksheet), and A-905 (Open Flame, Welding, and Grinding Permit) as described in theCombustible Loading discussion.

Add the following to the Combustible Loading discussion:

"There is no impact to the combustible loading analysis (ref. DA-ME-98-004) when the portablegas heaters are in service due to the hard piping of the gas supply. Transient combustibles aretracked administratively through FPS-16, Attachment 2 (Transient Combustible Permit). TheFire Marshal or designee reviews the permit for the proposed usage and/or storage oftransients, in regards to location, and includes potential restrictions to be followed as necessary.The Temporary Combustible Permit is periodically reviewed to ensure restrictions are beingfollowed and documented as part of the plant inspection program performed per A-54.7, FireProtection Tour. Additionally, during the initial installation of the portable heaters, via procedureM- 115, there is a step to notify the Fire Marshal prior to performing the procedure. This willensure the potential for a transient already in place in the Screen House to be evaluated forrelocation if it is in the proximity of the two portable gas fired heaters."

FPE RAI 051

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FPE RAI 06

LAR Attachment L, Approval Request 2, is for approval of wiring above suspended ceilings thatmay not comply with the requirements of NFPA 805, Section 3.3.5.1. The basis for the requestindicates that there are multiple areas where suspended ceilings are located, however only theControl Room (CR) was discussed in detail. Provide a description of the other fire areas thatcontain wiring above suspended ceilings, including proximity to fire areas containing nuclearsafety capability systems and equipment. Also, include a discussion on the type, use, andamount of wiring, proximity to combustibles, presence of ignition sources, as well as any firedetection or suppression features that may be installed.

Response

The Turbine Building Operating Level Conference Room and the MUX Room are incorrectlylisted as rooms with wiring above suspended ceilings. These rooms are closed, self containedunits (in essence "boxes") within larger rooms and won't be considered in this evaluation. LARAttachment A, Section 3.3.5.1 will be revised to delete discussion of these rooms from theCompliance Basis. The TSC (Technical Support Center) and adjoining hallway, along with theService Building Basement office areas are discussed in greater detail as shown below in therevised Approval Request 2:

Approval Request 2

NFPA 805, Section 3.3.5.1 states:"Wiring above suspended ceiling shall be kept to a minimum. Where installed, electrical wiringshall be listed for plenum use, routed in armored cable, routed in metallic conduit, or routed incable trays with solid metal top and bottom covers."NRC approval is requested for a Performance-Based (PB) method to demonstrate an equivalentlevel of fire protection for the requirement of NFPA 805, section 3.3.5.1 regarding wiring abovesuspended ceilings.

Basis for Request

Ginna has wiring above suspended ceilings that may not comply with the requirements of thiscode section. Suspended ceilings were identified in the following areas:

" Control Room* TSC (Technical Support Center) and adjoining hallway* Service Building Basement office areas

These areas are not risk significant with the exception of the Control Room. A walk down of theareas revealed that the majority of the wiring is communication systems or lighting systemsrelated.

Control Room

The FIN (Fix it Now) team determined there are no active power cables above the suspendedceiling of the Control Room. One cable was abandoned in place with no power to it. Overall the

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cabling in the Control Room has a very low possibility of a fire due to limited combustibleloading, discontinuity of combustibles, and the inherent features of the electrical circuit design.The Control Room HVAC system design and operation supports the rapid identification ofcombustion products if a fire were to occur in the MCR suspended ceiling. The normal ControlRoom HVAC system is located in the basement floor of the three-story Control Building and isconnected to the Control Room Envelope (CRE) by supply and return ducts that are located inthe stairwell. The normal HVAC system supports the NORMAL and PURGE modes ofoperation. The system includes a supply and return air fan and in the NORMAL mode providesfresh outside air and exhaust, coarse filtration, and heating or cooling via electric heating orchilled water cooling coils. The NORMAL system includes a separate fan for lavatory exhaust,which is isolated in the EMERGENCY mode of operation. In the PURGE mode of operation theNORMAL system has the same functions described above while also providing the maximumamount of fresh air and exhaust air to purge airborne contaminants from the CRE.

The normal HVAC system's outside air intake duct is equipped with redundant trains ofradiation, chlorine, and ammonia monitors. Any one of these six monitors reaching theirsetpoint will actuate the EMERGENCY mode of operation, employing the Control RoomEmergency Air Treatment System (CREATS), along with providing an alarm in the ControlRoom. The normal HVAC system is also equipped with a smoke detector that monitors returnair in the duct between the CRE and the return air fan, and provides an alarm in the ControlRoom. CREATS simply isolates and recirculates air within the CRE boundary in a closed-loopsystem. The existing fire detection system or the Control Room operators who are continuouslypresent in the area would quickly identify the presence of smoke while the air is beingrecirculated.

The observed video/communication/data cables above this suspended ceiling are approximatedto be less than 5% of the total space. Video/communication/data cables are low voltage. Theselow voltage cables are not generally susceptible to shorts which would result in a fire. TheControl Room does contain NSCA components/equipment. Portable fire extinguishers areavailable in this zone. Hose reels are available for use on the Turbine Building Operating floor,if required. Smoke and heat detectors provide area detection for the Control Room, althoughthey are not located above the suspended ceiling.

Technical Support Center (TSC) and adioining hallway

A walk down determined that the quantity of wiring which may not be plenum rated or routed ina metallic conduit, above these ceilings is minor. Approximating, less than 1% of the spaceabove the suspended ceilings in the TSC and adjoining hallways is occupied by electrical wiringthat may not be listed for plenum use, routed in armored cable, or routed in metallic conduit.The majority of this wiring is low voltage communications wiring or lighting systems related.There are no intervening combustibles in the space above the suspended ceilings since the vastmajority of equipment is metal, and all other wiring is routed in conduits.

The TSC hallway is located between the TSC and the Turbine Building Mezzanine and otherrooms including the TSC diesel generator room, TSC inverter room, and TSC battery room. 3hr fire door F7 (TSC south hallway door) is located between the TSC south hallway and theTurbine Building Mezzanine. 3 hr fire door F8 is located between the TSC south hallway to theTSC battery room, 3 hr fire door F9 is located between the TSC south hallway to the TSCinverter room, 3 hr fire door F10 is located between the TSC south hallway to the TSC dieselgenerator room, 3 hr fire door Fl 1 is located between the TSC south hallway to exterior, 3 hr

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rated fire door F12 between the TSC south hallway to the TSC office area, and a 3 hr rated firedoor F14 (TSC north hallway door) between the TSC north hallway. [References EPM-FPPR,33013-2617,1 and 21488-0119,3]. There also are 3 hr rated walls between the TSC north andsouth hallway and the Turbine Building Mezzanine, along with a 3 hr rated fire wall between theSafety Assessment System (SAS)/Power Plant Computer System (PPCS) computer room andthe Turbine Building Mezzanine [EPM-FPPR and penetration database].

There were no ignition sources observed during the walk downs and the proximity of the wiringto areas containing NSCA equipment is not an issue since there are 3 hr rated walls andpenetrations between these areas. There is also detection located in the hallways, and roomswithin the TSC along with S30 automatic sprinkler system which provides suppression coveragefor the TSC Diesel Generator Room and Operational Support Center. S27 is an automatichalon suppression system which provides suppression for the SAS/PPCS computer room andsubfloor. Portable fire extinguishers are available in this zone and in adjacent fire areas/zonesalong with hose stations.

Service Building Basement office areas

A walk down of the electric shop and mechanical maintenance administrative areas (fire zoneSB-1) was conducted due their proximity to the Primary Water Treatment room (fire zone SB-1WT) which contains NSCA equipment including the condensate storage (CST) tanks and CSTlocal level indicator. There was only (1) de-energized power cables observed above the electricshop ceiling. The quantity of electrical wiring that may not be listed for plenum use, routed inarmored cable, or routed in metallic conduit above the suspended ceilings above the electricshop and maintenance administrative areas was observed to be minor (less than 1% of space)and classified as video/communication/data cables. There are no NCSA components located infire zone SB-I. There are no ignition sources above these ceilings. Although there is a non-rated wall between the Primary Water Treatment room and the Service Building Basement, S19is an automatic sprinkler system that provides suppression protection in fire zone SB-1 and SB-1WT. Portable fire extinguishers are available in this zone, and there are hose stationsavailable in adjacent fire areas/zones as well as yard hydrant connections.

The basis for the approval request of this difference from NFPA 805, Chapter 3 requirements is:

• There are no ignition sources above these ceilings" The wiring above ceilings in offices, conference rooms, laboratories, lobbies, etc. do not

pose a hazard:" Low voltage is not susceptible to shorts causing a fire" There is a lack of continuity of combustibles" There is no equipment important to nuclear safety in the vicinity of these cables" Modification Design Process requires new installations to use plenum-rated equivalentor armored cable (Note: FAQ 06-0022 identified acceptable electrical cable flamepropagation tests). Plenum-rated cable is tested to NFPA-262.

* Power, control or instrumentation cables installed are either IEEE-383 qualified (orequivalent) or provided with a flame retardant coating.

Acceptance Criteria EvaluationNuclear Safety and Radiological Release Performance Criteria:

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The location of wiring above suspended ceilings does not affect nuclear safety since: (1) thespace enclosing these cables are non-combustible, (2) the location of wiring above thesuspended ceilings has a minimum amount of nearby ignition sources considering the adjacentpower, control, or instrument cables, and (3) the video/communication/data cables are lowenergy and therefore pose a low fire ignition hazard due to hot shorts. Therefore, there is noimpact on the nuclear safety performance criteria.

The wiring above the suspended ceilings has no impact on the radiological release performancecriteria. The radiological review was performed based on the potential location of radiologicalconcerns and is not dependent on the type of cables or locations of suspended ceilings.

Safety Margin and Defense-in-Depth:

The amount of non-rated and non-enclosed wiring above the ceilings in the Power Block isminor and does not present a significant fire hazard. Therefore, the safety margin inherent inthe analysis for the fire event has been preserved.

The three echelons of defense-in-depth are 1) prevent fires from occurring (hot work and otheradministrative control), 2) rapidly detect, control and extinguish fires that occur thereby limitingdamage (fire detection systems, automatic fire suppression systems, manual fire suppressionand pre-fire plans to aid the fire brigade), and 3) provide an adequate level of fire protection forsystems and structures, so that a fire will not prevent essential safety functions from beingperformed (fire barriers, fire rated cable, success path remains free of fire damage, recoveryactions). The prior introduction of non-listed video/communications/data cables routed abovesuspended ceilings does not directly result in compromising automatic fire suppression systems,manual fire suppression functions, or post-fire safe shutdown capability.

The inherent safety margin remains unchanged. The introduction of non-listed video /communications / data cables routed above suspended ceilings does not impact fire protectiondefense-in-depth. The video/communication/data cables routed above suspended ceilings doesnot directly result in compromising automatic fire suppression or detection functions, manual firesuppression function, or post-fire safe shutdown capability.

Conclusion:

NRC approval is requested for the acceptance of a Performance-Based (PB) method todemonstrate an equivalent level of fire protection for the requirement of NFPA 805, Section3.3.5.1 regarding the currently installed non-plenum listed cables routed above suspendedceilings.

Ginna determined that the performance-based approach/NFPA 805 alternative satisfies thefollowing criteria:

(A) Satisfies the performance goals, performance objectives, and performance criteriaspecified in NFPA 805 related to nuclear safety and radiological release;

(B) Maintains safety margins; and(C) Maintains fire protection defense-in-depth (fire prevention, fire detection, fire

suppression, mitigation, and post-fire safe shutdown capability.

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3.3.5.1Suspended

Wiring

Wiring above suspended ceiling shall be kept to aminimum. Where installed, electrical wiring shall belisted for plenum use, routed in armored cable,routed in metallic conduit, or routed in cable trayswith solid metal top and bottom covers.

Submit forNRC Approval

Ginna has wiring above suspended ceilings thatmay not comply with the requirements of this codesection. Suspended ceilings were identified in thefollowing areas:

* Turbino BuRpildring Oporating L9VOI ConferenceRnnm

* MUX Room (romputor room)TSC (Technical Support Center) and adjoininghallway

* Service Building Basement office areas* Control Room

See Attachment L of the Transition report forfurther details on the request for NRC approval forexisting wiring above suspended ceilings.

The wiring above suspended ceilings for futureinstallations is controlled administratively withinCNG-FES-007 [Attach. 3].

CNG-FES-007 rev.00015, Preparation ofDesign Inputs and ChangeImpact Screen

I I

Ginna LAR Rev 0 Page A-IS


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FPE RAI 07

LAR Attachment L, Approval Request 3, is for approval of the use of wiring that may not complywith the requirements of NFPA 805, Section 3.3.5.3. The Approval Request states that theperformance based (PB) method is specifically associated with the installation ofvideo/communication/data cables. However, the basis for the request includes a discussion ofpower cables used for modifications such as the spent fuel pool bridge crane which wasmodified using cables others than those meeting the Institute of Electrical and ElectronicsEngineers (IEEE) Standard 383, "IEEE Standard for Qualifying Class 1E Electric Cables andField Splices for Nuclear Power Generating Stations." It is unclear whether this request is for alltypes of cable or only video/communication/data cables. Clarify the scope of the request. Inaddition, provide further justification for the acceptability of using cable that does not complywith NFPA 805, Section 3.3.5.3. Include a qualitative or quantitative risk evaluation, as well as amore detailed discussion of how safety margin (SM) and defense-in-depth (DID) are maintainedusing this PB approach.

Response

Change compliance statement in the B-1 table section 3.3.5.3 from "Submit for NRC Approval"to "Complies via Previous Approval" and "Submit for NRC Approval".

The "Complies via Previous Approval" is applicable to all non video/communication/data cables,and the "Submit for NRC Approval" is applicable to the video/communication/data cables.Procedure EP-3-P-0504, "Electrical/I&C Analyses Impact Form and Load Growth ControlProgram", and corporate procedure CNG-FES-007, "Preparation of Design Inputs and ChangeImpact Screen," ensures that all new power, control or instrument cable installed will beconstructed to meet or exceed the requirements of:

IEEE 383-1974 orIEEE 1202-1991 orCSA (Canadian Standards Association) 22.2 No. 0.3 orNFPA 262 orUL 44, UL 83, UL 1581, UL 1666, or UL 1685

The compliance basis statement for "Complies via Previous Approval" is as follows:

The UFSAR [Sec. 9.5.1.2.4.8] states, "The cable insulation used at Ginna includes Kerite, oil-based rubber, neoprene, and polyvinyl chloride (PVC). The cables have, as a minimum, passedthe ASTM and UL horizontal and vertical flame tests. Power cables and PVC control cableshave passed the Consolidated Edison Bonfire Test. The majority of electrical cables werepurchased and installed prior to issuance of IEEE 383; however, the potential combustionproducts for the materials used at the station have been evaluated from generic test reports anddo not exhibit an unusual or significantly hazardous nature. All cables used for modificationsmeet IEEE 383 criteria unless specifically excepted. "A specific determination is made wheneverit is impracticable to meet IEEE 383 criteria for cables used in modifications. RGO03006 [Encl.1, Item 3.2.4, pg. 2] states, "SER Section 3.2.4 indicates that the licensee is investigating thefire characteristics, including fire resistance, of the cable insulation used in the plant. By letterdated April 30, 1979, the licensee provided a list of cable insulation types and quantities used inthe plant. The assumptions on Page I-1 of the licensee's study performed in response to SER

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Section 3.2. 1 obviate the need for a separate staff analysis of the fire characteristics of electricalcable insulation. We conclude that this item is acceptable."

Electric cable construction is acceptable per NRC SER Supplement No. 2 Fire ProtectionAccession # 8102270116, dated 02/06/1981, Section 3.2.4, [RG003006] and EPM-FPPRAppendix A section D.3.f.

The compliance basis statement for "Submit for NRC Approval" is as follows:

The approval request is for the video/communication/data cables that do not comply with therequirement of NFPA 805, section 3.3.5.3.

See Attachment L of the Transition report for further details on the request for NRC approval forthe video/communication/data cables that do not comply with this requirement.

Procedure EP-3-P-0504, "Electrical/I&C Analyses Impact Form and Load Growth ControlProgram", and corporate procedure CNG-FES-007, "Preparation of Design Inputs and ChangeImpact Screen," ensures that all new power, control or instrument cable installed will beconstructed to meet or exceed the requirements of:

IEEE 383-1974 orIEEE 1202-1991 orCSA (Canadian Standards Association) 22.2 No. 0.3 orNFPA 262 orUL 44, UL 83, UL 1581, UL 1666, or UL 1685

The following is a complete revision of Attachment L, Approval Request 3 from the requirements

of NFPA 805 Section 3.3.5.3:

NFPA 805, Section 3.3.5.3

NFPA 805, Section 3.3.5.3 states:

"Electric cable construction shall comply with a flame propagation test as acceptable to theAHJ. "

NRC approval is requested for the acceptance of a Risk-Informed Performance-Based (PB)method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805,Section 3.3.5.3 regarding an acceptable flame propagation test for electric cable construction.The scope of this request includes video/communication/data cables. NRC approval of a Risk-Informed PB method is needed to justify the use cable that may not comply with therequirements of this code section, which are not necessarily tested in accordance with the flamepropagation test requirements of IEEE-383, or any other qualification standard outlined in FAQ06-0022 as endorsed by the NRC.

Basis for Request

The video/communication/data cables are not necessarily tested in accordance with the flamepropagation tests outlined in the FAQ 06-0022 as endorsed by the NRC. These low voltage

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cables are not generally susceptible to shorts which would result in a fire, therefore self-ignitedfires are not a concern. An exposure fire could potentially ignite the cables, although the samefire would result in damage to other cables in the vicinity.

With the exception of the telephone communications room located in fire area BOP/ zones TB-2(Turbine Building Mezzanine), and TSC-1M (Administrative Computer Room), along with theControl Room (fire area CC/ zone CR), the remaining areas contain a limited quantity of thiswiring which is sufficiently dispersed, is considered an insignificant fire hazard, and is notcapable of causing fire damage to components necessary for safe shutdown.

The telephone communications room in fire area BOP/ zone TB-2 is protected by an automaticsprinkler system S38. Portable fire extinguisher, and hose stations are also available in this firearea. A fire in this room would be readily extinguished by the automatic actuation of sprinklersystem S38.

TSC-1M is protected by smoke detectors S34D1 and S34D2, portable fire extinguishers, andhose stations in adjacent areas. A potential fire in this area would be detected andconsequently extinguished.

BOP/CR is protected by a constantly manned area, Z19 smoke and heat detectors, portable fireextinguishers, and hose reels available in adjacent areas. A potential fire in this area would bereadily detected due to the constantly manned area, and consequently extinguished.

The combustible loading in these areas is tracked within DA-ME-98-004 and transientcombustibles are tracked administratively through FPS-16.

Acceptance Criteria EvaluationNuclear Safety and Radiological Release Performance Criteria:

Video/communication/data cables are low-voltage cable not susceptible to shorts that wouldresult in a fire. Therefore, there is no impact on the nuclear safety performance criteria.The flame propagation testing of electrical cable construction has no impact on the radiologicalrelease performance criteria. The radiological review was performed based on the potentiallocation of radiological concerns and is not dependent on the flame propagation tests of cables.

The limited use of video/communication/data cabling has been shown to be acceptable anddoes not create or pose an un-acceptable fire hazard. Therefore, the radiological releaseperformance criteria are also satisfied based on the determination of limiting radioactive release.

Safety Margin and Defense-in-Depth

The introduction of non-listed video/communication/data does not impact fire protectiondefense-in-depth. The video/communication/data cables do not directly result in compromisingautomatic fire suppression or detection functions, manual fire suppression functions, or post-firesafe shutdown capability.

Exposed video/communication/data cables, installed at Ginna, with cable construction that doesnot comply with a flame propagation test acceptable to the AHJ is not capable of causing firedamage to components necessary for safe shutdown due to the nature of the low voltage of

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these cables not being susceptible to hot shorts or being a significant fire hazard in areas otherthan the communications room, administrative computer room, and Control Room. Thesuppression system, smoke detectors, portable fire extinguishers, and hose reels is consideredadequate to prevent fire propagation in these areas. The Control Room is also constantlymanned. Therefore, the safety margin and defense-in-depth is maintained.

The three echelons of defense-in-depth are 1) prevent fires from starting (administrativeprocedures for combustible/hot work controls) via FPS-16, 2) rapidly detect, control andextinguish fires that do occur thereby limiting damage (fire detection systems, automatic/manualfire suppression, fire brigade/pre-fire plans), and 3) provide adequate level of fire protection forsystems and structures so that a fire will not prevent essential safety functions from beingperformed (fire barriers, success path remains free of fire damage, recovery actions).

Exposed video/communication/data cables, installed at Ginna, with cable construction that doesnot comply with a flame propagation test acceptable to the AHJ does not affect echelon 1 of thedefense-in-depth concept because cable construction is not involved with administrativeprocedures to prevent fire from occurring, and administrative procedures control and trackcombustibles at Ginna via FPS-16. In areas containing these cables which cannot becategorized as insignificant adequate detection, manual hose stream, and fire extinguishers areprovided to ensure the fire is rapidly detected and controlled/extinguished by the fire brigade.Therefore, echelon 2 of the defense-in-depth concept is maintained. The telephonecommunications room is a room within fire zone TB-2 (Turbine Building Mezzanine). Since it isprotected by an automatic sprinkler system, there is an adequate level of protection to preventthe spread of fire to systems and structures. The administrative computer room in fire zoneTSC-1M, is protected by smoke detectors and hose reels, along with 3 hour concrete block wallto adjacent fire zone TSC-1N (Technical Support Center) and TB-2 (Turbine BuildingMezzanine). The Control Room is protected by smoke and heat detectors, an automatic delugespray system which provides a water curtain on the wall between the Control room and theTurbine Building Operating Floor, along with 3 hour rated walls to the exterior YARD area.Therefore, echelon 3 of the defense-in-depth concept is maintained.

Conclusion

NRC approval is requested for the use of video/communication/data cables that are not testedto the flame propagation tests as endorsed by the NRC.

The engineering evaluation performed determined that the performance-based approachutilized to evaluate this difference from the requirements of NFPA 805, Chapter 3:

(A) Satisfies the performance goals, performance objectives, and performance criteriaspecified in NFPA 805 related to nuclear safety and radiological release;

(B) Maintains safety margins; and(C) Maintains fire protection defense-in-depth fire prevention, fire detection, fire suppression,

mitigation, and post-fire safe shutdown capability).

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3.3.5.3Electrical CableConstruction

Electric cable construction shall comply with aflame propagation test as acceptable to the AHJ. lNRG AP•p•,l

Complies viaPreviousApproval

The UFSAR [Sec. 9.5.1.2.4.8] states, "The cableinsulation used at Ginna includes Kerite, oil-basedrubber, neoprene, and polyvinyl chloride (PVC).The cables have, as a minimum, passed the ASTMand UL horizontal and vertical flame tests. Powercables and PVC control cables have passed theConsolidated Edison Bonfire Test. The majority ofelectrical cables were purchased and installed priorto issuance of IEEE 383; however, the potentialcombustion products for the materials used at thestation have been evaluated from generic testreports and do not exhibit an unusual orsignificantly hazardous nature. All cables used formodifications meet IEEE 383 criteria unlessspecifically excepted." A specific determination ismade whenever it is impracticable to meet IEEE383 criteria for cables used in modifications.RG003006 [End. 1, Item 3.2.4, pg. 2] states, "SERSection 3.2.4 indicates that the licensee isinvestigating the fire characteristics, including fireresistance, of the cable insulation used in the plant.By letter dated April 30, 1979, the licenseeprovided a list of cable insulation types andquantities used in the plant. The assumptions onPage I-1 of the licensee's study performed inresponse to SER Section 3.2.1 obviate the needfor a separate staff analysis of the firecharacteristics of electrical cable insulation. Weconclude that this item is acceptable."

Electric cable construction is acceptable per NRCSER Supplement No. 2 Fire Protection Accession#8102270116, dated 02/06/1981, Section 3.2.4[RG003006] and EPM-FPPR Appendix A sectionD.3.f.

Minor quantitioc of existingvido4,o.....• c .. At.Ond.. t;; cuabl'e ;and other coadetyP8s May not moot the requiremenRt of this code66ctiR. TherforoF, NRC approVal Of oxistin9cablec; tha4t do not mee~t thA requirement of NIFPA

" RG003006, SafetyEvaluation ReportSupplement 2, 2/6/81

" CNG-FES-007 rev.00015, Preparation ofDesign Inputs and ChangeImpact Screen

" EP-3-P-0504,rev. 01500,Electrical I and C AnalysesImpact Form

* rev 8.0,•EPM-FPPR ANL,rev.8.0,

Fire Protection ProgramReport (FPPR)

I ____________________________________________ .j ______________ I


Page 26 of 114


NFPA 805 Requirements/GuidanceChapter 3Reference


3.3.5.3Electrical CableConstruction(continued)

805, Section 3.3.5.3 and approval of apoFformanr.e baeod (121) methed to clommsnt~ratothe equiValency of future Gable typcc that may netMeet thic requirement i6 requested. SeeAttachmcnt: L to thc TraneitiOn Report.EP 3 P 0504 [SoG. 5.2] and GNG FES 00:7 [Attach.3] include accoptablo cable flame9 propagationtects, ac ic6ted in FAQ 06 0022, for Muturmed:Aiiat!GRc.NoEtei The 8XcoPtiGn to thic cection is not ondorccdby 1OCF=R5O.48 (c)(2)i(v' and has boon remoVed-.

(((

,The approval request is for thevideo/communication/data cables that do notcomply with the requirement of NFPA 805, section3.3.5.3.

See Attachment L of the Transition report forfurther details on the request for NRC approval forthe video/communication/data cables that do notcomply with this requirement.

Procedure EP-3-P-0504, "Electrical/l&C AnalysesImpact Form and Load Growth Control Program",and corporate procedure CNG-FES-007,"Preparation of Design Inputs and Change ImpactScreen," ensures that all new power, control orinstrument cable installed will be constructed tomeet or exceed the requirements of:

IEEE 383-1974 orIEEE 1202-1991 orCSA (Canadian Standards Association) 22.2 No.0.3 orNFPA 262 orUL44, UL 83, UL 1581, UL 1666, or UL 1685

K- 4d dJ

4- ______________________________________ j. ____________ L ____________________________________________________________

Ginna LAR Rev 0 Page A-lBGinna LAR Rev 0 Page A-18

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FPE RAI 08

LAR Attachment L, Approval Request 4, is for approval of the reactor coolant pump (RCP) oilmist that results from pump/motor operation. Provide the following additional information:

a. Further characterization of the misting in terms of location of deposition, and the firehazard associated with these deposition locations, including proximity to equipmentnecessary to meet nuclear safety performance criteria. Include the basis foracceptability.

b. What actions, if any, are taken to clean oil mist deposits from equipment surfaces(e.g., during maintenance outages).

Response

a. RCP oil misting has a potential to uniformly deposit on surfaces in the pump baylocated within Reactor Coolant System Loop A and Loop B. Oil has only been observedto accumulate on the equipment located within the pump bay when there was amalfunction of a component. However, a light sheen of oil is typically observed onequipment located within the pump bay due to misting, as well as on the walls of theRCP cubicle.

Regarding the proximity to equipment necessary to meet nuclear safety performancecriteria, RCP 'A' is associated with fire area/zone RC/T-LOOPA and RCP 'B' with firearea/zone RC/T-LOOPB [Ref. drawing 33013-2928,5]. The RCPs are located in thebasement of the Reactor Containment Building and extend vertically upwards as shownon drawing 33013-2131. The NSCA components located in fire zones RC-1, T-LOOPA,and T-LOOPB were evaluated for the potential of oil misting as shown in the followingtables:

NSCA equipment associated with Fire Area/Zone - RC/RC-1

EIN Description Location Room Comments123 Excess Letdown Flow 235', RC Basement TI-SE Not subject to oil misting. Concrete wall

Control Valve Regenerative HX separates RCP from component. Ref.Area 33013-2131.

200A Letdown Orifice 235', RC Basement Ti-SE Not subject to oil misting. Concrete wallRegenerative HX separates RCP from component. Ref.Area 33013-2131 and D304-0631.

200B Letdown Orifice 235', RC Basement TI-SE Not subject to oil misting. Concrete wallRegenerative HX separates RCP from component. Ref.Area 33013-2131 and D304-0631.

202 Letdown Orifice 235', RC Basement Ti-SE Not subject to oil misting. Concrete wallRegenerative HX separates RCP from component. Ref.Area 33013-2131 and D304-063 1.

294 Charging to Loop B 235', RC Basement TI-SE Not subject to oil misting. Concrete wallcold leg AOV Regenerative HX separates RCP from component. Ref.

Area 33013-2131 and D304-0631.296 Aux Spray AOV to 235', RC Basement TI-SE Not subject to oil misting. Concrete wall

pressurizer Regenerative HX separates RCP from component. Ref.Area 33013-2131 and D304-0631.

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EIN Description Location Room Comments312 Excess letdown heat 235', RC Basement TI-SE Not subject to oil misting. Concrete wall

exchanger divert to Regenerative HX separates RCP from component. Ref.VCTorRCDT Area 33013-2131.

392A Charging to Loop B hot 235', RC Basement TI-SE Not subject to oil misting. Concrete wallleg AOV Regenerative HX separates RCP from component. Ref.

Area 33013-2131 and D304-0631.386 RCP seal return bypass 235', RC Basement Ti-SE Concrete wall separates RCP from

Shield Wall B Area component. Ref. D304-0685. Confirmedwith system engineer component is notsubject to oil misting.

133 RHR to CVCS letdown 235', RC Basement TI-SE Not subject to oil misting. Concrete wallAOV Regenerative HX separates RCP from component. Ref.

Area 33013-2131 and D304-0685.701 RHR Pump suction 235', RC Basement TI- Not subject to oil misting per system

Loop A Area NE engineer. Location is between "B" RCP oiltank and "A" accumulator which is outsidethe loop. Ref. D304-0611 and D304-0612.

720 RHR Pump discharge 235', RC Basement TI-SE Not subject to oil misting. Concrete wallRegenerative HX separates RCP from component. Ref.Area D304-061 1. Confirmed with system

engineer component is not subject to oilmisting. Iocation is outside loop againstbio-shield 8' up.

852A RHR Pump discharge 235', RC Basement TI- Not subject to oil misting. Location isMOV Accumulator A NE outside the loop, adjacent to B RCP oil

Area collection tank, 4' off floor. Ref. D304-0611, D304-0612, and EP-VT-109.

852B RHR Pump discharge 235', RC Basement TI- Not subject to oil misting. Location isMOV Accumulator B SW outside the "B" loop area between the A and

Area B sumps. Ref. D304-061 I, D304-0612, andEP-VT- 109.

PT- RC Over Pressure 235', RC Basement TI- Not subject to oil misting. Location is450 Protection Transmitter Elevator Area NW across from elevator area outside loop. Ref.

EP-VT-109. Verified with system engineer.

PT- RC Over Pressure 235', RC Basement TI- Not subject to oil misting. Location is451 Protection Transmitter Elevator Area NW across from elevator area outside loop. Ref.

EP-VT-109. Verified with system engineer.PT- RC Over Pressure 235', RC Basement Ti-SE Not subject to oil misting. Location is by452 Protection Transmitter shield wall B area by east stairway. Outside

loop. Ref EP-VT-109. Verified withsystem engineer.

835A Accumulator A Fill 235', RC Basement TI- Location is directly off the 'A'Accumulator A NE Accumulator. Ref EP-VT-109 and D304-Area 0645. Not subject to oil misting per system

engineer.835B Accumulator B Fill 235', RC Basement TI- Location is directly off the 'B'

Accumulator B SW Accumulator. Ref. EP-VT-109 and D304-Area 0645. Nol subject to oil misting per system

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EIN Description Location Room Commentsengineer.

841 Accumulator A to 235', RC Basement Ti- Location is near the 'A' Accumulator. Ref.LOOP B MOV Accumulator A NE EP-VT-109, D304-0642, and D304-0643.

Area Not subject to oil misting per systemengineer.

865 Accumulator B to 235', RC Basement TI- Location is near the 'B' Accumulator. Ref.LOOP A MOV Accumulator B SW EP-VT-109, D304-0642, and D304-0643.

Area Not subject to oil misting per systemengineer.

878A SI Pump A discharge to 235', RC Basement TI-SE Concrete wall separates RCP fromB Hot Leg Regenerative HX component. Ref. 33013-213 1. Not subject

Area to oil misting per system engineer.878B S1 Pump A discharge to 235', RC Basement T1-SE Concrete wall separates RCP from

Loop B Cold Leg Regenerative HX component. Ref. 33013-2131. Not subjectArea to oil misting per system engineer.

878C SI Pump B discharge to 235', RC Basement Tl- Location is outside the 'B' loop areaLoop A Hot Leg Accumulator B SW between the 'A' and 'B' Sumps. Ref. EP-

Area VT-109. Not subject to oil misting persystem engineer.

878D SI Pump B to Loop A 235', RC Basement TI- Location is outside the 'B' loop areaCold Leg Accumulator B SW between the 'A' and 'B' Sumps. Ref. EP-

Area VT-109. Not subject to oil misting per_ system engineer.

NSCA equipment associated with Fire Area/Zone - RC/T-LOOPA

EIN Description Location Room CommentsEMS Steam Generator A 278', RC T3- Could be susceptible to oil misting, however a01A Oper Fl LOOPA buildup of oil on this component has not been

observed. Approx. 10' apart from RCP. Ref.D304-061 I.

PRC Reactor Coolant Pump 253', RC T2- Could be susceptible to oil misting, however a01A A Mezz LOOPA buildup of oil on this component has not been

observed. Ref. D304-0611, 33013-2101,33013-2131, 33013-2928,5.

700 RHR Pump Suction 235', RC Tl- Located 'A' Loop basement to the left of thefrom RCS Basement LOOPA 'A' Loop entrance. Ref. EP-VT-109.

Loop A Approximately 19' from RCP (Ref. D304-061 1area and D304-0612). Due to distance from RCP,

not susceptible to oil misting, and oil film hasnot been observed on this valve during outagewalk down per system engineer.

310 Excess Letdown AOV 253', RC T2- Approximately 6' from RCP 'A', on RCPMezz LOOPA platform. Ref. D304-063 1. Could be

susceptible to oil misting, however a buildup ofoil on this component has not been observed.

270A RCP Seal Outlet AOV 253', RC T2- Location is at the RCP 'A' (Ref. D304-0630).

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EIN Description Location Room CommentsMezz LOOPA Could be susceptible to oil misting, however a

buildup of oil on this component has not beenobserved.

NSCA equipment associated with Fire Area/Zone - RC/T-LOOPB

EIN Description Location Room CommentsEMS01 Steam Generator B 278', RC Oper Fl T3-LOOPB Could be susceptible to oil misting,B however a buildup of oil on this

component has not been observed.Approx. 10' apart from RCP. Ref.D304-061 1.

PRC01B Reactor Coolant 253', RC Mezz T2-LOOPB Could be susceptible to oil misting,Pump B however a buildup of oil on this

component has not been observed.Ref. D304-0611, 33013-2101, 33013-2131, 33013-2928,5.

721 RHR Discharge to 235', RC Basement T1-LOOPB Approximately 8' from RCP (Ref.Loop B Loop B area D304-0611 and D304-0612). Could be

susceptible to oil misting, however abuildup of oil on this component hasnot been observed.

427 Letdown Isolation 253', RC Mezz, RC T2-LOOPB Approximately 3' from RCP (Ref.AOV Coolant Pump 'B' D304-063 1). Could be susceptible to

Platform oil misting, however a buildup of oilon this component has not beenobserved.

955 Loop B Hot Leg 235', RC Basement TI-LOOPB Approximately 22' from RCP (Ref.Sample Isol Vlv Loop B area D305-0601). Due to distance from

RCP, not susceptible to oil misting,and oil film has not been observed onthis valve during outage walk downper system engineer.

270B RCP Seal Outlet 253', RC Mezz, RC T2-LOOPB Location is at the RCP 'B' (Ref. D304-AOV Coolant Pump 'B' 0630). Could be susceptible to oil

Platform misting, however a buildup of oil onI_ this component has not been observed.

The NSCA equipment* located in fire zone RC-1 was determined not to be subject to oilmisting from the RCPs due to the geometry of the reactor containment building. TheNSCA equipment located in fire zones T-LOOPA and T-LOOPB that are in the vicinity ofthe RCPs have a potential to be impacted by oil misting; however, there is no physicalevidence of oil accumulation on the equipment as verified during system engineeringoutage walk downs. Transient combustibles are tracked per FPS-16 and goodhousekeeping practices are followed at Ginna.

TE-2007-0042 and ECP-10-000066 installed a new labyrinth seal associated with theRCPs which reduced the loss of oil significantly per refueling cycle. RCP oil loss istracked by the system engineer.

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Based on the design of the RCPs, geometry of the reactor containment, the controls ofcombustibles, and the ability to track the loss of oil, the ability of the NSCA componentsto perform their intended function when needed is ensured.

* The NSCA equipment list that is evaluated in the above table is based on the NSCA

evaluation documented in EIR 51-9089546-001 revision 001A. Subsequent revisions to51-9089546-001 will not invalidate the overall conclusions of this approval request. Anyadditional NSCA equipment that may become a part of the NSCA equipment list thatmay be located in fire zone RC-1 will also not be subjected to accumulation from oil mistfrom the RCPs due to the geometry of the reactor containment building. Any additionalNSCA equipment that may become a part of the NSCA equipment list located in firezones T-LOOPA and T-LOOPB will be in the vicinity of the RCPs. However, based onoperating experience, the loss of oil from the Reactor Coolant Pumps is minor, tracked,and controlled. Oil misting has not been observed to accumulate on components withinthe Loops, absent a component failure.

b. Actions are taken to clean up any oil spills from leakage or misting during refuelingmaintenance outages. Oil deposits are tracked using the condition report process.

Again, it should be noted that oil loss has been significantly reduced with motorrefurbishments that added a labyrinth seal at the lower oil pot standpipe. Thismodification has resulted in a reduction in the oil lost from the lower oil pot through a fuelcycle. Actions are taken each refueling outage to wipe down any oil that may bedeposited on any surfaces in the pump bay.

The above information will be reflected in the revision to Attachment L of the LAR alongwith the addition of reference to EIR 51-9064339-003.

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PRA RAI 01

LAR page W-3 states that the variant condition includes all modifications for NFPA 805transition, which include proposed modifications that 1) deterministically resolve variances fromdeterministic requirements (VFDRs), or at least reduce their delta risk, or 2) reduce plant riskbut are not directly related to any particular VFDR. Clarify if "deterministically resolve VFDRs" isthe same as "make the plant compliant." Provide a table with the following information: a)modifications which make the VFDR compliant, b) modifications which help to reduce delta riskbut do not make the VFDR compliant, and c) modifications which reduce plant risk but are notdirectly related to any particular VFDR. For parts a) and b), indicate the VFDRs for which it iscredited and provide the technical basis, including an explanation on how the modificationresolves, or helps to resolve the VFDR, and any relevant supporting references (e.g. design,success criteria, calculations, etc.)

Response

The wording "deterministically resolve VFDRs" is the same as "make the plant compliant".

The table below lists, for each modification, the VFDRs that are deterministically resolved andthose VFDRs for which the modification provides a reduction in risk, along with a technicalbasis.

Delta risks are calculated as the difference between the risk in the post-transition plant, whichincorporates the planned modifications, and the risk of the deterministically compliant plant,which is defined as the post-transition plant where the VFDRs are deterministically resolved.Thus, the compliant plant incorporates the planned modifications. As a consequence, thereduction in risk from plant modifications is, above all, a reduction in plant risk rather than areduction in delta risk. There is a reduction in delta risk if the incorporated modification and/orinitial plant design actively mitigate the VFDR of concern.

PRA RAI 011

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VFDRDeterministically VFDR Whose Risk the

Modification Resolved Modification Helps Reduce Technical BasisESR-12-0123 VFDR-CC-011 N/A A MSO concern exists related to CST

Diversion to the Condenser. Firedamage to cable G1488 couldspuriously open main condenser hotwellmakeup AOV-4315 which would resultin premature draining of the CST and

damage the in-service AFW Pump(s)due to loss of suction. Control for thisAOV is not provided at a PCS. Themodification installs bi-stables such thatthe failure of a signal outside of thecontrol range will de-energize SVassociated with the dump AOV. Thisprevents the drain down and thusdeterministically resolves VFDR-CC-011.

ESR-12-0125 N/A VFDR-ABI-034 and VFDR-ABI- Modification ESR-12-0125 provides035 pressurizer pressure indication in the

control room, from two channelsprotected from fire damage in theControl Building and the Cable Tunnel.Because the cables run through theIntermediate Building, which Fire AreaABI includes, the modification cannot becredited to deterministically resolveVFDRs in that fire area, but it can becredited to help reduce the riskassociated with loss of pressurizerindication, by providing an additionaland reliable indication source.The pressurizer pressure indicationprovided by Modification ESR-12-0125

is also used to support ModificationsESR-12-0146, ESR-13-0028, ESR-13-0029, and ESR-13-0030. Thecorresponding impacted VFDRs areidentified under each of thesemodifications.

PRA RAI 012

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Modification Resolved Modification Helps Reduce Technical BasisESR-12-0126 N/A N/A Modification ESR-12-0126 provides

steam generator level indication in thecontrol room, from two channelsprotected from fire damage in theControl Building and the Cable Tunnel.While this modification was not found tohelp resolve any specific VFDR, it helpsreduce the risk associated with loss ofsteam generator level indication, byproviding an additional and reliable

indication source.The steam generator level indicationprovided by Modification ESR-12-0126

is also used to support ModificationESR-12-0128. The correspondingimpacted VFDRs are identified underthat modification.

ESR-12-0128 VFDR-BR1A-003, VFDR-ABI-012 Modification ESR-12-0128 providesVFDR-CC-017, and automatic closure of the main steamVFDR-CT-020. isolation valves (MSIVs), using the 2 out

of 2 low steam generator level protectedchannels of Modification ESR-12-0126.This modification was primarilydesigned to provide protection againstfire-induced damage in the controlbuilding and the cable tunnel (FireAreas CC, BRIA, BR1B, and CT), but isalso effective in other locations of theplant, provided that there is at least onefire scenario where the cablesassociated with the modification are notimpacted by the fire, which is the case,for example, in the auxiliary buildingportion of Fire Area ABI. Themodification can only be credited in thatpart of fire area ABI where the cablesare not located.

PRA RAI 013

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Modification Resolved Modification Helps Reduce Technical BasisESR-12-0129 N/A VFDR-ABBM-013, VFDR-ABBM- Modification ESR-12-0129 installs

019, VFDR-ABBM-023, VFDR- disconnect switches in the Main ControlABI-008, VFDR-CC-003, VFDR- Room (on the North side of the wall),CC-006, VFDR-CHG-010, which remove power to the positive andVFDR-CT-003, VFDR-CT-050, negative side of all conductors in anyVFDR-RC-003, VFDR-RC-007, cables in the PORV control circuit thatand VFDR-RC-015. could cause the PORV to spurious

open. The disconnect switches similarlyprevents spurious opening of the orificevalves, excess letdown valve 310, andletdown valve 371. Because properpolarity hot shorts might still defeat themitigative effects of the modification,there is no deterministic resolution ofVFDRs. However, the modificationhelps mitigate the risk associated withfailing to isolate PORVs, orifice valves,normal letdown, or excess letdown, forwhich several VFDRs have beenidentified.

ESR-12-0141 N/A VFDR-BOP-003, VFDR-BOP- Modification ESR-12-0141 provides013, VFDR-BR1A-009, and overcurrent protection for bothVFDR-CC-026. emergency diesel generators (EDGs) in

case of a fire outside the dieselgenerator rooms. Overcurrent failure ofthe EDGs was identified as challengingthe vital auxiliary nuclear safetyperformance criterion in several VFDRs.Accordingly, the modification eliminatesthe fire-induced risk associated with thistype of failure. Once the EDGs trip onovercurrent, the EDGs must be locallyshed, started, and loaded. As this isoutside the control room, it does notprovide a deterministic resolution of theVFDRs.

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Modification Resolved Modification Helps Reduce Technical BasisESR-12-0142 N/A VFDR-BR1B-003, VFDR-BR1B- Modification ESR-12-0142 ensures that

006, and VFDR-BR1B-012. the metal enclosed box lined withHemyc protects EDG power cables(including L0318) in Fire Area BR1B.Currently, the Hemyc is only credited forprotecting the enclosed cables for 25minute beyond the point where a HGLlayer damaging cables occurs. Whilethis modification does not provide a fireprotection level that meets thedeterministic requirements of NFPA 805(Section 4.2.3), it helps reduce the riskassociated with the loss of ElectricalTrain A, which challenges the vitalauxiliary nuclear safety performancecriterion, and for which several VFDRshave been identified.

ESR-12-0143 N/A VFDR-ABBM-030, VFDR-ABI- Modification ESR-12-0143 provides an013, VFDR-ABI-036, VFDR-CC- additional diesel generator that is014, VFDR-CC-019, VFDR-CC- capable of supporting that RCS injection027, VFDR-CT-021, VFDR-CT- pump installed per ESR-12-0144 and 1030, VFDR-CT-031, VFDR-CT- SAFW pump. This modification does not048, and VFDR-SH-006. provide a deterministic resolution of

VFDRs. Rather, by providing a reliableand diverse source of power for SAFWpumps, it helps reduce the risk

associated with the loss of decay heatremoval, for which several VFDRs havebeen identified.

PRA RAI 015

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Modification Resolved Modification Helps Reduce Technical BasisESR-12-0144 VFDR-ABI-016 VFDR-ABBM-007, VFDR-ABBM- Modification ESR-12-0144 provides a

008, VFDR-ABBM-010, VFDR- new RCS injection pump in the StandbyABBM-011, VFDR-ABBM-012, Auxiliary FeedWater (SAFW) buildingVFDR-ABBM-025, VFDR-ABI- with its dedicated support systems,002, VFDR-ABI-006, VFDR-ABI- including a 10,000 gallon tank, which022, VFDR-ABI-023, VFDR-ABI- used to be the SAFW Pump028, VFDR-BOP-004, VFDR- Condensate Test Tank (TCD01). WithBR1A-007, VFDR-BR1A-020, the implementation of this modification,VFDR-BR1B-016, VFDR-CC- the piping connections of TCDO1 with004, VFDR-CC-010, VFDR-CC- the SAFW system will be removed.022, VFDR-CC-023, VFDR-CC- Thus, this modification deterministically024, VFDR-CC-025, VFDR-CT- resolves VFDR-ABI-016, which004, VFDR-CT-006, VFDR-CT- identified a potential SAFW flow025, VFDR-CT-026, VFDR-CT- diversion to TCDO1 (that is, the potential027, VFDR-EDG1A-010, VFDR- failure of the decay heat removalEDGIB-010, VFDR-PA-008, success path identified in the VFDR willVFDR-SH-013, VFDR-SH-014, not exist in the post-transition plant). Forand VFDR-YARD-008. all other VFDRs linked to Modification

ESR-12-0144, the modification providesno deterministic resolution, but areduction in risk, because the diverse

charging system helps mitigatechallenges to the reactor pressure andinventory nuclear safety performancecriterion.

PRA RAI 016

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Modification Resolved Modification Helps Reduce Technical BasisESR-12-0146 N/A VFDR-ABBM-019, VFDR-CC- Modification ESR-12-0146 provides

003, and VFDR-CT-050. automatic closure of letdown valve 371,using the 2 out of 2 low pressurizerpressure protected channels ofModification ESR-12-0125. Becauseproper polarity hot shorts might still

defeat the mitigative effects of themodification, there is no deterministicresolution of VFDRs. However, themodification helps mitigate the riskassociated with failing to isolate theletdown valve, for which several VFDRshave been identified. This modificationwas primarily designed to provideprotection against fire-induced damagein the control building and the cabletunnel (Fire Areas CC, BR1A, BR1B,and CT), but is also effective in otherlocations of the plant, provided thatthere is at least one fire scenario wherethe cables associated with themodification are not impacted by thefire, which is the case, for example, inFire Area ABBM.

ESR-11-0305 N/A VFDR-ABI-014 and VFDR-RC- Modification ESR-11-0305 installs011 upgraded seals for the reactor coolant

pumps. Due to the industry issues

associated with this modification, othermodification alternatives are beingexamined. Both the current proposedmodification and any alternativemodification will not providedeterministic resolution of any VFDR,but these will help to reduce the riskassociated with the failure of the reactorcoolant pump seals, for which VFDRshave been identified.

PRA RAI 017

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Modification Resolved Modification Helps Reduce Technical BasisESR-11-0050 N/A VFDR-ABBM-030, VFDR-ABI- Modification ESR-11-0050 provides a

013, VFDR-ABI-036, VFDR-CC- large dedicated water source to feed the014, VFDR-CC-019, VFDR-CC- SAFW pumps, a 1000kW DG that can027, VFDR-CT-021, VFDR-CT- power the pumps, a new minimum flow030, VFDR-CT-031, VFDR-CT- line, and new manual discharge by-pass048, and VFDR-SH-006. valves. Although the primary concern

for these VFDRs is another watersource beyond service water, thismodification provides additionalimprovements. Even so, thismodification does not provide adeterministic resolution of VFDRs.Rather, by providing a diverse source ofwater for SAFW pumps, it helps reducethe risk associated with the loss of

decay heat removal, for which severalVFDRs have been identified.

ESR-12-0412 VFDR-BOP-012 VFDR-ABBM-038, VFDR-BOP- A hot short on Cable L0365 or L0475and VFDR-EDG1B- 003, VFDR-BOP-013, VFDR-CC- could prevent an EDG from being014 044, and VFDR-CT-049. started and could also fail the EDG fuel

transfer oil pumps. Modification ESR-12-0412 provides fusing that wouldpreclude that failure. Accordingly, thismodification provides a deterministicresolution for two VFDRs (VFDR-BOP-012 and VFDR-EDG1B-014) focused onthat issue. The modification does notdeterministically resolve the otherVFDRs linked with the modification,because these VFDRs identifyadditional issues that the modification

does not address.

PRA RAI 018

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VFDR

Deterministically VFDR Whose Risk the

Modification Resolved Modification Helps Reduce Technical Basis

ESR-13-0028 VFDR-CC-041 and VFDR-ABI-031 Modification ESR-13-0028 providesVFDR-CT-047. automatic reactor trip by opening of the

480 VAC power supply upstream of the

MG-sets, using the 2 out of 2 lowpressurizer pressure protected channels

of Modification ESR-12-0125. Thismodification provides a diverse means

to ensure a reactor trip in the fire areas

that the modification was designed to

protect, which includes Fire Areas CC

and CT, thereby providing adeterministic resolution of VFDR-CC-041 and VFDR-CT-047. Fire Area ABIincludes the intermediate building where

cables associated with the modificationrun. As such, no deterministicresolution is provided, but themodification reduces the risk associated

with the failure to trip the reactor in

scenarios not associated with the PZRpressure cables.

ESR-13-0029 N/A VFDR-ABBM-023, VFDR-ABI- Modification ESR-13-0029 provides

008, VFDR-CC-006, VFDR-CT- automatic closure of PORVs, using the003, and VFDR-RC-007. 2 out of 2 low pressurizer pressure

protected channels of Modification ESR-

12-0125. Because proper polarity hotshorts might still defeat automatic

PORV closure, the modification doesnot provide a deterministic resolution of

VFDRs. However, it helps mitigate therisk associated with spurious PORV

opening, for which several VFDRs have

been identified. This modification wasprimarily designed to provide protection

against fire-induced damage in the

control building and the cable tunnel(Fire Areas CC, BRIA, BR1B, and CT),

but is also effective in other locations ofthe plant, provided that there is at least

one fire scenario where the cables

associated with the modification are notimpacted by the fire, which is the case,for example, in Fire Area ABBM.

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Modification Resolved Modification Helps Reduce Technical BasisESR-13-0030 N/A VFDR-CC-002 and VFDR-CT- Modification ESR-13-0030 removes

007 power from the control circuits of

several containment ventilation isolationvalves, using the 2 out of 2 lowpressurizer pressure protected channelsof Modification ESR-12-0125. Becauseproper polarity hot shorts might stilldefeat the mitigative effects of themodification, no deterministic resolutionof VFDRs is provided. However, it helpsmitigate the risk associated with the lossof control of the isolation valves, forwhich several VFDRs have beenidentified.

ESR-12-0423 N/A N/A Although initially identified as an issue

under CR CR-2007-008452, this ESR isnow closed. Further detailed plantwalkdowns determined that thesprinklers were properly located and incompliance with the code. This isdocumented within ESR-12-0423.Therefore, this ESR has been closed.

ESR-11-0421 N/A N/A Modification ESR-11-0421 modifies thediesel fire pump control panel circuit toisolate the remote start circuit from thecontrol panel microcontroller. While thismodification was not found to helpresolve any specific VFDR, it helpsreduce the risk associated with thefailure of the diesel fire pump.

PRA RAI 0110

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PRA RAI 02

Different VFDRs identify different fire scenarios plant response challenges, yet the samemodification or recovery actions may be seen to be credited for the VFDRs. Given the differentfire scenarios expected to be encountered, how was it determined that the plannedmodifications can be credited as they have been in the FPRA for the VFDFRs? Discuss thetechnical basis, including success criteria considerations, or new supporting calculations, foreach modification or recovery action credited in the Fire PRA fire scenarios, and cite relevantreferences.

Response

VFDRs are linked to a plant modification or a recovery action based on the reading of the issueidentified in the VFDR and the mitigation effects afforded by modifications and recovery actions.

Refer to the response of PRA RAI 01 for the table giving the specific technical basis linking

modifications to VFDRs.

The table below provides the technical basis linking credited recovery actions to VFDRs.

Recovery Action VFDR Crediting the Recovery Technical Basis

Action

AFHFDSUPPL-3 VFDR-CC-018 and VFDR-CT-022 Recovery Action AFHFDSUPPL-3 consists ofAlso: VFDR-ABBM-030, VFDR- supplying alternate sources of water for the AFWABI-013, VFDR-ABI-036, VFDR- system once the initial water supply is depleted. For

CC-014, VFDR-CC-019, VFDR- the preferred AFW system, this is the two smallCC-027, VFDR-CT-021, VFDR- condensate storage tanks. For the post modificationCT-030, VFDR-CT-031, VFDR- (ESR-1 1-0050) stand-by AFW (SAFW) system, this isCT-048, and VFDR-SH-006. the large new tank. The AFW systems are designed to

use lake water as provided by the service water

pumps. In cases where service water is lost, the initialAFW water supplies must be replenished. Although

large new tank does have over 12 hours of condensategrade water, even that must eventually be refilled. As

such, this action represents re-filling an initial AFWwater supply as the tanks empty when required. Thisaction is credited in all fire scenarios that damageenough equipment where secondary system is lost

and the normal heat removal path (i.e. through themain condenser) is not available, residual heatremoval is lost (i.e. shutdown cooling), and service

water to the AFW pumps is lost. For the VFDRs listed,the only recourse is to re-fill the AFW tanks. VFDR-

CC-018 and VFDR-CT-022 are listed due to the loss of

indication on the preferred AFW CSTs. Althoughstrictly speaking, due to the new larger tank dedicated

to the SAFW pumps, the smaller CSTs associated withpreferred AFW pumps are not required, the recovery

action is credited because water must be replenished

to an AFW pump.

PRA RAI 021

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Recovery Action VFDR Crediting the Recovery Technical Basis

Action

AXHFDSAFWX-2 VFDR-ABBM-030, VFDR-ABI- Recovery Action AXHFDSAFWX-2 consists of aligning013, VFDR-ABI-036, VFDR-CC- a Standby AFW (SAFW) pump. For these associated014, VFDR-CC-019, VFDR-CC- VFDRs, this align is done locally using the

027, VFDR-CT-021, VFDR-CT- Modifications associated with ESR-11-0050. This030, VFDR-CT-031, VFDR-CT- modification includes a dedicated large condensate

048, and VFDR-SH-006 storage tank, a diesel generator to locally power thepumps, and a local minimum flow system that cannotbe damaged by a fire outside the SAFW building. It is

credited, in conjunction with Recovery ActionAFHFDSUPPL-3, to help resolve VFDRs where the

turbine-driven AFW pump is found to be failedbecause of fire-induced damage, or where control of

the SAFW from the Control Room is lost, butrecoverable with local alignment.

DCHFDTSCLT VFDR-ABBM-014, VFDR-CC-029, Recovery Action DCHFDTSCLT consists of locallyand VFDR-CT-033. aligning the TSC diesel generator or 100kW portable

DG ultimately to provide power to the battery chargers.It is credited to help resolve VFDRs that identified the

need to have power for long-term DC availability. Thisrecovery action is used in conjunction with Recovery

Action FSHFDTSCLT-DR.

FSHFDTSCLT-DR VFDR-ABBM-014, VFDR-CC-029, Recovery Action FSHFDTSCLT-DR consists of locallyand VFDR-CT-033. using the TSC bus work or 100kW DG to ultimately

power a 125 VDC charger. It is credited to help resolve

VFDRs that identified the need to have power for long-term DC availability. This recovery action is used inconjunction with Recovery Action DCHFDTSCLT.

DGHFDER-DG- VFDR-ABBM-009, VFDR-ABI- Recovery Action DGHFDER-DG-LOCAL consists ofLOCAL 020, VFDR-BOP-003, VFDR- locally starting an emergency diesel generator

BOP-013, VFDR-BR1A-009, (KDG1A or KDG1B). It is credited to help resolveVFDR-BRIA-018, VFDR-BR1B- VFDRs that point to a potential loss of credited 480004, VFDR-CC-008, VFDR-CC- VAC Bus 14, 16, 17, or 18.

020, VFDR-CC-021, VFDR-CC-026, VFDR-CC-044, VFDR-CT-

012, VFDR-CT-018, VFDR-CT-023, VFDR-CT-023, VFDR-

EDG1B-003, VFDR-and EDG1B-

012.

FSHFDMCBDC-B VFDR-CC-003, VFDR-CC-006 Recovery Action FSHFDMCBDC-B consists of locally

depowering DC loads. It is credited to help resolve

VFDRs that call for this recovery action.

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Recovery Action VFDR Crediting the Recovery Technical BasisAction

RCHFDMAKEUP VFDR-ABBM-007, VFDR-ABBM- Recovery Action RCHFDMAKEUP consists of locally008, VFDR-ABBM-010, VFDR- aligning and starting the new injection system,ABBM-011, VFDR-ABBM-012, associated with Modification ESR-12-0144. It isVFDR-ABBM-025, VFDR-ABI- credited to help resolve the VFDRs where there is a002, VFDR-ABI-006, VFDR-ABI- challenge to the reactor inventory control. Inventory022, VFDR-ABI-023, VFDR-ABI- control loss is a function of the reactor system integrity028, VFDR-BOP-004, VFDR- being lost as well as an injection source (charging,BR1A-007, VFDR-BR1A-020, safety injection pumps, or the new injection pumpsVFDR-BR1B-016, VFDR-CC-004, (ESR-12-0144) being lost. A contributing factor to theVFDR-CC-010, VFDR-CC-022, loss of the currently installed charging and safetyVFDR-CC-023, VFDR-CC-024, injection systems is the associated normal boratedVFDR-CC-025, VFDR-CT-004, water supply which is the refueling water storage tankVFDR-CT-006, VFDR-CT-025, (RWST). The RWST inventory can be depleted due toVFDR-CT-026, VFDR-CT-027, RWST drain down.VFDR-CT-028, VFDR-EDG1A-010, VFDR-EDG1B-010, VFDR-PA-008, VFDR-SH-013, VFDR-SH-014, and VFDR-YARD-008.

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PRA RAI 04

For VFDR-CC-044, discussed in the LAR, modification ES-12-0142 provides a 45 minute fireprotection for cable 00687, yet this cable is not mentioned in the VFDR. Please clarify.

Response

Modification ESR-12-0142 was mistakenly included in the disposition of VFDR-CC-044. Inreality, this modification is not credited for a fire in Fire Area CC. Modification ESR-12-0412 iscredited.

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PRA RAI 05

Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall beacceptable to the NRC. RG 1.205 identifies that recovery actions must be addressed. TableW-4 of the LAR provides the results of the additional risk of recovery actions. However, it is notclear if both previously approved and new recovery actions are included in the results. Discusshow your treatment of previously approved and new recovery actions credited in the Fire PRA isconsistent with the guidance in RG 1.205. Discuss your treatment of primary and secondaryrecovery actions credited as well.

Response

Conservatively, no recovery actions were credited in the delta risk evaluation as beingpreviously approved. Although it is possible that some recovery actions could be consideredpreviously approved, the majority of the delta risk is due to the human actions associated withalignment of the new equipment. As such, the conservative delta risk calculation is not overlyconservative.

PRA RAI 07 provides details about how recovery actions were developed and credited for theR.E. Ginna fire PRA. The response to PRA RAI 07 also includes details about the definitions ofrecovery actions that could be considered previously approved.

Neither RG 1.205 nor the associated FAQs explicitly define Primary and Secondary recoveryactions. Our interpretation of the terms "primary" and "secondary" actions fit similar definitionsthat we have developed in the HRA for R.E. Ginna's fire PRA. Primary recovery actions areconsidered to be the "New Recovery Actions" and "Previously Approved Recovery Actions",which are defined in detail in the response to PRA RAI 07. Similarly, secondary actions areconsidered to be "Previously Approved PCS Actions" and "Other Local Actions", which are alsodefined in detail in the response to PRA RAI 07. While we don't explicitly define these actionsas "primary" and "secondary" actions, the analysis is sufficient to support the requirements inRG 1.205 and the associated FAQs.

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PRA RAI 06

Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall beacceptable to the NRC. RG 1.205 identifies that recovery actions must be addressed. Forrecovery actions associated with modifications, please provide a general discussion on themethod of quantifying the human error probability (HEP), the HEPs assigned, and discuss howthe verification of assumptions made in quantifying the HEP (e.g., timing) are addressed in theimplementation items in Attachment S of the LAR.

Response

The recovery actions associated with proposed modifications are evaluated in a similar fashionto the actions associated with the existing procedures and equipment. For example, theyundergo a feasibility assessment according to the exact format of the feasibility FAQ 07-030 andare evaluated using detailed quantitative analysis with the EPRI HRA Calculator to estimate anHEP for use in the fire PRA.

The main difference in the evaluation of recovery actions associated with modifications is thatthe PRA team works with Operations supervisors to establish a proposed procedure outline.The PRA team specifically lays out with Operations what needs to be done and when it needs tobe done. Once Operations agrees on the reasonableness of the proposed approach, operatorsare interviewed as if the procedure exists and the procedure outline is modified as appropriateaccording to the findings from the interview.

Regarding HEP timing, the design specifications for the modified systems are developed suchthat the existing thermal/hydraulics (T/H) calculations can still be used.

For example, the current charging system is credited as providing 75 gpm to the RCS.Calculations show that with 75 gpm injection, the RCS pressure will drop below 1500 psiabefore equilibrium is achieved. Therefore, the PRA specified that the new injection pump mustbe able to deliver 75 gpm at an RCS pressure of 1500 psia. This inherently requires a higherdischarge head of the pump to account for line losses between the pump and the RCS, which inturn factors into the pump design specs.

The new injection system design specs were included in the details behind Item 9 of LARAttachment S Table S-2 Plant Modifications Committed, ESR-12-0144: Installing a new injectionpump in the SBAFW building.

The design specs related to this modification correspond to the timing in the HEPs evaluated forthe recovery actions listed in LAR Attachment G as "Operators locally align and start the newinjection system associated with Plant Modification ESR-12-0144".

In this way, the modifications are correlated to the assumptions made in the HEPs calculated forthe recovery actions.

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PRA RAI 07

With respect to Frequency Asked Questions (FAQ) 08-0054, Rev 1 and FAQ- 07-0030, discusswhich method(s) were employed to evaluate the additional risk of recovery actions from firescenarios outside the main control room, fire scenarios involving main control roomabandonment, or fire scenarios which may involve actions both in the MCR and at a primarycontrol station (PCS). Include a discussion of how the guidance for the additional risk ofrecovery actions was evaluated with FAQ 08-0054, Rev. 1 and the section "Additional Risk ofRecovery Actions - Alternate or Dedicated Shutdown." Discuss how primary control stations,planned modifications, procedures, and recovery actions were considered. Also, for the maincontrol room additional risk of recovery action evaluation, describe the compliant case used. Inparticular discuss how shutdown panels ABELIP and IBELIP are treated for the compliant plantgiven that the safe shutdown strategy has changed to use new equipment and does not requireuse of these two shutdown panels, according to the LAR Attachment G.

Response

FAQ-0054 discusses options for assessing the delta-risk of recovery for dedicated and alternateshutdown (i.e., MCR abandonment scenarios). It first discusses a bounding approach, and thenstates:

"If this bounding treatment is judged to be overly conservative, then it will be necessaryto further refine the fire PRA so that those recovery actions are isolated and treatedseparately in the Fire PRA so that their specific risk contribution can be determined."

Ginna determined that the bounding approach was overly conservative, and so has modeledeach recovery (and non-recovery) action associated with MCR abandonment explicitly in theFPRA, as allowed by the FAQ. How this is done is addressed within the sections below.

FAQ-0030 lists a number of processes that can be used to evaluate the additional risk of arecovery action. One of those processes is stated as follows:

"Model the recovery action explicitly in the Fire PRA, with an appropriate human errorprobability and calculate the CDF (LERF). Subtract the CDF (LERF) obtained byeliminating the VFDR in the PRA model to create a compliant case. This gives the ACDFand ALERF associated with performing the action compared to providing a deterministicresolution. "

One way to eliminate the VFDR is if the recovery action were possible to be performed fromeither the control room (for non-abandonment scenarios) or from the PCS (for abandonmentscenarios). By definition, actions taken in the MCR or at the PCS are not recovery actions. Asstated in FAQ-0030:

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"Activities that occur in the main control room as a result of fire damage in the plant arecompliant with NFPA 805 Section 4.2.3.1 and do not require an evaluation of the

additional risk of their use.

"Activities at primary control station(s), including activities to enable or activate the

primary control station(s) meeting the criteria set forth in Step 1, are free of fire damage

from the primary control station are compliant with NFPA 805 Section 4.2.3.1 and do not

require an evaluation of the additional risk of their use."

This is how Ginna has chosen to model the compliant plant. The VFDR has been eliminated in

the compliant plant by assuming the compliant plant has all of the actions taking place in theMCR or at the PCS (the ABELIP, the IBELIP, or a theoretical new PCS). Per the quotes above,such an arrangement would be compliant with NFPA 805 Section 4.2.3.1. In such a case, thedelta-risk would be calculated by replacing the HEP of the recovery action (non-compliant, orvariant, case) for the HEP of the action if it were being performed in the MCR or at the PCS

(compliant case). The cognitive portion of each action is the same and takes place in the MCR,

and so is not modified between the variant case and the compliant case. The execution portionof the action is moved from a plant location to the MCR or the PCS, and so needs to bemodified. Rather than perform a specific analysis to determine the HEP for the execution

portion at its new, non-recovery location, Ginna chose to bound the delta by assigning anexecution HEP of zero to the compliant plant action.

In order answer the rest of this RAI, it is necessary to discuss the various types of actions (from

an NFPA-805 perspective) that are considered in the FPRA. The actions considered in the

Ginna NFPA 805 FPRA are of five types: Each type and its treatment are discussed below.

New Recovery Actions - These actions are associated with the operation of the equipment usedto place a standby AFW pump and a new injection pump (ESR-12-0144) into operation using anew dedicated diesel generator (ESR-11-0050). This is the primary equipment credited withproviding heat removal and inventory control for scenarios where either or both is lost. Since

there are a number of plant areas where the NSCA showed that either or both of these couldoccur, which are therefore VFDRs because they fail the associated nuclear safety performancecriteria; these are clearly recovery actions since they keep one train free from fire damage. Thisapplies to both MCR abandonment and non-abandonment cases. The delta-risk of recovery isbased on a compliant plant where these actions could be performed from the MCR or a PCS.The delta-risk is bounded by assuming that the execution portion of these actions has an HEP

of zero (the cognitive portion occurs in the control room, even in the abandonment case, and so

is unchanged).

Previously Approved Recovery Actions - The actions are associated with de-energizing certain

panels in the battery rooms. Since there are a number of plant areas where the NSCA showedthat equipment tied to these panels could cause a loss of a nuclear safety performance criterionif the panels remain energized, which are therefore VFDRs, these are clearly recovery actions

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since they keep one train free from fire damage. This applies to both MCR abandonment andnon-abandonment cases. The delta-risk of recovery is based on a compliant plant where these

actions could be performed from the MCR or a PCS. The delta-risk is bounded by assuming

that the execution portion of these actions has an HEP of zero (the cognitive portion occurs inthe control room, even in the abandonment case, and so is unchanged).

Previously Approved PCS Actions - The designated PCS at Ginna is the IBELIP and ABELIP.

In the Appendix R plant, these actions were the primary credited actions for MCR abandonmentscenarios. However, in the post-transition plant it was decided to first use the equipmentassociated with the strategy to use the SBAFW pump and new injection pump. Therefore,these actions do not serve to keep one train free from fire damage - they provide defense-in-

depth for the new recovery actions. In addition, although in the abandonment scenario theseactions are no longer the first actions performed, the IBELIP and ABELIP still maintain their pre-transition designation as a PCS at Ginna and so actions performed there are PCS actions.

Since FAQ-0030 states "Activities that take place at primary control station(s) or in the MainControl Room are not recovery actions, by definition." the IBELIP and ABELIP actions are notrecovery actions and there is no requirement to perform a delta-risk calculation.

Other Local Actions - There are other actions modeled in the FPRA that take place outside the

main control room as part of the standard AOPs and EOPs, but are not for the purpose ofrecovering a success path to meet the nuclear safety performance criteria. FAQ-0030 states"Actions taking place outside the main control room that are modeled in the PRA but are not

involved with demonstrating the availability of a success path to meet the Nuclear SafetyPerformance Criteria are not considered recovery actions requiring the evaluation of additionalrisk required by NFPA 805 Section 4.2.4.". Therefore, the delta risk associated with theseaction is not calculated.

MCR Actions - There are, of course, numerous actions that take place in the MCR. Of particularnote for NFPA 805 are actions to close certain LOCA and LERF paths and to diagnose certain

functional losses. Ginna will install a dedicated panel in the control room along the exit path thatwill have protected controls for accomplishing the actions and a protected instrument train thatwill be available if all other instrumentation fails. Procedures will be modified to instruct the useof this panel for non-abandonment scenarios when MCB actions are unavailable and forabandonment scenarios when preparing for and executing abandonment. Since these occurwithin the control room, they are not recovery actions and the delta risk associated with these isnot calculated.

Overall Discussion of Compliant Case for MCR Abandonment

* In the compliant plant, the new recovery actions are assumed to take place at a PCS

instead of away from the PCS. This could be either by adding the controls for the newactions to the already existing ABELIP and IBELIP, or at a new PCS specifically for thatpurpose.

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" In order to simplify the analysis of delta-risk, and to render the location of the PCS usedto take those actions moot in the analysis, the execution HEP for the actions is set tozero for the compliant case.

" The execution portion of the previously approved recovery actions is also set to zero forthe compliant case.

• The ABELIP and IBELIP actions that are still in the procedures are not recovery actionsbecause (1) they no longer are credited as the means to meet a nuclear safetyperformance criterion and (2) they are previously-approved PCS actions that take placeat a previously-approved primary control station. Therefore, they are not changed in thecompliant plant model.

• The actions that take place in the control room, including the cognitive decision to

abandon and implement the abandonment procedures and the actions that take place atthe new dedicated panel located in the MCR, are by definition not recovery actions and

so are not adjusted in the compliant model.

Further discussion on how the procedures, and the revisions to the procedures, were addressedis contained in the responses to PRA RAIs 05, 06, and 23.

Further discussion of compliant plant modeling is contained in the response to PRA RAI 11.

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PRA RAI 08

In the LAR Attachment G, Table G-i, explain what is meant by "Risk" In the column "RA/PCS?"

Response

As stated on page G-3 of the LAR, the RA/PCS column indicates that these are actionsnecessary to address risk; however, the column headings in the table indicate "RA/PCS", whichis does not clearly correlate to the text on page G-3. The table will be updated to categorizeactions as "RA", consistent with the column heading for Table G-I. The text will be updated todefine the items in the table as "RA" instead of "RISK" as follows:

"Activities that are identified as Recovery Actions that are necessary to address risk areidentified as RA."

Additionally, to further clarify the text associated with Table G-I, the following paragraph will be

deleted from page G-3:

"Table G-1 - Recovery Actions and Activities Occurring at the Primary Control Station(s) identifythe activities that occur at the primary control stations. Activities necessary to enable theprimary control stations are also identified in Table G-1 as primary control station(s) activities(identified as PCS). These activities do not require the treatment of additional risk. Activities thatare identified as Recovery Actions that are necessary to address risk are identified as RISK."

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PRA RAI 09

The LAR Attachment G indicates that PCS actions are in Table G-1; however, only recoveryactions appear to be in Table G-1. Please provide the PCS actions, or clarify the information inTable G-1.

Response

There are no recovery actions that take place at a PCS that are required to resolve a VFDR;therefore, the title of Table G-1 will be updated to "Recovery Actions". The text on page G-3 willbe updated to reflect the updated table title. This text currently states that the PCS actions arenot required to resolve any VFDRs, so this text remain as-is. The updated text will read:

"For the reasons stated above, Table G-1 - Recovery Actions does not include any requiredactivities that occur at the primary control stations (i.e., PCS actions are no longer required inorder to address any VFDRs), because with the installation of the new equipment, there arenone required. Since the new equipment is not controlled from a PCS, all actions associatedwith it are considered RAs. Activities that are identified as Recovery Actions that are necessaryto address risk are identified as RA in the G-1 Table."

To further clarify the text associated with Table G-1, the following paragraph will be deleted frompage G-3:

"Table G-1 - Recovery Actions and Activities Occurring at the Primary Control Station(s) identifythe activities that occur at the primary control stations. Activities necessary to enable theprimary control stations are also identified in Table G-1 as primary control station(s) activities(identified as PCS). These activities do not require the treatment of additional risk. Activities thatare identified as Recovery Actions that are necessary to address risk are identified as RISK."

Additionally, there is an error correction that will be made to the table as well. Action BRIA-RSKRA-2 will be updated to the category "RA" in the table, not "PCS".

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PRA RAI 10

Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall beacceptable to the NRC. RG 1.205 identifies that recovery actions must be addressed. Explainwhich recovery actions and modifications are being proposed to meet risk acceptance criteria?

Response

The approach taken for the Ginna NFPA 805 project is that the recovery actions andmodifications included in the post-transition plant are those required in order to meet the RG1.174 requirements for CDF, LERF, A-CDF, and A-LERF, per RG 1.205. Therefore, allrecovery actions and modifications are being proposed to meet the risk criteria. For furtherdiscussion of recovery actions, see responses to PRA RAIs 5, 7, and 11. For the modificationrelationship to VFDRs, see response to PRA RAI 1.

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PRA RAI 11

Describe the method(s) used to determine the fire area changes in risk (delta-CDF and delta-LERF) reported in the LAR Appendix W for fire scenarios outside the main control room, firescenarios involving main control room abandonment, or fire scenarios which may involveactions both in the MCR and at a PCS. The description should include a summary of PRAmodel additions or modifications needed to determine the reported changes in risk. If any ofthese model additions used data or methods not included in the FPRA peer review pleasedescribe the additions.

With respect to FAQ 08-0054, Rev 1 and FAQ 07-0030, indicate which FRE method describedin the FAQs was used to evaluate the VFDRs. Include in the discussion how recovery actions,whether primary or secondary, are treated. Also, for the compliant case, it is noted in the LARthat recovery actions are set to success to represent a plant where the VFDR is deterministicallyresolved. Discuss the practice of setting recovery actions to success and its equivalency toremoving the VFDR.

Response

The method used to determine the delta CDF and delta LERF are described in "R. E. GinnaNuclear Power Station NFPA 805 Fire Risk Evaluations" (HAI-0028-0011-002-003). Sections5.3.2, 5.3.3, and 5.3.4 provide an overview of the post-transition plant, the deterministicallycompliant plant, and the methods for evaluating VFDRs and calculating delta risks, respectively.

Post-Transition Plant

The post-transition plant (or variant plant) is the pre-transition plant withproposed modifications, including procedure changes required for the NFPA 805transition. In addition, the post-transition plant credits recovery actions to helpresolve several VFDRs.

Deterministically Compliant Plant

For the NFPA 805 transition, the deterministically compliant plant at REG is thepost-transition plant with VFDRs assumed deterministically resolved. That is, it isa hypothetical post-transition plant in which the existing VFDRs have beenresolved and the fire areas meet the requirements of Section 4.2.3 of NFPA 805.

Methods for Evaluation of VFDRs

To evaluate the delta risks, that is, the difference in CDF (and LERF) betweenthe post-transition plant and the deterministically compliant plant, each of the 329open items identified as a VFDR in the NSCA Report is examined individuallyand a path forward for its resolution is developed.

The recovery actions that were considered in the evaluation of the delta risks are included inthese sections.

PRA RAI 11

1

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The methods used to develop the fire PRA were all included in the peer reviewed plant model.There were some data updates made to the model between the peer review and the submittal,but, there were no new methods introduced between the peer review and the submittal thatwould require an additional peer review.

FRE methods that were used to evaluate the VFDRs are described in sections 5.3.2 and 5.3.4of "R. E. Ginna Nuclear Power Station NFPA 805 Fire Risk Evaluations". This report wasgenerated considering guidance from FAQ 08-0054, Rev 1 and FAQ 07-0030 as stated in thereferenced sections as well as the introductory material in section 5.2.

For the compliant case, if a recovery action was required to meet a critical safety function in agiven fire area, then that recovery action was assumed to be done from the control room in thecompliant plant model (the execution portion of the HEP was set to zero). Additional detailedinformation about recovery actions can be found in the responses to PRA RAI 05 and PRA RAI07.

PRA RAI 11

2

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PRA RAI 12

For the main control room analysis, detailed human reliability analysis (HRA) was performedfollowing NUREG-1921. Confirm that the feasibility of the operator actions was considered,consistent with NUREG-1921 guidance, in addition to the detailed reliability analysis. For thoseHEPs which were screening values, describe the methodology since, based on the audit, itappears to be different from the main control room screening approach described in NUREG-1921.

Response

The majority of the time, a fire in the Ginna main control room (MCR) causes abandonment.Ginna has modeled each recovery action associated with MCR abandonment explicitly in theFPRA, as allowed by FAQ-0054. How this is done is addressed in detail in the response toPRA RAI 07, but as the RAI suggests, detailed HRA was used for the quantification of theserecovery actions.

Beyond the qualitative assessment associated with detailed analysis in the HRA Calculator,these recovery actions were evaluated according to the exact format of FAQ 07-030 (cited in thefeasibility assessment section of NUREG-1921), which lists 11 required criteria for establishingthe feasibility of recovery actions. Each recovery action was evaluated for feasibility against the11 criteria, and the results are documented in Appendix I of the Fire HRA notebook (G1-HRA-F001, Rev: 3). The feasibility analysis was reviewed by Operations personnel, who concludedthat the assessment of each item was appropriate. It is expected that these actions will beincluded as time critical operator actions in procedure A-601.10 Time Critical ActionManagement Program to ensure that all plant and procedure changes which may impact thefeasibility of these actions are evaluated.

There are also very rare cases where the fire in the control room is small enough such thatabandonment is not required, either because equipment damage is not sufficiently severe orenvironmental conditions do not warrant leaving the MCR.

To address this situation, the baseline fire case HFEs were reviewed to identify those whichwould be impacted by a fire in the MCR that did not require abandonment. Screening HEPvalues were assigned to the in-MCR fire HFEs according to the following criteria:

" If the baseline fire HEP was less than 0.05, 0.1 was assigned as the in-MCR fire value." If the baseline fire HEP was greater than 0.05, 2 times the baseline value was assigned

as the in-MCR fire value.

During the various iterations of fire PRA model quantification, the cutsets were reviewed forsignificant HFE contributors to risk. If in-MCR HFEs were found to exist in dominant cutsets,they were reviewed and, where justifiable, re-quantified with detailed analysis to replace theinitial screening values. The adjustments made to the baseline fire case HEPs for this detailedanalysis are described in section 4.2.3 on In-MCR Fire Evaluation of the Fire HRA notebook(G1-HRA-FOO1, Rev: 3). In general, the adjustments include:

" Increasing the median response time" Increasing the manipulation time" Increasing cognitive and execution complexity

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0 Adjusting for smoke and accessibility factors

There are also a few cases where the screening values used in the internal events model werecarried over to the fire model. These account for 6% of the internal events HEPs (10-of-164).Two of these are 1.0, one is 0.9, two are 0.25, three are 0.1, and two of the LERF HEPs are0.04. The two LERF HEPs are associated with mechanical only penetrations and are notsignificant to fire risk.

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PRA RAI 13

Provide a discussion of your procedure(s)/process(es) for plant change evaluations post-transition. Include a discussion on how post-transition guidance for plant change evaluationsaddresses key uncertainties, assumptions, sensitivity analyses, and peer review facts andobservations (F&Os) (e.g., unaddressed F&Os).

Response

Changes affecting the design, operation, or maintenance of the plant are performed inaccordance with fleet procedure CNG-CM-1.01-1003, "Design Engineering and ConfigurationControl." CNG-CM-1.01-1003 ensures that reviews are performed to determine if plant changesimpact the fire protection program documentation. This procedure has been revised to reflectthe requirements of NFPA 805. Engineering standard CNG-FES-007, "Preparation of DesignInput and Change Impact Screen," which compliments CNG-CM-1.01-1003, has also beenrevised to add a section for NFPA 805 applicability criteria and required actions. Revisions toPRA documents and analyses are governed by procedure CNG-CM-1.01-3003, "ProbabilisticRisk Assessment Configuration Control" and may be invoked by CNG-CM-1.01-1003. CNG-CM-1.01-3003 includes procedures for updates that may involve addressing key uncertainties,assumptions, sensitivity analyses, and peer review F&Os.

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PRA RAI 14

Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall beacceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodologyfor conducting a fire PRA and endorses, with exceptions and clarifications, Nuclear EnergyInstitute (NEI) 04-02, revision 2, as providing methods acceptable to the staff for adopting a fireprotection program consistent with NFPA-805.

Describe how suppression is treated in your analysis. Address suppression with regards toprotection of cable trays. Also describe your treatment of suppression with regards toprogression of events.

Response

Suppression treatment in the Ginna fire PRA is documented in section 6.1.3.3 of G1-FSS-FOO1.For suppression credit in the progression of events specifically:

For each fire protection feature under consideration and for each fixed or transient ignitionsource identified, a fire protection evaluation was conducted to determine if the fire protectionfeature is expected to perform its intended function, (i.e., the suppression system is expected tocontrol/suppress the fire considering the postulated fire generated conditions). For water basedfire protection features with a tray immediately above the ignition source, sprinklers are onlycredited for protection after damage to the first tray. The first tray directly above the ignitionsource is failed and then sprinklers are credited. Targets above the first tray are damaged withthe credit of the suppression failure probability. Therefore, the credit for interveningcombustibles is only applied when confirmed by walkdowns that sprinklers are located betweencable trays and with the conservatism that the first tray would be damaged. For the gas basedsystems (i.e. Halon) credit is provided for preventing tray damage above panels. All equipmentinside the panel is considered damaged as well as any cables in connecting conduits. G1-FSS-F001 will be updated to include Halon activation calculation for time to activation to support thecredit for Halon suppression to prevent damage to items not in direct contact with the ignitionsource.

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PRA RAI 16

A number of VFDRs involve HEMYC wrap. Identify if any other VFDRs in the LAR involveperformance-based evaluations of wrapped or embedded cables. If applicable, describe howwrapped or embedded cables were modeled in the Fire PRA including assumptions and insightson how the PRA modeling of these cables contributes to the VFDR delta-risk evaluations,including the modeling method and assumptions associated with the cables.

Response

HEMYC is currently only credited in Battery Room B. 0028-0018-000-001 HEMYCEVALUATION Rev 2 documents the qualification and modeling of the HEMYC in the Fire PRA.HEMYC is only credited for a group of cables which the analysis indicates will be damaged at25 minutes, after the HGL of the room reaches thermoplastic damage criteria. These cables,protected by the HEMYC, are failed at 25 minutes, with the associated non-suppressionprobability.

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PRA RAI 17

The staff's review of the LAR Attachment H did not identify FAQ 08-0053, "Kerite-FR CableFailure Thresholds." Clarify whether Kerite insulated cables are utilized at the plant, and if theyare included in the Fire PRA. If so, describe their modeling such as the cable damagethreshold.

Response

Kerite cable is present in the R.E. Ginna plant. However, in the Fire PRA, all targets areassumed to be of thermoplastic material for damage criteria purposes with one exception. Theexception applies to cables in 5 conduits in Battery Room A. These 5 conduits were verified tohave thermoset material (Ginna Key Input 83, EWR-1444, and PCR-98-015) and the damagecriteria of 330 C and 11 kW/m 2 described in NUREG/CR-6850 Table 6-2 was assigned to them.In summary, Kerite cables within the scope of the Fire PRA are assigned thermoplastic damagecriteria per Table 6-2 in NUREG/CR-6850.

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PRA RAI 20

RG 1.174, Revision 2, identifies that key sources of model uncertainty should be identified andsensitivity analysis performed or reasons given as to why they are not appropriate for theapplication. In the ASME/ANS PRA standard a source of model uncertainty is labeled as "key"when it could impact the PRA results that are being used in a decision, and consequently, mayinfluence the decision being made.

With respect to SR FSS-E3, provide a list of the input parameters were considered foruncertainty intervals, and their uncertainty treatment for the Fire PRA.

Response

The uncertainty analysis is documented in G1-UNC-FOO1, "Fire PRA Notebook Uncertainty andSensitivity Analysis (UNC)". All of the fire-related tasks were considered for uncertainties.Many of the inputs are conservative or deterministic in nature. These inputs are not subject touncertainty analysis. Table 3 of G1-UNC-FOO1 provides a summary of the uncertainty treatmentfor parameters of each fire PRA task.

Monte Carlo sampling was performed to propagate parametric uncertainty through the GinnaFPRA model. The analysis was performed on the following parameters:

" Fire ignition frequencies" Human error probabilities* Existing internal events component random failure probabilities and unavailability's* Circuit failure probabilities.

The parameters that were evaluated for sensitivity analyses included:

" Human error probabilities• Common cause factors" Circuit failure probabilities" Supplement 1 ignition frequencies with an alpha factor of 1 or less" Automatic Suppression.

Detailed information on the parameters used in the uncertainty and sensitivity analyses can befound in G1-UNC-FOO1.

The peer review team categorized the Fire Uncertainty SR, FSS-E3, as being CC III in LTR-RAM-Il-12-066, "Ginna Fire Peer Review".

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PRA RAI 21

Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptableto the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting afire PRA and endorses, with exceptions and clarifications, NEI 04-02, revision 2, as providingmethods acceptable to the staff for adopting a fire protection program consistent with NFPA-805. RG 1.200 describes a peer review process utilizing an associated ASME/ANS standard(currently ASME/ANS-RA-Sa-2009) as one acceptable approach for determining the technicaladequacy of the PRA once acceptable consensus approaches or models have been established.ASME/ANS-RA-Sa-2009 describes when changes to a PRA should be characterized as a "PRAupgrade." Address the following questions regarding the internal events or fire PRA peer review andchanges made to the internal events or fire PRA subsequent to your most recent full-scope peerreview:

a) Did the peer reviews for both the internal events and fire PRAs consider the clarifications andqualifications from Regulatory Guide (RG) 1.200, Revision 2, "An Approach for Determiningthe Technical Adequacy of Probabilistic Risk Assessment Results for Risk-InformedActivities," March 2009 (ADAMS Accession No. ML09041 0014) to the ASME/AMS PRAStandard? If not, provide a self-assessment of the PRA model for the RG 1.200 clarificationsand qualifications and indicate how any identified gaps were dispositioned.

b) ASME/ANS-RA-Sa-2009 describes when changes to a PRA should be characterized as a"PRA upgrade," and RG 1.200 Revision 2 provides clarification on PRA upgrade. Identifyany such changes made to the internal events or fire PRA subsequent to your most recentfull-scope peer review. Also, address the following:

If any changes are characterized as a PRA upgrade, indicate if a focused-scope peerreview was performed for these changes consistent with the guidance in ASME/ANS-RA-Sa-2009, and describe any findings and their resolution.

ii. If a focused-scope peer review has not been performed for changes characterized asa PRA upgrade, describe what actions will be implemented to comply with theASME/ANS standard.

iii. If any changes for which the methodology utilized in the current fire PRA differs fromthat evaluated by the peer review such that a reduced capability category wouldresult, describe what actions will be implemented to comply with the ASME/ANSstandard. If this means that CC-Il or greater is not met, then provide justification as towhy this is acceptable for transition to NFPA 805.

Response

a) Yes, the clarifications and qualifications from Regulatory Guide (RG) 1.200, Revision 2, "AnApproach for Determining the Technical Adequacy of Probabilistic Risk Assessment Resultsfor Risk-Informed Activities," March 2009 to the ASME/AMS PRA Standard were used for thepeer reviews for both the Internal Events and Fire PRAs.

b) There were no changes to the Internal Events PRA or Fire PRA after their respective peerreviews that would be considered PRA Upgrades. There were no methodology changesmade to either analysis post-peer review. There were several significant modelimprovements made to the Fire PRA after the peer reviews, but which were not characterizedas a PRA Upgrade based on the ASME/ANS-RA-Sa-2009 definition.

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PRA RAI 23

Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall beacceptable to the NRC. Section 2.4.4.1 of NFPA-805 further states that the change in publichealth risk arising from transition from the current fire protection program to an NFPA-805 basedprogram, and all future plant changes to the program, shall be acceptable to the NRC. RG1.174 provides quantitative guidelines on core damage frequency, large early releasefrequency, and identifies acceptable changes to these frequencies that result from proposedchanges to the plant's licensing basis and describes a general framework to determine theacceptability of risk-informed changes. Table S-3 of the LAR provides items (procedurechanges, process updates, and training to affected plant personnel) that will be completed priorto the implementation of the new NFPA 805 fire protection program.

Implementation item 19 states "The following procedure change to be implemented as part ofNFPA 805 transition provides a reduction in risk: A procedural change, not to be implementeduntil all required modifications are installed, will eliminate ER-Fire 2,3, 4, 5, and 6." Pleasediscuss this item.

Response

The ER- FIRE series of procedures address fires in areas where, under Appendix R rules, therewere effects on safe shutdown equipment and possible spurious actuations. These procedurescontain actions to respond to these possible effects. ER- FIRE.1 is used when a fire causesmain control room abandonment. ER- FIRE.2, 3, 4, 5, and 6 are used when a fire occurs inother safe shutdown areas. Because the latter five procedures are event-based, rather thansymptom based, there is uncertainty as to the severity of fires in these areas that would requireimplementation of the procedure. If a fire occurs in a non-control room area, but is not severethen certain steps in the procedure actually remove available equipment from service. This canincrease risk. Equipment should only be removed from service when the removal is required tomitigate fire effects. The best way to achieve this goal is to create symptom based procedures.In support of this goal, all of the credited actions in the fire PRA model are being relocated fromER- FIRE.2, 3, 4, 5, 6 procedures to alternative actions in the emergency operating procedures(EOPs). The alternative actions will be symptom based. For example, if the charging pumpsand safety injection pumps cannot be started from the control room, then the alternative actionwould be to implement a fire attachment that starts the new injection pump from the stand-byAFW pump room. Since the current procedures are a key part of the Appendix R program,these procedures will be retained until transition to NFPA 805 is complete. The transition notonly includes implementing the modifications, but it also includes adding all the requiredalternative actions as symptom based responses to the EOPs. Once this occurs, ER- FIRE.2,3, 4, 5, and 6, will be cancelled as the key actions are integrated into the EOPs.

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PRA RAI 24

Section 2.4.3.3 of NFPA 805 states that the probabilistic safety assessment (PSA) (PSA is alsoreferred to as PRA) approach, methods, and data shall be acceptable to the authority havingjurisdiction (AHJ), which is the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting amethodology for conducting a fire PRA and endorses, with exceptions and clarifications,NEI 04-02, revision 2, as providing methods acceptable to the staff for adopting a fire protectionprogram consistent with NFPA-805. RG 1.200 describes a peer review process utilizing anassociated ASME/ANS standard (currently ASME/ANS-RA-SA-2009) as one acceptableapproach for determining the technical adequacy of the PRA once acceptable consensusapproaches or models have been established. The primary results of a peer review are theF&Os recorded by the peer review and the subsequent resolution of these F&Os.

Address the following questions on the dispositions to the internal events F&Os and SupportingRequirement (SR) assessment identified in Attachment U of the LAR that have the potential toimpact the fire PRA results and do appear to be fully resolved.

a) SC-A2. Confirm that the Fire PRA supporting calculations are based on the plant poweruprate.

b) SY-A10. Provide additional information on how the disposition was addressed and theevaluation of the impact on the Fire PRA.

c) SY-A14. Provide additional information on how the disposition was addressed and theevaluation of the impact on the FPRA.

d) QU-E4. Clarify and discuss whether internal events PRA key sources of uncertainty andassumptions remained key sources of uncertainty and assumptions in the FPRA, anddiscuss the results and significance of sensitivity analyses for them.

Response

a) The plant power uprate (1811 MWth) was used in the PCTRAN evaluations thatsupported development of the internal events PRA model success criteria that was usedas the basis for the Fire PRA.

b) The cited F&O was associated with the Feed and Bleed modeling in the internal eventsPRA. Specifically, the F&O stated that "The logic does not include 75 gpm charging flowwhich is noted in the Success Criteria notebook as required to support single PORVsuccess." The 75 gpm charging flow requirement was added to the internal events faulttree prior to development of the Fire PRA. This update to the model resolved the citedF&O. The Fire PRA used the revised internal events fault tree as a basis for the firemodel; therefore, the issue is addressed in the internal events PRA as well as the FirePRA.

c) The unavailability event (CDAACITYWATER) was added to the internal events model torepresent the loss of city water to the SAFW for loss of off-site power. This updateresolved the cited F&O. The internal events model was used as the basis for the FirePRA, therefore, the issue is addressed in the internal events PRA as well as the FirePRA.

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d) Internal events PRA key sources of uncertainty and assumptions were carried through tothe Fire PRA and were evaluated as part of G1-UNC-FOO1, "Fire PRA NotebookUncertainty and Sensitivity Analysis (UNC)". The sensitivity analyses included thepotential for a methodology change for the calculation of HEPs, circuit failure likelihoods,and common cause failure probabilities. Each set of probabilities was evaluatedindependently using the calculated 95th percentile and a 0 for each probability for eachaffected event.

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PRA RAI 26

Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall beacceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodologyfor conducting a fire PRA and endorses, with exceptions and clarifications, NEI 04-02, revision2, as providing methods acceptable to the staff for adopting a fire protection program consistentwith NFPA-805. In letter dated July 12, 2006, to NEI, the NRC established the ongoing FAQprocess where official agency positions regarding acceptable methods can be documented untilthey can be included in revisions to RG 1.205 or NEI 04-02. Methods that have not beendetermined to be acceptable by the NRC Staff require additional justification to allow the NRCStaff to complete its review of the proposed method.

Identify and describe all UAMs or deviations from NUREG/CR-6850, its supplements, andapproved FAQs, and clarify whether guidance from the June 21, 2012, memo from NRC to NEI,"Recent Fire PRA Methods review Panel Decisions and EPRI 1022993, 'Evaluation of PeakHeat Release Rates in Electrical Cabinets Fires"' was used in applying related methods. Foridentified deviations from NUREG/CR-6850 that fall outside this guidance memo, provide asensitivity study that estimates the impact of their removal on the LAR Table W-4 CDF, LERF,delta-CDF, and delta-LERF.

Response

The methods used for the development of the R.E. Ginna Fire PRA are all approved methodsas outlined by current guidance from NUREG/CR-6850, the approved FAQs, and the interimguidance provided by the NRC (e.g. circuit failure likelihoods and the Guidance Letter DatedJune 21st 2012, etc.). No unapproved methods (UAMs) were used to develop the fire PRA.Sensitivity studies are not necessary since Ginna opted to use all approved methods in itsanalysis.

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PRA RAI 31

Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall beacceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodologyfor conducting a fire PRA and endorses, with exceptions and clarifications, NEI 04-02, revision2, as providing methods acceptable to the staff for adopting a fire protection program consistentwith NFPA-805. The ASME/ANS PRA standard includes supporting requirements for seismic-fire interactions are a part of the Fire PRA review due to the inclusion of this issue in theASME/ANS PRA standard. The seismicities for central and eastern U.S. were changed as aresult of the USGS re-evaluation (USGS, "2008 NSHM Gridded Data, Peak GroundAcceleration"), based on reanalysis of the New Madrid earthquakes. Indicate if your evaluationof seismic-fire interactions includes the results of the USGS re-evaluations. Describe how theUSGS re-evaluation was addressed.

Response

The seismic-fire interactions assessment, as defined in the standard is a purely qualitativeassessment of vulnerabilities. As noted in the Ginna seismic-fire notebook, this task is focusedon identifying ignition sources that have a seismic failure mode at very low accelerations andthat are not present in the absence of a seismic event (e.g., a flammable liquid cabinet, etc),and the plant response to such fires considering the effect that the seismic event could have onthe detection and suppression capabilities. The definition of "low accelerations" is a qualitativedefinition, and is not affected by the USGS re-evaluation - the search was to identify things thatwere obviously seismically weak. Since there is no quantification involved, the assessment isindependent of return period.

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PRA RAI 32

The main control board fire scenario methods require clarification and justification ASME/ANSPRA standard with respect to NURGE/CR-6850 guidance. Please provide additionalinformation on the following. ASME/ANS PRA standard

a) The main control room analysis does not follow NUREG/CR-6850 Appendix L for themain control board (MCB) method of finding the conditional probability of damage whichaccounts for the severity factor, the non-suppression probability, and the separationbetween fire-damage targets. While the MCB analysis uses the fire ignition frequencyfrom NUREG/CR-6850, Supplement 1 (FAQ 08-0048, Section 10.2.1), the approachdiffers from that in NUREG/CR-6850 Appendix L. Describe the approach taken in lieu ofthat described in Appendix L (e.g., use of Figure L-1), and provide justification for itsapplication. Include a discussion on how it addresses the considerations which weretaken into account for Figure L-1, and its applicability to the MCB analysis

b) Describe the treatment in the MCR risk analysis of the spread of fire between panelsgiven that there is no partition between panels and how it is consistent with NUREG/CR-6850 guidance. If different, provide justification for its use.

Response

a) The justification for the assumed values for characterizing the different sequences isbased on a comparison with values obtained from the approach described in Appendix Lof NUREG/CR-6850:

The following table lists the three panel propagation impacts considered in the GinnaMCB propagation analysis. The table also compares the analysis with results that wouldbe obtained using the guidance in Appendix L of NUREG/CR-6850. The likelihoodvalues listed in the table for the Ginna MCB analysis refers to the probability ofdamaging the assigned targets given a fire starting in a panel.

Ginna 6850FPRA Appendix L

Damage Scope Likelihood LikelihoodVery Localized 0.632 8.5E-3Full Panel is Lost 0.331 5.OE-3Full Panel and Adjacent Panels are Lost 0.0369 3.5E-3

A very localized impact is considered to be, for example, a tight grouping of handswitches. This is conservatively developed based on the worst case grouping onswitches/indication on the panel. This is assumed to be a fire duration 3 minutes or lessusing the control room non-suppression probability with no severity factor applied. TheNUREG/CR-6850 Appendix L method includes a severity factor integrated with the nonsuppression probability as a function of distance. For the case of no propagation outsidethe point of ignition (i.e., a zero distance), a value of 8.5E-3 is used as the conditionallikelihood of that damage outcome given a main control board fire using NUREG/CR

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6850 Appendix L. For a full panel being lost in the Ginna FPRA, a value of 0.33 is usedas the conditional likelihood of a full panel being lost. This value is the non suppressionprobability at a time of 10 minutes. In contrast, this likelihood would be about 5.OE-3using NUREG/CR 6850 Appendix L at a distance of 0.5 m. If 10 minutes is exceeded inthe Ginna FPRA, then all adjacent panels are damaged as well at a 3.69% likelihood.Using Appendix L, the likelihood of two or three panels being lost would be about 3.5E-3given a distance of 1 m. Notice that severity factors are not included for theswitch/instrument damage evaluation. Even the smallest Ginna damage likelihoodexceeds the largest value from NUREG/CR-6850 Figure L-1.

The only time severity factors are consider in the Ginna main control room analysis is forcontrol room abandonment. This is solely used for hot gas layer development whichforces abandonment.

b) The NUREG/ CR 6850 Appendix L approach is developed for a typical main controlboard which does not include walls that separate the instrumentation and controls withinthe board. Yet, the likelihood of damage progress beyond 0.1 m per NUREG/CR 6850Appendix L Table is about 8E-3. This is far lower than the smallest Ginna conditionallikelihood given a control board fire the damage will propagate to two or more panels(i.e. 3.69%).

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PRA RAI 33

The peer review assigned Not Applicable to SRs PRM-B3, PRM-B4, PRM-B6, PRM-B8,PRM-B9, PRM-B10, and PRM-B15. Please provide the basis for this assignment.

Response

These seven PRM Supporting requirements were considered Not Applicable based on thefollowing:

PRM-B3, PRM-B4, PRM-B6: For all Capability Categories, PRM-B3 requires: "IDENTIFYany new initiating events arising from the considerations of the ES and CS technicalelements that might result from a fire event that were not included in the Internal EventsPRA including those arising from the consideration of spurious actuation."

Section 3.1 of the PRM Notebook describes the basis for the selection of TIRXTRIP, reactortrip, as the initiating event for all fire scenarios. This is justified by the structure of the faulttree logic. This includes a detailed explanation of why the fire-induced component failurescombined with the use of the TIRXTRIP initiator addresses all fire induced initiators. In thetop logic for the sequences, all potential initiating events which can cause the initiating eventcondition, either by themselves or in combination with equipment failures, are input into thelogical representation of the sequence (for example under the small LOCA top logicsequence, the initiating event for a small pipe break LOCA is included as well as the reactortrip initiator combined with equipment failure (e.g. stuck open PORV), which causes a smallLOCA). All sequences that represent equivalent sequence progressions are combined in thefault tree. Since the FRANX or CAFTA quantification applies fire impacts to equipment, thefire-induced component failures combined with the use of the TIRXTRIP initiator addressesall fire induced initiators. Similarly, the inputs into the required mitigating system failures(including support systems) include both equipment failures and the initiating events thatcould result in the failure of the mitigating system (for example, the initiating event for loss ofall CCW is included under the logic for failure of both CCW pumps). As with the initiatorportion of the fault tree, since fire impacts are tied directly to equipment failures, use of theTIRXTRIP initiator addresses all fire induced failures of equipment and no other initiators areneeded. Therefore, in evaluation of individual fire scenarios, the fire induced failure ofcomponents that are both part of an initiator and the mitigating/support systems are properlymodeled in the fault tree logic. No new initiating events or accident sequences wereidentified.

In their assessment of SRs PRM-B3, the peer review team reviewed this information alongwith the ES (as documented in G1-ES-FO01) and CS (as documented in G1-CS-FO01)tasks, and determined that no new initiating events were identified, therefore PRM-B3 wasNot Applicable.

PRM-B4 and PRM-B6 address modeling of new unique fire initiating events and accidentsequences. Because no new unique fire initiators were identified and no new event treemodels were developed, PRM-B4 and PRM-B6 do not apply.

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PRM-B8: For all Capability Categories, PRM-B7 requires: "IDENTIFY any cases where newor modified success criteria will be needed to support the Fire PRA consistently with theHLR-SC-A and HLR-SC-B of Part 2 and their supporting requirements."

Section 3.1 of the PRM Notebook addresses the applicability of Internal Event successcriteria for the Fire PRA. The HRA Task Report addresses success criteria issuesassociated with HRA timing. No new or modified success criteria were identified. In theirassessment of PRM-B7, the peer review team agreed that no new cases where new ormodified success criteria would be needed to support the fire PRA were identified. Successcriteria were as used in the internal events PRA model. PRM B8 applies only if new casesare indentified per PRM-B7. Because no new cases were identified, SR PRM-B8 is notapplicable.

PRM-B9: Model changes involving new logic or new basic events are summarized inAppendix B of the PRM Notebook. The model changes were associated primarily with MSOmodeling. PRM-B9 requires that HLR-SY-A and HLR-SY-B be followed for any cases wherenew system models or split fractions are needed, or existing models or split fractions need tobe modified to include fire-induced equipment failures, fire-specific operator actions, and/orspurious actuations. Because no new system models were required to be added in order toconstruct the fire PRA model, and existing system models and split fractions were notmodified, this SIR is not applicable.

PRM-B10: This SR applies to systems and equipment that were included in the InternalEvents PRA model but were not selected in the Equipment Selection Task of the Fire PRA.For R.E. Ginna, all systems and equipment that were included in the Internal Events PRAwere selected in the ES element (as documented in G1-ES-FO00I); therefore, this SR is notapplicable.

PRM-B15: For all Capability Categories, PRM-B14 requires: "IDENTIFY any new accidentprogressions beyond the onset of core damage that would be applicable to the Fire PRAthat were not addressed for LERF estimation in the Internal Events PRA." Section 3.5 of thePRM Notebook (G1-PRM-FO01) describes how the Level 2 PRA model fully is integratedwith the internal events Level 1 PRA. Because the Level 2 model is linked to Level 1 in thequantification, initiating events and accident sequences are the same as described for theLevel 1 PRA model. No new accident sequences beyond the onset of core damage wereidentified for the LERF estimation in the fire PRA model. In their assessment of PRM-B14,the peer review team agreed that no new accident sequences beyond the onset of coredamage were identified for the LERF estimation. PRM B15 applies only if new sequencesare indentified per PRM-B14. Because no new sequences were identified, SR PRM-B15 isnot applicable.

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PRA RAI 34

For SR SF-A5 it appears that the normal expected practice of fire-fighting personnel is creditedto meet this supporting requirement. With respect to fire brigade training procedures related toearthquakes, it was confirmed at the audit that these procedures do not exist. Normal expectedpractice of fire-fighting personnel may not be sufficient for seismic-fire scenarios. Pleaseassess the seismic-fire scenarios against the normal expected practice and discuss theconclusion on the need to develop training procedures which account for potential earthquakefire-related challenges.

Response

The requirements of SF-A5 and the Ginna Fire Brigade Training procedure were reviewed andassessed as to the extent to which training has prepared the Fire Brigade for fighting fires in thewake of an earthquake.

(a) The Fire Protection Program Report [ EPM-FPPR, Vol. I, Part II Sec. 9.0] establishes therequirements for Fire Brigade personnel, equipment, training and procedures for FireBrigade response.

The classroom instruction includes detailed review of fire fighting strategies within the FireResponse Plans (FRPs), and other related procedures. The FRPs are structured to be used

as a guideline due to the massive number of variables present in a potential fire scenario.Personnel are trained to respond as a fire-fighter would and to adapt to the situation.Outside of the classroom, training includes "live fire" training at an off-site facility wherepersonnel are subjected to building damage scenarios and multiple inputs while fighting

multiple real fires. In-plant drills also include some scenarios with complications, such as atransformer explosion. Training is conducted on the basis that building or system damage is

to be expected in a fire event without focusing on a specific cause (e.g. seismic). Ginna'sprocedures follow a symptomatic approach to best equip the Fire Brigade personnel forresponse to conditions which are not predictable.

(b) The firefighting equipment at Ginna is diverse and varied. There are hose reel stations,extinguishers, sprinkler systems, outside hydrants, a drafting station, hose houses,"appendix R" lockers at multiple locations, an on-site response vehicle and offsite firedepartment vehicles. With this type of variety, the ability to extinguish a fire in a fire event,including a seismic event, is considered achievable based on the varied locations of thisequipment. Fire brigade access routes are also able to be varied to match conditions andthe Fire Brigade is trained to be familiar with the plant configuration such that they canselect alternate routes in the event of obstructions.

(c) The potential for an earthquake to compromise one or more of the above features has beenassessed and is not determined to warrant any additional changes.

Some specific existing procedural guidance that could be applicable to a seismic scenarioincludes the following:

The Fire Response Procedure FRP-0.0, Major incident at Ginna, outlines the Fire Brigade's

Captain's Guidelines, how to utilize the appropriate Fire Response procedures, identification of

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major equipment, identification of potential hazards, fire protection features such as detection,fire barriers, specialized firefighting equipment, and Control Room/Operations Guidelines. (FRP)procedure set includes handling large fires involving large quantities of flammable liquids or oiland/or high voltage electrical equipment inside or outside the plant.

ER-SC.4, Earthquake Emergency Plan, includes actions to assess the potential damage to thefire protection yard loop and various block walls. Fire is not specifically addressed in thisprocedure since other procedures provide that guidance.

Response to a large area fire such as might result from a seismic event that results in a totalloss of AC power is outlined in ECA-0.0. Procedural guidance is included in procedure ER-D/G.1 for a loss of AC and DC power.

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PRA RAI 35

Fire-induced instrument failure should be addressed in the HRA per NUREG/CR-6850 andNUREG-1921, and instrumentation credited for operator actions in the HRA should be verifiedto be available. Therefore, please address the following:

a) Describe how fire-induced instrument failure (including no readings, off-scale readings,and incorrect/misleading readings) is addressed in the fire HRA. Include discussion ofinstrumentation that was modeled explicitly in the fault trees, the success criteriaassumed for this modeling, and how explicit modeling of instrumentation was done in theevaluation of HEPs.

b) Confirm that instrumentation credited in the HRA been verified to be available for the firescenarios in which they are credited.

Response

a) The Ginna alarm response procedures (ARMs) and Emergency Operating Procedures(EOPs) were evaluated through the Equipment Selection task of the Fire PRA [G1-ES-F001] to determine where it is necessary to model instrumentation required to supportoperator actions during a fire. In conjunction with this review, the Fire HRA task identifiedHFEs where the reliability of operator diagnosis could be affected by degradedinstrumentation. The potential for instrumentation to initiate undesired operator actionwas also considered.

Alarm Response Procedures were reviewed to identify new, undesired operator actionsthat could result from spurious illumination of a main control board (MCB) annunciator(e.g., due to verbatim compliance with the instruction in an alarm response procedure,when separate confirmation is not available or required).

Each of the human actions credited in the fire PRA were reviewed to identify any newprocedural flow diversions that could result from fire induced erroneous indication (e.g.operations starts bleed and feed due to a false low SG water level which diverts fromAFW alignment). Therefore, the Emergency Operating Procedures (EOPs) werereviewed in the context of the credited Human Failure Events (HFEs).

The Appendix R Safe Shutdown Equipment List (SSEL) was also reviewed to evaluatethe instrumentation and cues and their potential operator action impacts.

For each HFE, an initial set of indication failures which would degrade the ability ofoperators to successfully complete the action (but not guarantee failure of the action)was determined. This was based on the specific procedure steps required to completethe HFE, as well as general knowledge of indication available in the control room foroperators to use. For example, some procedure steps may explicitly state whichindicator to use, while others may only refer to the parameter required (e.g., checkpressurize pressure, etc.). The nature of the degraded instrumentation performance wasspecified, such as reading off-scale high, reading zero, indicating red, or one of two flowindicators reading >X gpm. A second initial set of indication failures which wouldguarantee failure of the action was developed in a similar fashion. These initial indicationsets were developed by the PRA group. Once the initial 'degraded' and 'failed' indication

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sets were identified, interviews were performed with operations personnel to confirm theinitial indication sets, or to determine more appropriate indication sets if the original setswere not accurate.

This detailed instrumentation assessment, logged as PRA Evaluation Request (PRAER)No. G1-2010-001 in the Constellation document tracking system, is included in theGinna Fire HRA Notebook [-HRA-FO01] as Appendix C. Attachment 1 of the PRAERdocuments the detailed review of all relevant procedures and instrumentation for eachHFE. Table 4 presents a listing of the HFEs identified in the PRAER as requiringinstrumentation along with the relevant instrumentation.

The following three cases of effects of instruments on operator reliability wereconsidered in the Fire HRA to address the range of fire impacts:

1. Action is normal because cues are available2. Action reliability is reduced by degraded cues from instruments caused by fire3. Action is completely failed because of no cues

For case 1, no adjustments were made to the HFE on the basis of cue degradation(although other adjustments for fire-related impacts may have been made to timing, forexample).

For case 3, an HFE was not modeled and the instrumentation cues were hardwired intothe fault tree logic as leading directly to functional failure. For example, when power islost, the power range neutron flux indicators fail low. This can cause operations tobelieve that the rods have inserted if the rod display screen is dark.

Most of the HFEs credited in the model fall under case 2 and are associated with someform of indication (and any needed power inputs) in an "OR" relationship in the fault tree.The logic for 'degraded' human actions was combined with a new 'degraded' HRA eventwhose value was calculated in the HRA Calculator. Most of these instrumentation-impacted HFEs were variants of existing internal events that were modified to reflectdegraded cues to the operator and designated with a suffix of -Dl. (The exceptions arethose actions that do not contain a diagnosis component and are direct proceduralsteps. In those instances, the associated but separate HFE for the diagnosis actionwould include any indication dependencies.)

The degraded instrumentation modifications to existing HFEs were made in the HRACalculator using the guidance specified in Ginna PRA Assumption 85 (included in theGinna Fire HRA Notebook as Figure 4-7):

The general impact to the HFEs is to increase the delay time, increase the medianresponse time, and increase the cognitive recovery dependency. (Note: No increase inexecution dependency, since once operators determine the need to perform an action,there should be no impact on the execution)

In addition, where the analysis determined that cues were inaccurate due to fire, theCBDTM Cognitive Unrecovered trees Pc-a and Pc-d branch selections were adjusted toreflect this.

PRA RAI 352

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b) For each operator action in the fire PRA, the required indication was directly modeled inthe fault tree in the same fashion as all the other credited equipment. As with anycredited equipment, the instrumentation appears in both the equipment selectionnotebook and the cable selection notebook. Any fire scenario that degraded instrumentcables or tubing for related indications either failed the operator action or degraded theoperator action as described in Section A of this RAI response.

Therefore, any required instrumentation associated with a credited operator action hasbeen verified to be available for the fire scenarios in which the operator action iscredited.

PRA RAI 353

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PRA RAI 37

SRs in the ASME/ANS RA-Sa-2009 standard include updating ignition frequencies if outliers arefound (SR IGN-A4) using a Bayesian update process or equivalent statistical process (SRIGNA6). The fire ignition frequencies are generic and Bayesian updating was not performed.Please provide justification for not performing Bayesian updating with plant-specific data.

Response

The evaluation of the applicability of generic frequencies for the Ginna Fire PRA was performed

and documented in Sections 2.2 and 2.3 of the "Fire PRA Notebook Ignition Frequency (IGN)"

G1-IGN-F001, Revision 2. The results of that evaluation are summarized and complemented

with additional information below.

Fire events specific to the R.E. Ginna nuclear plant were collected from condition reports (CRs)and the Ginna Fire Legacy Action Reports (ARs). The records were collected starting in year2000. Fire events prior to the year 2000 belong to the time period used in EPRI's Fire EventsDatabase to calculate the generic ignition frequencies, and as such are not included in thisupdate to avoid double counting. The events were obtained by searching the R.E. Ginna ARand CR databases. Using keywords: "fire," "fires," "smoke," "explosion," and "ignition," thedatabases were queried from 01/01/2000 to 03/29/2012, then manually screened to actual fires,using the criteria given in Section C.3.2 of NUREG/CR-6850, which include three categories:"potentially challenging", "not challenging", and "undetermined". The evaluation found a total of8 plant-specific events that were categorized as either "potentially challenging" or"undetermined".

Of the 8 events, 3 of them were found to belong to Bin 15- Plant-Wide Electrical Cabinets,defined in Chapter 6 of NUREG/CR-6850. Section 6.5.2 of that NUREG indicates that if multiplefire events fall into the same bin, the analyst should investigate the severity of those events andattempt to identify any common causes, because a common cause may indicate a problemspecific to the plant, which may warrant a plant-specific evaluation. The NUREG furtherindicates that if no commonality can be identified as the root causes of the fires, it may be anindication of the large variations in reporting practices throughout the commercial nuclearindustry, rather than a problem in the plant analyzed. A review of the severity of the three eventsshowed that two of them were classified as "Undetermined" as to whether they werechallenging. Further, no common cause was identified between the three. Accordingly, it wasdeemed that the three events did not point to an underlying problem specific to the R.E. Ginnapower plant, such that no plant-specific data update was considered necessary for that bin.

2 of the events were found to belong to Bin 14, Plant-Wide Electric motors. In this case, nocommon cause between the two fires was identified, and neither fire impacted componentsother than the motor itself. Again, no plant-specific data update was considered necessary.

The other 3 fires were all in separate bins, and use of generic data was found to be appropriate.

Based on the above discussion, the generic frequencies were determined to be appropriate for

the R.E. Ginna Fire PRA.

PRA RAI 371

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PRA RAI 39

Discuss your plans for modeling the charging pump motor vari-drive replacement, including thelocation of the variable drives, in the Fire PRA, and whether or not this change is part ofImplementation Item 9 in the LAR, Attachment S, Table S-3, "Implementation Items."

Response

The charging pump vari-drive replacement is a modification that finished after the NFPA805 firemodel was developed. As such, as part of our long term configuration management programthis modification will be incorporated into our risk model. This modification was not credited inthe LAR submitted model. The modification is part of Implementation Item 9 in the LAR,Attachment S, Table S-3, "Implementation Items." All changes affecting the design, operation,or maintenance of the plant are performed in accordance with fleet procedure CNG-CM-1.01-1003, which has been updated to reflect NFPA 805 requirements, as described in PRA RAI 13.

PRA RAI 391

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PRA RAI 43

Table W-4, "Ginna Fire Area Risk Summary," of the LAR provides the risks associated with theVFDRs. This table reports LERF and A LERF values for a number of fire areas such asEDG1A, EDGIB, OFFSIE, PA, SAF, SH, STA13ACH, and YARD to be "0.OOE+00." As thesefire areas contribute to CDF, it not is clear how these areas can have zero contribution to LERFand delta-LERF. Confirm that none of the fire-induced core damage sequences in these fireareas lead to large early release sequences (justify the zero value contributions), provide theactual calculated values for these fire areas, or explain what these values are meant torepresent.

Response

The LERF quantification approach was the same as the CDF quantification approach with thetruncation lowered by a factor of ten. In certain cases where the LERF values were very small,the truncation level would preclude the production of cutsets; therefore, showing a LERF valueof zero. This result is a function of the software limitation and actually indicates that the resultsare lower than the attempted truncation value. These zero values are considered negligible.The zero values will be replaced with epsilon, which will indicate that the values are negligibleinstead of zero. Ep~ilon will be defined appropriately in the footnotes of Table W-4 to indicatethat the value is less than a particular threshold for CDF and LERF that indicates the values arenegligible.

PRA RAI 431

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Programmatic RAI 01

Based on the NRC staffs review of the LAR and during the subsequent audit, it was determinedthat the licensee did not adequately describe the effects of NFPA 805 on the configuration andchange control processes.

Describe how the various configuration control and change control procedures are implementedtogether to ensure compliance with the NFPA 805 change evaluation and configuration controlrequirements.

Response

Changes affecting the design, operation, or maintenance of the plant are performed inaccordance with procedure CNG-CM-1.01-1003, "Design Engineering and ConfigurationControl." CNG-CM-1.01-1003 ensures that reviews are performed to determine if plant changesimpact the fire protection program documentation. This procedure has been revised to reflectthe requirements of NFPA 805. Engineering standard CNG-FES-007, "Preparation of DesignInput and Change Impact Screen," which compliments CNG-CM-1.01-1003, has also beenrevised to add a section for NFPA 805 applicability criteria and required actions.

Programmatic RAI 011

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Programmatic RAI 02

NFPA 805, Section 2.7.3.4, "Qualification of Users", states that cognizant personnel who use

and apply engineering analysis and numerical models (e.g., fire modeling techniques) shall becompetent in that field and experienced in the application of these methods as they relate tonuclear power plants, nuclear power plant fire protection, and power plant operations.Describe how the training program will be revised to support the NFPA 805 change evaluationprocess, including positions that will be trained and how the training will be implemented (e.g.,classroom, computer-based, reading program).

Response

Training Review Requests will be initiated, and Training Change Orders initiated, followingdetermination of a need for training by the Curriculum Committee. A Training Needs Analysiswill be completed to determine training that will need to be conducted to support implementationof the NFPA 805-based Fire Protection Program. This will include determining the variouslevels of training that will be conducted and the recipients of such training, as well as how thetraining will be implemented. Procedure CNG-TR-1.01-1014, "Engineering Support PersonnelTraining Program," is the governing procedure to maintain and enhance the EngineeringSupport Personnel (ESP) training and qualification program. CNG-TR-1.01-1014 has beenrevised to list the qualifications for NFPA 805, and changes have been made to the ESPprogram to include the qualification of cognizant individuals.

The design inputs and change impacts screening process described in CNG-FES-007 has beenrevised, and components have been flagged in our controlled document management system

as NFPA 805 components. Procedure CNG-TR-1.01-1014, "Engineering Support PersonnelTraining Program," ensures that cognizant personnel who use and apply CNG-CM-1.01-1003are competent in the application of methods for reviewing the impact of changes to fireprotection program documentation.

Current CENG Design and PRA staff members are required to maintain qualification cards toensure these personnel have the appropriate training and technical expertise to performassigned work, including the use of engineering analyses and numerical models.

Ginna will maintain qualification requirements for individuals assigned to perform NFPA 805related tasks. Position Specific Guides will be developed to identify and document required

training and mentoring to ensure cognizant individuals are appropriately qualified to performassigned work per the requirements of NFPA 805, Section 2.7.3.4. Qualification requirementsare contained in procedure CNG-TR-1.01-1014.


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Programmatic RAI 03

Based on the NRC staffs review of the LAR and during the subsequent audit, it was determinedthat the licensee did not adequately describe the integration of combustible loading controls andFire Probabilistic Risk Assessment (FPRA) requirements.

Describe how the combustible loading program will be administered to ensure that FPRAassumptions regarding combustible loading are met.

Response

As part of the NFPA 805 transition, R.E. Ginna will develop an integrated combustible loadingprogram to include insights from the Fire PRA. R.E. Ginna Fire Protection and Fire PRAengineers will define combustible control levels to ensure practical and effective implementationthat considers relevant risk insights. Risk insights will include considerations for fullcompartment burn fire areas, transient influence factors developed in the Fire PRA, and firescenarios impacting structural steel elements. The objective of the integrated combustibleloading program is to provide plant procedures with guidance for transient combustible controlsconsistent with the requirements of the NFPA 805 fire protection program.


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SSA RAI 01

NFPA 805, Section 1.3.1 states the nuclear safety goal that a fire during any operational mode

and plant configuration will not prevent the plant from achieving and maintaining the fuel in a

safe and stable condition. LAR section 4.2.1.2 states that safe and stable conditions can bemaintained indefinitely until a decision is made to transition to residual heat removal (RHR)cooling. The LAR summarizes the means to maintain safe and stable conditions for extended

periods of time, including inventory control, decay heat removal, electrical systems, and dieselfuel supplies. Provide additional discussion of the actions necessary beyond 24-hours to meetthe specific nuclear safety performance criteria and maintain safe and stable conditions.Discuss the risks associated with accomplishing these actions.

Response

The following is background information that forms the bases for the specific responses forNFPA-805 Performance Criteria (a) through (c).

The specific capabilities that will be required to meet the performance criteria beyond 24 hoursinclude the availability of procedures and personnel to perform the necessary repair/recovery of

equipment needed to maintain safe and stable conditions for the extended period of time. Toaccomplish this goal, existing Emergency Operating Procedures (EOPs) and other Emergency

Response Organization (ERO) procedures are currently in place to assist the plant operatingstaff with options to proceed and implement such actions and/or repairs.

Following the initial establishment of safe and stable conditions, the ability to control reactor

pressure, inventory, and temperature requires limited operator involvement as the actions arecharacterized by simple manipulations of valves and/or pump controls, and processinstrumentation is readily available at appropriate locations. LAR Attachment C, "NEI 04-02

Table B-3 - Fire Area Transition", lists a "Method of Accomplishment" for each performance

goal. These methods can be maintained beyond 24 hours.

A fire affecting plant equipment will result in activation of the Emergency Response

Organization (ERO). This activation results in staffing the ERO facilities within one hour.Therefore, the Technical Support Center (TSC), Operational Support Center (OSC) and

Emergency Operations Facility (EOF) staff would be in place to provide additional expertise andresources to address plant issues. Procedures exist to ensure adequate staffing of the EOF,TSC, OSC, and Operations Shifts for indefinite periods of time. The OSC would provide overall

coordination of repair and corrective actions, as directed by TSC personnel. The TSC,Operations Shifts, and Fire Brigade Captain would review existing Fire Response Procedures(FRPs) for the affected fire area(s) for the impact of the fire on affected plant equipment.

Frequent drills and exercises are conducted with the Emergency Response Organization toevaluate and maintain these capabilities.

SSA RAI 011

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While the "Method of Accomplishment" (from LAR Attachment C) can be sustained for more

than 24 hours, it is more likely that restoration of offsite power to 480VAC busses, reliable 125VDC power, operable redundant equipment, and available power to equipment, will beaccomplished within that time frame per TSC/EOF/OSC repair actions. If it is determined thatInventory and Pressure Control or Decay Heat Removal can be better accomplished usingsafety injection and/or residual heat removal systems, then cooldown and lowering of reactorcoolant system (RCS) temperature and pressure are options that would be available.

The following modifications are referred to in Table S-2 (Plant Modifications Committed) ofAttachment S of the Ginna Transition Report (LAR), and may be utilized to maintain safe andstable conditions:

" Plant Modification ESR-12-0144 will install a new standby charging pump in the standby

auxiliary feedwater (SBAFW) building with its dedicated support systems, including a10,000 gallon charging tank. A new batching station will also be installed in the SBAFWbuilding to supply borated water once the 10,000 gallon tank is depleted.

" Plant Modification ESR-1 1-0050 will provide a water source that will be free of theeffects of fires outside the SBAFW building. A new 160,000 gallon Condensate StorageTank will be installed, and will be able to supply the SBAFW pumps for a minimum of 12hours with no operator action, longer if operator action is taken. ESR 11-0050 will alsoprovide a new 1000KW diesel generator that is capable of powering 2 charging pumps,

one SBAFW pump, and a battery charger.

" Plant Modification ESR-12-0143 will provide a second new 1000KW diesel generatorthat is also capable of powering 2 charging pumps, one SBAFW pump, and a battery

charger.

Actions necessary beyond 24 hours to meet specific nuclear safebj performance criteria andmaintain safe and stable conditions, in general, include options such as:

" Alternate sources of inventory to provide assurance of adequate supply of RCSmakeup water, auxiliary feedwater and diesel fuel oil, for Inventory and PressureControl, Decay Heat Removal and maintaining Vital Auxiliaries

" Alternate pumps to provide assurance of adequate flow of RCS makeup water,

auxiliary feedwater and diesel fuel oil, for Inventory and Pressure Control, DecayHeat Removal and maintaining Vital Auxiliaries

* Alternate power sources to provide assurance of adequate electrical power to

needed or redundant equipment

* Resources from the Emergency Response Organization (ERO) to develop

contingency actions as needed to ensure that alternate means are available tomaintain nuclear safety performance criteria

The risks associated with the required actions are discussed at the end of this response.

SSA RAI 012

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REACTIVITY CONTROLAdequate negative reactivity insertion will have been established in the first 24 hours.Therefore, continued CVCS injection of borated water, as discussed under "Inventory andPressure Control" below, will continue to provide the required negative reactivity to maintain

subcritical conditions.


INVENTORY AND PRESSURE CONTROLActions that would be necessary after 24 hours to maintain Inventory and Pressure Control mayinclude any or all of the following actions:

* Minimize loss of reactor coolant by continuing to identify and isolate sources of RCSleakage. TSC/OSC personnel would evaluate and provide this, assisted by on-shift

operators.* Provide alternate means of primary system injection, utilizing available power

supplies and available pumps. TSC/OSC personnel would evaluate and providealternate means of RCS injection if a normal power supply is not available to anavailable charging pump.

* Additional supplies of boron are maintained on-site, and would be used perprocedure S-11 or to replenish the new 10,000 gallon Charging Tank (to be installed

per ESR-12-0144), as noted below.

* Provide alternate sources for primary system makeup:o Replenish refueling water storage tank (RWST). Plant procedures, including

the following, already provide this direction:

* ER-RWST.1, "Alternate RWST Makeup"* S-11, "Batching Tank"* S-9J, "Blending to RWST"* EOP ATT-18.0, "Attachment SFP-RWST"

o Utilize CVCS Hold-up Tanks (HUTs). Plant procedure S-3.2D (TransferringWater from CVCS HUTs to RWST or SFP) provides this direction, andseveral other plant procedures could be used to send water to the CVCSHUTs from Monitor Tanks, reactor makeup water (RMW) tank, etc.

o Replenish new 10,000 gallon Charging Tank with borated water from newbatching station in the standby auxiliary feedwater (SBAFW) Building.


DECAY HEAT REMOVALActions that would be necessary after 24 hours to maintain Decay Heat Removal may includeany or all of the following actions:

SSA RAI 013

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* Maintain Steam Relief capability. Maintaining the plant at hot shutdown is controlled byplant procedure 0-2.1 (Normal Shutdown to Hot Shutdown).

o Steam Generator (SG) Atmospheric Relief Valves (ARVs) would be the preferredpath for steam relief. Instrument air is the normal motive force for operation ofthe ARVs.

o If instrument air has not been restored within 24 hours, installed backup nitrogensupply systems are used to control the ARVs. The nitrogen supply systems arecapable of controlling the ARVs for eight hours without requiring a bottle change.Plant procedure P-15.20 (Change Nitrogen Bottles for Atmospheric ReliefValves) provides direction for the routine replacement of expended bottles asneeded. Spare replacement nitrogen bottles are stored on-site.

o If ARVs are not available, automatic operation of main steam safety valves(MSSVs) on increasing pressure will continue to be required

" Maintain Spent Fuel Pool (SFP) cooling to ensure SFP temperature remains less than180°F. Note that, for a complete and extended loss of SFP cooling, SFP "Time to Boil"is typically on the order of several days, and has not been calculated as less than 24hours except during full-core off-loads. Options for restoration of SFP cooling include:

o ER-CCW.1, "Fire Water Cooling to CCW and A SFP Heat Exchangers"o EOP ATT-30.0, "Attachment SFP Cooling Restoration"o ER-SFP.1, "Loss of Spent Fuel Pool Cooling"

" Provide alternate means of auxiliary feedwater (AFW) flow to SGs, utilizing availablepower supplies and available pumps. TSC/OSC personnel would evaluate thealternatives, and ensure the availability of whichever AFW pump (SBAFW pump "C" or"D", turbine-driven auxiliary feedwater (TDAFW) pump, or motor-driven auxiliaryfeedwater (MDAFW) pump "A" or "B") would be used to feed the SG. Note that PlantModifications ESR-11-0050 and ESR-12-0143 will provide additional 1000KW dieselgenerators, each of which is capable of powering 2 charging pumps, one SBAFW pump,and a battery charger.

" Provide additional water inventory to supply AFW pumps:o Operate the GE Betz Water Treatment System to maintain a supply of

secondary-grade makeup water to the Condensate Storage tanks (CSTs). IfBusses 13 or 15 are not available and cannot be made available, power can besupplied to the GE Betz system from an independent offsite power source(Sodus Circuit 5241 to AC power Panel "ACPDPWW36"). Plant procedure T-7.3(GE Betz Water Treatment System Operations) provides this direction.

o Transfer water from the new 160,000 gallon Condensate Storage Tank (to beinstalled per ESR-11-0050).

* Provide alternate sources of water. (Note that Lake Ontario is an inexhaustible watersource.) Refer to procedure ER-AFW.1 (Alternate Water Supply to the AFW Pumps),which lists several options, including:

o Transferring water from the 100,000 gallon Outside Condensate Storage Tank(OCST) to the normal CSTs.

o Align City Water to SBAFW pump suction

SSA RAI 014

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o Align Service Water to AFW pump suctiono Connect plant Fire Water system and fire water pump to fill CSTso Connect City Water fire hydrant to fill CSTs


VITAL AUXILIARIES

A - ELECTRICAL SYSTEMSActions that would be necessary after 24 hours to maintain and/or provide alternate electricalsystems to support safe and stable plant conditions may include any or all of the followingactions:

* Maintain and/or restore offsite power to 480 VAC busses, 120 VAC instrument loads,and 125 VDC loads

o If offsite power needs to be restored, the EOF would obtain necessary resourcesto perform any actions needed. ERO procedure EPIP-1-18 (DiscretionaryActions for Emergency Conditions) identifies sources for alternate AC and DCpower supplies and alternate means of providing power.

" Maintain reliable power (or backup power) from the emergency diesel generators(EDGs) to 480 VAC busses. This would involve ensuring adequate diesel fuel suppliesand maintaining EDG support component capability.

" When AC power is available from offsite power and/or the EDGs, DC power is alsoavailable, since the DC busses are normally powered by battery chargers supplied byAC power.

* Restore AC and DC power using existing plant procedures. Other options for restoringpower include:

o ER-ELEC.4, "TSC DIG Feed to BUS 16 to Supply Charging Pumps, InstrumentBus D, and Battery B"

o ER-ELEC.3, "Emergency Offsite Backfeed Via Main and Unit Transformers"o ER-FIRE.1, Attachment 8, "Long Term DC Power Supply"o EOP ATT-24.0, "Attachment Transfer Battery to TSC"

" Evaluate the adequacy and reliability of 480 VAC and 125 VDC power to neededequipment. TSC/OSC personnel could implement actions necessary to re-powerneeded equipment from alternate power sources, as discussed in the Inventory Controland Decay Heat Removal sections above. Such actions could involve field layout andinstallation of electrical cables from available (energized) breaker cubicles directly toneeded equipment. The capability to identify unique solutions to re-power equipment isevaluated during ERO drills and exercises. Dedicated electrical cables for selected re-powering are identified in plant procedures and stored for use when needed. Examplesinclude:

o ER-FIRE.1, Attachment 8, "Long Term DC Power Supply"o ER-FIRE.3, Attachment 11, "Power Restoration to SAFW"

SSA RAI 015

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B - EMERGENCY DIESEL GENERATORS AND DIESEL FUEL SUPPLIESActions that would be necessary after 24 hours to maintain diesel operation, diesel cooling, and

maintain an adequate supply of diesel fuel oil to ensure long-term emergency diesel generator(EDG) operation to support safe and stable plant conditions may include any or all of the

following actions:" Continue to review plant procedure T-27.4 (Diesel Generator Operation). T-27.4 is

performed whenever an EDG is operating.* Ensure adequate cooling to the EDGs. Refer to plant procedure ER-D/G.2 (Alternate

Cooling for Emergency D/Gs), which lists several options, including:

o Aligning Alternate Cooling from City Water to a EDGo Establishing D/G Alternate Cooling using B5b pump

" Plant procedure ER-D/G.1 (Restoring D/Gs) provides options for supplying fuel to theEDGs:

o Each Day Tank can be filled from the opposite train Fuel Oil Transfer Pump orfrom an external source

o Each Fuel Oil Transfer Pump can be aligned to the opposite train on-site Fuel OilStorage Tank

* Plant procedure T-27.3 (Fuel Oil Transfer To Emergency / Fire / Portable Air

Compressor /6188 Portable Tank/ Diesels) provides direction for receipt of fuel from off-site suppliers or from the two buried tanks that are located outside the protected area butwithin the Owner-Controlled Area (OCA)

" Trucking facilities exist in the Rochester area to ensure that, within 8 hours, fuel oil can

be delivered to the two fuel oil storage tanks located on-site (within the protected area)or to the two buried fuel tanks located within the OCA. Ordering and accepting deliveryof fuel is a routine activity that can be anticipated in ample time to ensure its continuousavailability long term.

C - INSTRUMENT AIRActions that would be necessary after 24 hours to ensure Instrument Air is available to support

safe and stable plant conditions may include any or all of the following actions:

" Verify the integrity of instrument air piping

• Identify and repair/replace air piping and/or tubing, as needed.

• Provide 480 VAC power to Bus 13 and/or Bus 15 to power an Instrument AirCompressor

* Provide alternate supplies for Instrument and/or Service Air. Refer to procedure T-2F

(Backup Air Supply), which lists several options:o Supply Service Air from the Breathing Air Compressoro Supply Service Air from Diesel Driven Air Compressor

o Supply Instrument Air From The Service Air System Via Diesel Air Compressor

SSA RAI 016

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D - OTHER EQUIPMENT/SYSTEMSActions that would be necessary after 24 hours to ensure other equipment/systems areavailable to support safe and stable plant conditions may include any or all of the following

actions:

" Monitor operating equipment using normal operator plant surveillance, procedural

controls and enhanced observation of key components

" Ensure necessary support systems are available as needed. EOF/TSC/OSC wouldprioritize and restore as appropriate or feasible.


PROCESS MONITORINGActions that would be necessary after 24 hours to achieve and maintain the necessaryindication may include any or all of the following actions:

* Ensure the actions listed for "Vital Auxiliaries" are successful in restoring andmaintaining reliable offsite power to 120 VAC instrument loads and 125 VDC loads.

" Ensure the actions listed for "Vital Auxiliaries" are successful in restoring and

maintaining reliable on-site backup power from the EDGs.

" Continue to monitor and to repair/restore, as necessary, parameters and parameter

redundancy.

RISKS ASSOCIATED WITH THESE ACTIONSGinna qualitatively evaluated the actions and activities required to maintain these conditions,and concluded that the risk impact of the failure of actions to maintain safe and stable conditions

beyond 24 hours is deemed to be very low.

" Each of the functions required to maintain safe and stable conditions (Reactivity Control,

Inventory and Pressure Control, Decay Heat Removal, Vital Auxiliaries, ProcessMonitoring) has multiple success paths.

" There is a long time available to establish alternative long-term configurations for

equipment and power supplies, and long periods of time before depletion of commoditiessuch as fuel oil and nitrogen become concerns. Replenishing of these commodities ispart of the ERO procedures and are routine actions.

" Shift staffing requirements are adequate to provide all of the operators required to

perform actions. ERO facilities would be staffed continuously. The availability of thesesupplemental resources to perform any of these actions or activities further ensure thatthese longer-term actions will be reliably accomplished.

* Existing plant procedures provide direction for many of the actions that would be

completed* ERO processes ensure that any actions directed from the TSC will be controlled by the

OSC, using procedures EPIP-1-10 (Operational Support Center (OSC) Activation), and

SSA RAI 017

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EPIP-1 -12 (Control of Emergency Maintenance Assessment and Repair Teams). Theseprocesses ensure that extensive planning and pre-job briefs will be conducted before

commencement of any field activities.When new equipment is installed per the ESRs listed above, procedures will bedeveloped to provide direction for appropriate actions to operate this equipment.

SSA RAI 018

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SSA RAI 02

Provide the following pertaining to non-power operations (NPO) discussions provided in Section4.3 and Attachment D of the LAR:

a) LAR Section 4.3.2 states that based on incorporation of the recommendations from thepinch point evaluations into appropriate plant procedures, the performance goals fornon-power operations will be fulfilled and the requirements of NFPA 805 will be met. Ata high level, identify and describe the changes to outage management procedures, riskmanagement tools, and any other document resulting from incorporation of Key SafetyFunctions (KSF) identified as part of NFPA 805 transition. Include changes to anyadministrative procedures such as "Control of Combustibles".

b) For those components which had not previously been analyzed in support of the at-power analysis or whose functional requirements may have been different for the non-power analysis, provide a list of the additional components and a list of those at-powercomponents that have a different functional requirement for NPO. Describe thedifference between the at-power safe shutdown function and the NPO function. Includewith this list a general description by system indicating why components would beselected for NPO and not be included in the at-power analysis.

c) Provide a list of KSF pinch points by fire area that were identified in the NPO fire areareviews including a summary level identification of unavailable paths in each fire area.Describe how these locations will be identified to the plant staff for implementation.

d) During NPO modes, spurious actuation of valves can have a significant impact on theability to maintain decay heat removal and inventory control. Provide a description ofany actions being credited to minimize the impact of fire-induced spurious actuations onpower operated valves (e.g., Air Operated Valves (AOVs) and Motor Operated Valves(MOVs) during NPO (e.g., pre-fire rack-out, actuation of pinning valves, and isolation ofair supplies).

e) During normal outage evolutions certain NPO credited equipment will have to beremoved from service. Describe the types of compensatory actions that will be usedduring such equipment down-time.

f) The description of the NPO review for the LAR does not identify locations where KSFsare achieved via recovery actions or for which instrumentation not already included inthe at-power analysis is needed to support recovery actions required to meet the KSFs.Identify those recovery actions and instrumentation relied upon in NPO and describehow recovery action feasibility is evaluated. Include in the description whether thesevariables have been or will be factored into operator procedures supporting theseactions.

SSA RAI 021

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Response

Part a

Ginna will ensure that outage management procedures, risk management tools, and otherprocedures that will incorporate Key Safety Functions (KSFs) are identified and changed asappropriate.

The following procedures currently implement shutdown risk and the essential work planningand implementing process. These and other procedures will be reviewed and revised asnecessary to implement these changes and requirements to incorporate guidance from the NPOreview. Procedures to be considered for revision include:

" IP-OUT-2, OUTAGE MANAGEMENT* IP-OPS-3, CONDUCT OF LOWER MODE OPERATIONS• A-3.1, CONTAINMENT STORAGE AND CLOSEOUT INSPECTION* A-54.7, FIRE PROTECTION TOUR* A-202, THE FIRE PROTECTION PROGRAM AND GINNA STATION STAFF

RESPONSIBILITIES FOR FIRE PROTECTION" A-601.13, FIRE PROTECTION / APPENDIX R COMPENSATORY ACTIONS" A-601.14, APPENDIX R PROGRAM CONTROL" FPS-16, BULK STORAGE OF COMBUSTIBLE MATERIALS AND TRANSIENT FIRE

LOADS" CNG-CM-1.01-3004, PRA PROCESS FOR INTERNAL EVALUATIONS" CNG-MN-4.01-1001, WORK ORDER EXECUTION AND CLOSURE PROCESS" CNG-MN-4.01-1002, WORK ORDER INITIATION,SCREENING AND PRIORITIZATION" CNG-MN-4.01-1003, WORK ORDER PLANNING" CNG-OM-1.01-1000, OUTAGE MANAGEMENT" CNG-OM-1.01-1001, SHUTDOWN SAFETY MANAGEMENT PROGRAM" CNG-OP-4.01-1000, INTEGRATED RISK MANAGEMENT

In preparing the revisions, Ginna will consider the need to include direction to minimize transientcombustibles, evaluate the need for fire tours, evaluate control of ignition sources, and considerother preventive measures throughout the plant during NPO, especially in the areas identifiedas having pinch points.

Part b

The NPO Modes Analysis identified systems used for accomplishment of required KSFs andgrouped those components making up success paths into function codes. Because they werenot credited in the at-power analysis, cable selection was not originally performed for the 153components listed in Table SSA Q2-1. (The majority of equipment required to maintain theNPO KSFs is the same as that required to safely shutdown the plant while at power.)

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Table SSA Q2-1 lists the additional 153 components that had not been analyzed in support ofthe at-power analysis, but have had cable selection completed and are addressed in the NPOanalysis.

Table SSA Q2-1: NPO Components Not in SSD

NPO NPOComponent ID SYS TRAIN Component Description Normal Required

Position Position

ill cvC A RMW TO BA BLENDER FLOW CONTROL VLV HCV-111 C C

133 RHR 3 RHR TO CVCS LETDOWN 0 C

135 cvC 3 LOW PRESS LTDN PRESS CONTROL VLV PCV-135 0 0

252 cvc N/A CVCS LETDOWN TO RHR IV 0 0

253 cvc N/A CVCS LETDOWN TO CHARGING PUMP SUCTION IV 0 C

261 CVC N/A VCT H2 INLT MANUAL BLK VLV 0 C

262 CVC N/A VCT NITROGEN INLET MANUAL BLK VLV 0 C

350 CVC B EMERGENCY BA SUPPLY VLV MOV C 0

358 CVC N/A RWST MAKEUP AOV BYPASS VALVE C 0

624 RHR 3 RHR HX B OUTLET 0 0

625 RHR 3 RHR HX A OUTLET 0 0

626 RHR 3 RHR HX BYPASS C C

782 SEP N/A LOW SUCTION ISOL VLV TO SPENT FUEL POOL RECIRC C 0PUMPS (ALT)

808 SIS N/A RWST TO RFW PURIFICATION PUMP IV C 0

819 FPC N/A REF WTR PURIF PUMP TO REGEN HX IV C 0

821 FPC N/A REF WTR PURIF PUMP TO REGEN HX IV C 0

1721 WDS A RCDT OUTLET ISOL VALVE AOV-1 721 C C

1726 WDS N/A RCDT PMP A DISCH ISOL VLV 0 C

1727 WDS N/A RCDT PMP B DISCH ISOL VLV 0 C

8667 SEP N/A ISOLATION GATE VALVE FROM SFP PUMP B TO SFP C 0HEAT EXCHANGER B

1003A WDS A RCDT OUTLET ISOL VALVE AOV-1003A C C

1003B WDS B RCDT OUTLET ISOL VALVE AOV-1003B C C

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Position Position

110A CVC A BA TO BA BLENDER FLOW CONTROL VALVE HCV-1 1 OA O,C 0

110B CVC B RMW FLOW CONTROL VLV AOV-110B C 0

13A OAC N/A 115 KV SWITCHYARD STATION 13A E E

1811A WDS N/A RCDT PUMPS DISCHARGE TO RHR HX B C 0

1811B WDS N/A RCDT PUMPS DISCHARGE TO RHR HX A C 0

1813A RHR A REACTOR COOLANT DRAIN TANK PUMP SUCTION FROM C OCCONTAINMENT SUM

1813B RHR B REACTOR COOLANT DRAIN TANK PUMP SUCTION FROM C OCCONTAINMENT SUM

1815A SIS A SAFETY INJECTION PUMP C SUCTION VALVE 0 0

1815B SIS B SAFETY INJECTION PUMP C SUCTION VALVE 0 0

25A/11T-12A- MAC A TRANSFORMER 11/BUS 12A SYNC CHECK RELAY E ESYNC

25AX/11T- TRANSFORMER 1 1/BUS 12A SYNC CHECK AUXILIARY E E12A-SYNC RELAY

25B/11T-12B- MAC B TRANSFORMER 11/BUS 12B SYNC CHECK RELAY E ESYNC

25BX/1 1T- TRANSFORMER 1 1/BUS 12B SYNC CHECK AUXILIARY E E12B-SYNC RELAY

52/13SS MAC N BREAKER FOR PXTBSS013 (STATION SERVICE C CTRANSFORMER 13)

52/14SS MAC A/N STATION SERVICE TRANSFORMER 14 SUPPLY C C

52/15SS MAC N BREAKER FOR PXTBSS015 (STATION SERVICE C CTRANSFORMER 15)

52/16SS MAC B/N STATION SERVICE TRANSFORMER 16 SUPPLY C C

52/17SS MAC B/N STATION SERVICE TRANSFORMER 17 SUPPLY C C

52/18SS MAC A/N STATION SERVICE TRANSFORMER 18 SUPPLY C C

52/6T13A72 OAC N/A 6T13A72 BREAKER FOR 115KV C O,C

52/76702 OAC N/A CIRCUIT 767 STATION 13A TO 12B BREAKER 76702 C C

52/7T1352 OAC N/A CIRCUIT 7T STATION 13A TO 12A BREAKER 7T1352 C C

52/7T13A72 OAC N/A 7T13A72 BREAKER FOR 115KV C O,C

52/8X13A72 OAC N/A 8X13A72 BREAKER FOR 115KV C O,C

704A RHR A RHR PUMP A SUCTION MOV 0 0

704B RHR B RHR PUMP B SUCTION 0 0

767-PILOT- OAC N/A 767 PILOT WIRE PROTECTION E EWIRE

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Position Position

7T-LINE-DIFF-PROT OAC N/A 7T DIFFERENTIAL CURRENT RELAY, GINNA SIDE E EPROT

822A RHR N/A CVCS LETDOWN TO RHR IV 0 0

822B RHR N/A RHR PUMPS MINIMUM RECIRCULATION VLV 0 C

825A SIS A RWST TO SI PUMP A 0 0

825B SIS B RWST TO SI PUMP B 0 0

851A RHR A RHR PMP SUCT FROM CONT SUMP C 0

851B RHR B RHR PMP SUCT FROM CONT SUMP C 0

852A RHR A RHR PMP DISCH C 0

852B RHR B RHR PMP DISCH C 0

86/11A-DIFF MAC B BUS 11A DIFFERENTIAL LOCKOUT RELAY F F

86/11B-DIFF MAC B BUS 11B DIFFERENTIAL LOCKOUT RELAY F F

86/12A-DIFF MAC B BUS 12A DIFFERENTIAL LOCKOUT RELAY F F

86/12B-DIFF MAC B BUS 12B DIFFERENTIAL LOCKOUT RELAY F F

86B/12A-DIFF MAC A BUS 12A DIFFERENTIAL BACKUP LOCKOUT RELAY F F

86B/12B-DIFF MAC B BUS 12B DIFFERENTIAL BACKUP LOCKOUT RELAY F F

871A SIS A SI PUMP C DISCHARGE TO LOOP B MOV-871A 0 0

871B SIS B SI PUMP C DISCHARGE TO LOOP A MOV-871B 0 0

878A SIS A SI PUMP DISCHARGE TO B HOT LEG C O,C

878B SIS B SI PUMP DISCHARGE TO B COLD LEG C 0

878C SIS A SI PUMP DISCHARGE TO A HOT LEG C O,C

878D SIS B SI PUMP DISCH MOV C 0

896A SIS A RWST TO CNMT SPRAY & SI PUMPS MOV 0 0

896B SIS B RWST TO CNMT SPRAY & SI PUMPS MOV 0 0

ACPDPDG01 LAC A DIESEL GENERATOR A HEAT TRACE PANEL E E

ACPDPDG02 LAC B DIESEL GENERATOR B HEAT TRACE PANEL E E

BA-CTRL CVC 2 BORIC ACID BLEND CONTROL E E

CIA02A IAS A INSTRUMENT AIR COMPRESSOR A C C

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Position Position

CIA02B IAS B INSTRUMENT AIR COMPRESSOR B C C

CIA02C [AS B INSTRUMENT AIR COMPRESSOR C C C

CSA02 PSA A SERVICE AIR COMPRESSOR E,D E

DG-SYN DGS A/B DIESEL GENERATOR SYNCH CONTROL F F

EAC02A RHR N/A RESID HEAT REMOVAL HX A F F

EAC02B RHR N/A RESID HEAT REMOVAL HX B F F

EAC06A RHR A RHR PUMP COOLER A F F

EAC06B RHR B RHR PUMP COOLER B F F

EAC13 SFP N/A SPENT FUEL POOL HEAT EXCHANGER B F F

EAC14 SFP N/A SPENT FUEL POOL HEAT EXCHANGER A F F

EIA01A IAS N/A IA CMPRSR A AFT CLR F F

EIA01B IAS N/A IA CMPRSR B AFT CLR F F

FI-626 RHR 3 RHR FLOW INDICATOR F F

FOX3 IAC 2 FOXBORO INSTRUMENT RACK 3 E E

FT-110 CVC 1 BORIC ACID FLOW TO BLENDER MAGNETIC FLOW XMTR F F

FT-111 CVC 1 REACTOR MAKEUP WATER FLOWRATOR MTR F F

FT-115A CVC 2 RCP A SEAL INJECTION FLOW XMTFR, F F

FT-116A CVC 2 RCP B SEAL INJECTION FLOW XMTR F F

FT-128 CVC 4 CHARGING LINE FLOW XMTR F F

FT-689 RHR 1 RHR LOOP FLOW TRANSMITTER F F

INVTAMSAC RPS N/A AMSAC INVERTER E E

LI-i102 CVC 2 BORIC ACID STORAGE TANK A LEVEL INDICATION AT F FMCB

LI-106 CVC 3 BORIC ACID STORAGE TANK B LEVEL INDICATION AT F FMCB

LI-112 CVC 2 VCT LEVEL INDICATION (MCB) F F

LI-139 CVC 1 VCT LEVEL INDICATION F F

LI-171 CVC 3 BORIC ACID STORAGE TANK B LEVEL INDICATION AT F FMCB

LI-172 CVC 2 BORIC ACID STORAGE TANK A LEVEL INDICATION AT F FMCB

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Position Position

LI-432A RCS N/A RCP LOOP A DUAL LVL IND F F

LI-432B RCS N/A RCS LOOP B DUAL LVL IND F F

LI-942 RHR 1 CONTAINMENT SUMP B LEVEL INDICATOR F F

LI-943 RHR 3 CONTAINMENT SUMP B LEVEL INDICATOR F F

MCCC-NVLS LAC A MOTOR CONTROL CENTER C NON VITAL LOAD SHED E E

MCCD-NVLS LAC B MOTOR CONTROL CENTER D NON VITAL LOAD SHED E E

MCCK LAC A MOTOR CONTROL CENTER K E E

NI-31B NIS 1 NIS SOURCE RANGE INDICATION F F

NI-32B NIS 3 NIS SOURCE RANGE INDICATION F F

PAC05 FPC A REFUELING WATER PURIFICATION PUMP E E

PAC07A SFP A SPENT FUEL POOL RECIRCULATING PUMP A E E

PCH03A CVC A BORIC ACID TRANSFER PUMP A E E

PCH03B CVC B BORIC ACID TRANSFER PUMP B E E

PCH08A CVC A REACTOR MAKEUP WATER PUMP A D D

PCH08B CVC B REACTOR MAKEUP WATER PUMP B D D

PPSA OAC A PREFERRED POWER SUPPLY A E E

PPSB OAC B PREFERRED POWER SUPPLY B E E

PWD10A WDS A REACTOR COOLANT DRAIN TANK PUMP A E,D E

PWD10B WDS B REACTOR COOLANT DRAIN TANK PUMP B E E

PX13AO06 OAC N/A # 6 POWER TRANSFORMER AT STATION 13A E E

PX13AO07 OAC N #7 POWER TRANSFORMER AT STATION 13A E E

PXYD012A OAC A/N STA AUX XFMR 12A E E

PXYD012B OAC BIN STA AUX XFMR 12B E E

TCH04 CVC N/A VOLUME CONTROL TANK F F

TCH07A CVC N/A BORIC ACID STORAGE TANK A F F

TCH07B CVC N/A BORIC ACID STORAGE TANK B F F

TI-409A-1 RCS 1 RCS LOOP A HL INDICATION (MCB) F F

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Position Position

TIA04A IAS N/A INSTRUMENT AIR COMPRESSOR A RECEIVER F F

TIA04B IAS N/A INSTRUMENT AIR COMPRESSOR B RECEIVER F F

TIA04C IAS N/A INSTRUMENT AIR COMPRESSOR C RECEIVER F F

TIA-635 SFP A/B SPENT FUEL POOL HIGH TEMPERATURE 115 DEGF F FHIGH-LOW LEVEL 20" 12"

TRANSF-12A- STATION AUXILIARY TRANSFORMER 12A BACKUPBU-PROT LOCKOUT RELAY

TRANSF-12A- MAC N STATION AUX XFMR 12A DIFFERENTIAL RELAY F FDIFF

TRANSF-1 2A-DINF-1- MA AN STATION AUXILIARY TRANSFORMER 12A DIFFERENTIALDIF- MAC AIN LCOTRAYF F

LOCKOUT LOCKOUT RELAYTRANS F-i12A-NEUT-GRD- MC / STATION AUX XFMR 12A TIME OVERCURRENT GROUNDNEUT-GRD- MAC A/N BAKPRLYF F

BU BACKUP RELAYTRANSF-12A- MAC N STATION AUXILIARY TRANSFORMER 12A INST AND TIME F F

OC-PROT OC RLY

TRANSF-12B- MAC B/N STATION AUXILIARY TRANSFORMER 12B BACKUP F FBU-PROT LOCKOUT RELAY

TRANSF-12B- MAC BIN STATION AUX XFMR 12B DIFFERENTIAL RELAY F FDIFF

TRANSF-12B3- STATION AUXILIARY TRANSFORMER 12B DIFFERENTIALDIF- MAC BIN LOCKOUT RELAYF

LOCKOUTTRANSF-12B3- STATION AUX XFMR 12B TIME OVERCURRENT GROUNDNEUT-GRD- MAC N BACKUP RELAYF

BU

TRANSF-12- MAC N STATION AUXILIARY TRANSFORMER 12B INST & TIMEOC-PROT OC RELAY

TRANSF-6- OAC N/A TRANSF-6-DIFF PROTECTION F FDIFF

UVBUS11A/11 MAC A UNDERVOLTAGE RELAYS AND POTENTIAL F FTRANSFORMERS 11A/12A

UVBUS11B/21 MAC B UNDERVOLTAGE RELAYS AND POTENTIAL F FTRANSFORMERS 11B/12B

UVBUS13 LAC A/N 480V BUS 13 UNDERVOLTAGE PROTECTION F F

UVBUS14 LAC A BUS 14 UV CIRCUITRY F F

UVBUS15 LAC BIN 480V BUS 15 UNDERVOLTAGE PROTECTION F F

UVBUS16 LAC B BUS 16 UV CIRCUITRY F F

UVBUS17 LAC B BUS 17 UV CIRCUITRY F F

UVBUS18 LAC A BUS 18 UV CIRCUITRY F F

Note:

" Normal NPO Position is the position of the component at the start of NPO which isdependent on the specific mode of operation or POS

" Required NPO Position is the position of the component that is required to ensure the

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KSF Path the component is associated with is successful

Abbreviations of positions used are as follows:C = ClosedD = De-energized

E = EnergizedF = Functional

N = Offsite PowerN/A = Not Applicable0 Open

T = Throttle

As discussed in FAQ 07-0040, the components within the required KSF success paths are

compared to the population of components contained in REG's Safe Shutdown Equipment List(SSEL) to determine if the component's function is already addressed as part of the safe

shutdown analysis. The table below lists components that are required for both safe shutdownand for NPO, but have different Normal or Required positions. If a component's function wasappropriately addressed in the REG SSEL, no further action is required; otherwise additional

cable selection was performed.

Table SSA Q2-2 lists those at-power components that have a different functional requirement

for NPO.

Table SSA Q2-2: Common SSD and NPO Components with Different FunctionalRequirements

SSD POSITIONS NPO POSITIONS

NPOSSD Normal Hot Shutdown NPO Normal Rqr

SYS TRAIN COMPID Position Position Position RequiredPosition

RCS B 430 c c c O,c

RCS B 515 0 c 0 O,c

RCS A 516 0 c 0 O,C

RHR A 700 c c 0 O,c

RHR B 701 c C O O,C

RHR A 720 c c 0 0

RHR B 721 c c O O

RHR A 856 0 c 0 O,c

SWS B 4613 0 c 0 0

SWS A 4614 0 c 0 0

SWS A 4615 0 O,C 0 0

SWS A 4616 0 O,C 0 0

SWS B 4664 0 c 0 0

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SSD POSITIONS NPO POSITIONS

NPOSSD Normal Hot Shutdown NPO Normal Rqr

SYS TRAIN COMPID Position Position Position RequiredPosition

SWS A 4670 0 C 0 0

SWS B 4734 0 OC 0 0

SWS B 4735 0 O,C 0 0

CVC A 392A C OC 0 0

RCS B 431C C C C OC

MAC BIN 52/12AX C C 0 0

MAC A/N 52/12BY C C 0 0

LAC A/N 52/BT14-13 0 0 0 O,C

LAC B/N 52/BT16-15 0 0 0 OC

EAC A 52/EG1A1 0 C 0 OC

EAC A 52/EG1A2 0 C 0 OC

EAC B 52/EG1B1 0 C 0 OC

EAC B 52/EG11B2 0 C 0 0,C

CCW A 738A C C 0 0

CCW B 738B C C 0 0

RHR A 850A C C C O,C

RHR B 850B C C C OC

MAC N BUS11A F F E E

MAC N BUS11B F F E E

MAC N BUS12A F F E E

MAC N BUS12B F F E E

RHR A PAC01A D D E E

RHR B PAC01B D D E E

CCW A PAC02A E E,D E E

CCW B PAC02B E E,D E E

SFP B PAC07B E D E E

CVC A PCH01A E E,D E E

CVC B PCH01B E E,D E E

CVC B PCH01C D E,D D E

SIS A PSI01A D D D E

SiS B PSI01B D D D E

SIS A/B PSI01C D D D E

SWS A PSW01A E E,D E E

SWS PSW01B E E,D E E

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Those components that are common between the safe shutdown and NPO analysis can havedifferent Normal and Required positions. This is because NPO normal position is the position ofthe component at the start of NPO, which is dependent on the specific mode of operation orPOS. For example, the existing NSCA may credit the valve in the closed position; however, thevalve may be required open for shutdown modes of operation. Similarly, NPO required position

is the position of the component that is required to ensure the KSF Path the component isassociated with is successful, which may be different from required hot shutdown (HSD)position.

For the following systems, components have been selected for NPO and not included in the at-power analysis because:

" RCS - The pressurizer safety and relief valves may opened or closed for specificSuccess Paths for Reactivity Control (RXC) or Inventory Control (INV) or Decay HeatRemoval (DHR).

" RHR - The valves may be opened or closed for specific Success Paths for RXC, INV, or

DHR or for isolation of paths not used for RXC, INV or DHR. To ensure Success Paths,both trains of pumps are analyzed as energized if the train is available.

* SWS - The valves are opened to provide SW cooling to the Support functions ofInstrument Air Compressors, CCW heat exchangers, and SFP heat exchangers. Toensure Success Paths, both trains of pumps are analyzed as energized if the train isavailable.

* CVC - The valve may be opened for specific Success Paths for RXC or INV. To ensure

Success Paths, both trains of pumps are analyzed as energized if the train is available.* MAC - Both trains of both 4160 VAC busses are analyzed as energized if the train is

available, and offsite power is analyzed in the normal "50/50" alignment.

• LAC - Bus Tie breakers are opened or closed to provide power to the Support functionof Instrument Air.

* EAC - To ensure Success Paths, both trains of emergency on-site power are analyzedas energized if the train is available and offsite power is not available.

* CCW - The valves are opened to provide CCW cooling for the CCW Support function.To ensure Success Paths, both trains of pumps are analyzed as energized if the train is

available.

* SFP - To ensure Success Paths for the SFP Cooling Support function, the pump is

analyzed as energized if the train is available.

* SIS - To ensure specific Success Paths for RXC and INV, both trains of pumps areanalyzed as energized if the train is available.

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Part c

Pinch Points were identified (on a fire zone basis), based on the loss of a KSF. A "No" in thepinch point column indicates that no KSFs were lost in this fire zone. A "Yes" in this columnindicates that one or more KSFs were lost in this fire zone, and therefore a pinch point isconsidered to exist. Fire Zones are then categorized as follows:

* Category 1 Fire Zones are not pinch points, as they were found to have at least onesuccess path for each KSF. Standard "Defense-in-Depth" (DID) strategies, as specifiedin procedure IP-OUT-2, are adequate to address risk. No recommendations foradditional fire protection measures during High Risk Evolutions (HREs) are made forthese areas.

* Category 2 Fire Zones are pinch points as every success path was lost for at least oneKSF. These KSF success paths can be preserved through fire protection/fire preventionactions, including the verification of functionality of available fire detection andsuppression during HREs.

Table SSA Q2-3 below, from EIR 51-9177694-000, Appendix B - NPO Pinch Point Assessment,provides summary level identification of KSF pinch points on a fire zone / fire area basis. Thetable identifies the unavailable KSF Success Paths associated with each pinch point, and therecommendations for addressing the pinch points.

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Table SSA Q2-3: Summary Level Identification of KSF Losses and Pinch Points

KSF's Lost or Impacted

DHR INV RXC Recommendations

Fire SPT PWR Pinch (See Table 8-1 of this EIRFire Zone Area DHR SFP CVCS RHR SI AC/RHR CVCS RHR SI Point? Category for description) Suppression Detection

ABB ABBM L L L L L L L L L I I Yes 2 1A and/or 2A and/or 3A YES YES

ABM ABBM L L L L L L L L L L L Yes 2 1A and/or 2A and/or 3A YES YES

ABO ABI L L L L L L L L L L L Yes 2 1A and/or 2A and/or 3A YES YES

AHR CC L L L L L L L L L L L Yes 2 1A and/or 2A and/or 3A YES YES

AVT BOP I I I I I I I I I I I No 1 Not a pinch point. Noaction needed.

BR1A BR1A L L L L L L L L L L L Yes 2 1A and/or 3B and/or 5 NONE YES

BRIB BR1B L L L L L L L L L L L Yes 2 1A and/or 3B and/or 5 NONE YES

CHG CHG L I L L L L L L L I I Yes 2 1A and/or 3B and/or 5 NONE YES

Not a pinch point. NoCPB ABI I I I I I I I I I I I No 1 actin n t.ded

action needed.

CR Cc L L L L L L L L L L L Yes 2 1A and/or 2A and/or 3A YES YES

CT CT L L L L L L L L L L L Yes 2 1A and/or 2A and/or 3A YES YES

EDG1A EDG1A I L L I L L L L L I L Yes 2 1A and/or 2A and/or 3A YES YES

EDG1B EDG1B I L L I L L L L L I L Yes 2 1A and/or 2A and/or 3A YES YES

GAB BOP I I I I I I I I I I I No 1 Not a pinch point. Noaction needed.

Not a pinch point. NoH2 BOP I I I I I I I I I I I No 1 actinnteded

action needed.Not a pinch point. No

IB-0 ABI I I I I I I I I I I I No 1 actin neededaction needed.

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Fire SPT PWR Pinch (See Table 8-1 of this EIRFire Zone Area I Point? Category for description) Suppression Detection

IBN-1 ABI L I L L L L L L L I I Yes 2 1A and/or 2A and/or 3A YES YES

IBN-2 ABI I I I I I I I I I I I No 1 Not a pinch point. No


IBN-3 ABI I I I I I I I I I I I No 1 actin n ted Ndaction needed.

Not a pinch point. NoIBN-4 ABI I I I I I I I I I I I No 1 actin n ted Nd


IBS-1 ABI I I I I I I I I I I I No 1 actin n t.dedaction needed.

IBS-2 ABI I I I I I I I I I I I No 1 Not a pinch point. Noaction needed.

IBS-3 ABI I I I I I I I I I I I No 1 Not a pinch point. Noaction needed.

N2 ABI I I I I I I I I I I I No 1No apic p in. oaction needed.

PA-NE PA I L L L L L L L L L L Yes 2 1B and/or 3B and/or 5 NONE NONE

Not a pinch point. NoPA-NW PA I I I I I I I I I I I No 1 actin n teded

action needed.

PA-SE PA I I I I I I I I I I I No 1 Not a pinch point. Noaction needed.

PA-SW PA I I I I I I I I I L L Yes 2 1B and/or 3B and/or 5 NONE NONE

RC-1 RC L I L L L L L L L I I Yes 2 1A and/or 3B and/or 5 NONE YES

RC-2 RC L I L L L L L L L I I Yes 2 1A and/or 3B and/or 5 NONE YES

RC-3 RC I I L I L L L L L I I Yes 2 1A and/or 3B and/or 5 NONE YES

RR CC L L L L L L L L L L L Yes 2 1A and/or 2A and/or 3A YES YES

Not a pinch point. NoSAF SAF I I I I I I I I I I I No 1 actinnteded

action needed.

SB-1 BOP I I I I I I I I I L L Yes 2 1A and/or 3B and/or 5 YES NONE

Not a pinch point. NoSB-1HS BOP I I I I I I I No 1 actinne.dNo

I I Iaction needed.

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Fire SPT PWR Pinch (See Table 8-1 of this EIRFire Zone Area DHR SFP CVCS RHR SI ACIRHR CVCS RHR SI Point? Category for description) Suppression Detection

SB-1WT BOP I I I I I I I I I No 1 Not a pinch point. Noaction needed.

SB-2 BOP I I I I I I I I I I I No 1 Not a pinch point. Noaction needed.

SH-1 SH I L L I L L L L L I L Yes 2 1A and/or 2A and/or 3A YES YES

SH-2 SH I L L I L L L L L I L Yes 2 1A and/or 2A and/or 3A YES YES

SH-3 SH I I I I I I I I I I I No 1 Not a pinch point. Noaction needed.

TB-1 BOP I L L I L L L L L L L Yes 2 1A and/or 2A and/or 3A YES YES

TB-1FP BOP I I I I I I I I I I I No 1 Not a pinch point. Noaction needed.

TB-2 BOP I L L L L L L L L L L Yes 2 1A and/or 2A and/or 3A YES YES

Not a pinch point. NoTB-3 BOP I I I I I I I I I I I No 1 actinneededaction needed.

T-LOOPA RC L I L L L L L L L I I Yes 2 1A and/or 3B and/or 5 NONE YES

T-LOOPB RC L I L L L L L L L I I Yes 2 1A and/or 3B and/or 5 NONE YES

TO BOP I I I I I I I I I I I No 1 Not a pinch point. Noaction needed.

T-PRZR RC L I L L L L L L L I I Yes 2 1B and/or 3B and/or 5 NONE NONE

TREACTOR RC L I L L L L L L L i i Yes 2 iA andior 3B and/or 5 NONE YES

TSC-1M BOP I I L L L I L I I I I Yes 2 1A and/or 2A and/or 3A YES YES

TSC-1N BOP I I L L L I L I I I I Yes 2 1A and/or 2A and/or 3A YES YES

TSC-1S BOP I L L L L L L L L I L Yes 2 1A and/or 2A and/or 3A YES YES

1 A and/or 2A and/or 3AYARD YARD I L L I L L L L L I L Yes 2 and/or YES YES_____and/or 10

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Referring to Table SSA Q2-3, the KSFs are categorized with codes assigned to each KSF - FireZone pair. Three codes have been established to summarize fire impacts:

" "I" (Impacted): At least one of the success paths associated with the KSF is affected, i.e.,

a component of the success path or any of its associated cables within the fire zone areimpacted whereby that path can no longer be assured of being functional. However, at

least one other success path within that KSF is still available.

* "U" (Lost): When all available success paths for a given KSF are impacted.* "N" (None): When no impacts to the KSF are identified.

The KSF pinch point locations listed in Table SSA Q2-3 will be identified to the plant staff

through changes to the outage management procedure that governs fire protection "Defense-in-Depth" (DID) features and shutdown risk management, as listed in the Part a response above.

Proposed options to reduce fire risk will include:" Limit hotwork in this fire zone during HRE conditions

" Prohibit hotwork in this fire zone during High Risk Evolutions

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

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

* Limit transient combustible storage in this fire zone during HRE conditions

" Prohibit transient combustible storage in this fire zone during FIRE conditions

* Provide a firewatch (continuous or periodic) in this fire area during HRE conditions

Part d

There are no actions, including pre-staging actions, (e.g., pre-fire rack-out, locally pinning ofvalves, isolation of air supplies) that are credited to minimize the impact of fire-induced spuriousactuations on power-operated valves. Additional actions are not relied upon as a strategy toreduce fire risk. The assessment of potential risk reduction options (including input fromOperations personnel) concluded that the actual additional risk posed by fire is best controlledthrough the options listed in NRC FAQ 07-0040. Specifically, the Ginna NPO strategy does not

credit the following methods:

" Recovery Actions - Reliance on recovery actions during an outage is difficult tocharacterize for feasibility due to the many variables that could exist, such as blockage

of normal routes, scaffolding impact on lighting, equipment/material staging andmovement, supplementary work force augmentation, planned equipment out of serviceor unavailable for service and resultant off-normal system lineups, etc.. For this reason,recovery actions are viewed as less predictable with respect to reliability and uncertaintyin comparison to the risk reduction options selected.

* Configuration Changes - The use of limited configuration changes to address in apreemptive manner certain high consequence fire-induced failures, most notably

spurious operations of key valves, was considered. However, after discussions withOperations personnel it was concluded that the reduction in operational flexibility torespond to a broader range of potential accidents and abnormal conditions outweighs

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the marginal improvement in risk reduction associated with fire-induced spuriousoperations.

Part e

In the event that NPO credited equipment is deliberately removed from service, Ginna willconsider appropriate contingency measures to reduce fire risk at the impacted locations. IP-OUT-2 addresses pre-outage review for defense in depth. A process currently exists forproviding compensatory actions for Appendix R equipment removed from service (refer to GinnaProcedure A-601.13, "FIRE PROTECTION / APPENDIX R COMPENSATORY ACTIONS"). A-601.13 uses several pre-fire compensatory actions that are consistent with the options endorsedby NRC FAQ 07-0040, and will be utilized at Ginna. These options are identified in Part "c" ofthis response. A similar approach will be used for NFPA 805 NPO credited equipment.

Part f

There are no recovery actions relied upon for the NPO analysis. However, recovery actionscould provide an option to respond to plant conditions, equipment alignments, or equipmentremoved from service during an outage. Note that, because no recovery actions are inherentlyrelied upon, there are no instruments relied upon to provide operator cues for recovery actions.Instruments which are part of the non-power operations analysis are not credited for initiation ofany action.

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SSA RAI 03

LAR Attachment S, Table S-2, "Plant Modifications Committed" lists the proposed modificationsS2-1; S2-2; S2-3; S2-4; S2-5; S2-6; S2-7; S2-8; S2-9; S2-10; S2-11;S2-12; S2-13; S2-14; S2-15; S2-16; and S2-18. With respect to fire risk reduction measures currently in place, provide astatement regarding whether or not fire risk reduction measures have been implemented inaccordance with the plant's fire protection program for the listed modifications.

Response

With respect to fire risk reduction measures currently in place, these fire risk reductionmeasures are implemented in accordance with the plant's fire protection program.

Fire risk reduction measures that have been implemented include, as appropriate andapplicable, such generic or modification-specific measures as:

" Confirming existing procedural direction addresses actions to compensate forcapabilities lost due to a fire, such as including direction for the use of alternateindication that may not be impacted by a fire.

Where applicable, procedures were revised to enhance the procedure and ultimatelyreduce the risk to the plant. The Procedure Change Request Process requires that aFire Protection Program and Appendix R Conformance Review Screen form becompleted for each procedure change in accordance with the Fire Protection ProgramProcedure (A-202). This form is reviewed by the Fire Protection Engineer and FireMarshal. Changes to Fire Response Procedures (FRPs) and Emergency ResponseProcedures (ER-FIRE) also require a 10 CFR 50.59 review.

" FPS-16 (Bulk Storage of Combustible Materials and Transient Fire Loads) was alsorevised to reduce the risk of fire through defense in depth measures such as limitingcombustibles and tracking transients, and additional restrictive controls on combustiblematerials in plant areas with Hemyc wrap fire barrier.

" Permanent combustible loading is also tracked within the Combustible Loading designanalysis DA-ME-98-004, through the incorporation of completed combustible loadingworksheets. This is a part of the modification process and the fire protection program inaccordance with CNG-CM-1.01-1003, A-202, and EP-3-P-0132.

* Performing a minimum of (3) inspections per week in plant areas outside containmentthat have Hemyc'fire barrier wraps as specified in Fire Protection Program Procedure A-202. These inspections are being performed to ensure control of activities andconditions that affect fire risk in "B" Battery Room, Intermediate Building Clean Sidebasement elevation 253'-8", Intermediate Building Clean elevation 267'-3", AuxiliaryBuilding basement, and Auxiliary Building Mezzanine. Any deficiencies that couldchallenge the Hemyc barriers are immediately corrected, or a fire watch is established.The inspections will continue in the interim period between now and when all the NFPA805 modifications are completed.

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Plant Fire Protection Tours are also performed in accordance with procedure A-54.7.These tours are a risk reduction measure for the defense in depth aspect of preventingfires from occurring.

Control of all plant modifications in regards to the impact on the fire protection program.These reviews are performed and documented in accordance with EP-3-P-0132 [Sec. 5,Attach. 1-9] and CNG-CM-1.01-1003 [Attach. 12].

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SSA RAI 04

LAR Table S-2 describes modification ESR-1 1-0050 as "provide a water source that will be freeof the fire's effect and supported by a DG in the SB AFW"

Provide a more detailed description of this modification that includes identification of theplanned physical hardware changes or additions, the communications between local controlstations, the operation of the equipment, and the command and control process with the newmodifications.

Response

The Diesel Driven Standby Auxiliary Feedwater (ESR-1 1-0050) Project is providing the StandbyAuxiliary Feedwater (SB AFW) pumps with an additional source of de-ionized water (160,000gallon tank) and power (1 MWe diesel generator).

The new tank will be located adjacent to the SB AFW Building. The tank contents will be analternate source of water for the SB AFW pumps. Piping and manual valves will be added sothat the SB AFW pumps can draw suction from the new tank and the existing safety relatedservice water system will still be available. SB AFW pump discharge can be recirculated,through a breakdown orifice, back to the tank to facilitate pump testing, or can be discharged tothe steam generators. Additionally, piping will be added in order to transfer de-ionized waterfrom the tank into the condensate system in the Auxiliary Building, or from the de-ionized waterheader to fill the new tank. Piping will be added to allow for a continuous flow of heated water,via a circulating pump and heater, to provide freeze protection for the tank. The tank will beequipped with level and temperature indication, overflow piping, vacuum/relief valve, vent, drain,sample line, manways, and numerous spare penetrations to support any future needs. The newtank will be of sufficient capacity to ensure the removal of decay heat immediately following areactor trip for approximately 24 hours.

The new Diesel Generator (DG) and associated distribution equipment and cabling will beinstalled in an addition to the existing SB AFW Building. It will have the capability of providingpower to both SB AFW pumps. The DG and normal loads will be powered from off-siteresidential power which is completely independent from the plant and the normal offsite powerto the plant. The DG can be locally started or will auto start upon loss of residential power topower the normal loads. Starting of the SB AFW pumps with the DG supplying power will beaccomplished manually using newly installed transfer switches and associated distributionequipment.

Once an Auxiliary Operator or other designated individual is dispatched by the on-shiftOperations staff to start the DG and/or start the SB AFW pumps if the DG is already running,communications will be maintained via radio between control stations. Instructions and

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command will come from Shift supervision and will be directed and controlled per new orrevised Emergency Response Procedures.

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Joseph Pacher Fax: CENGOS - NRC: Home Page · 2014. 10. 3. · Joseph Pacher Site Vice President...

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