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Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 - Request for Additional Information Regarding the National Fire Protection Association Standard 805 License Amendment Request
	ML15229A143
Person / Time
Site:	Calvert Cliffs
Issue date:	02/09/2015
From:	Gellrich G H; Exelon Generation Co
To:	; Document Control Desk, Plant Licensing Branch 1
References
	Download: ML15229A143 (89)
	v • d • e

~Exeton GenerationeGeorge GelirichSite Vice PresidentCalvert Cliffs Nuclear Power Plant1650 Calvert Cliffs ParkwayLusby. MD 20657410 495 5200 Office717 497 3463 Mobilewww~exeloncorp.comgeorge~gellrich@exeloncorp.com10 CFR 50.90February 9, 2015U. S. Nuclear Regulatory CommissionATTN: Document Control DeskWashington, DC 20555Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2Renewed Facility Operating License Nos. DPR-53 and DPR-69NRC Docket Nos. 50-317 and 50-318

Subject:

Request for Additional Information Regarding the National Fire ProtectionAssociation Standard 805 License Amendment Request

References:

1.Letter from G. H. Gelirich (CCNPP) to Document Control Desk (NRC), datedSeptember 24, 2013, License Amendment Request re: Transition to 10 CFR50.48(c) -NFPA 805 Performance Based Standard for Fire Protection2. Letter from N. S. Morgan (NRR) to G. H. Gelirich (CCNPP), datedJanuary 12, 2015, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 -Request for Additional Information Regarding the National Fire ProtectionAssociation Standard 805 License Amendment Request (TAC Nos. MF2993and MF2994)in Reference 1, Calvert Cliffs Nuclear Power Plant, LLC submitted a license amendmentrequest to transition to 10 CFR 50.48(c) -NFPA 805 Performance Based Standard for FireProtection. In Reference 2 the NRC staff requested additional information regarding thisamendment request. Attachment (1) and the Enclosure provide the response to the request foradditional information. The schedule for providing responses to individual questions wasprovided in Reference 2. Enclosure 1 contains markups of the original license amendmentpackage pages and supersedes the previously provided pages.This additional information does not change the No Significant Hazards Determination providedin Reference 1. No regulatory commitments are contained in this letter.Should you have questions regarding this matter, please contact Mr. Douglas E. Lauver at(410) 495-5219.

Document Control DeskFebruary 9, 2015Page 2I declare under penalty of perjury that the foregoing is true and correct.February 9, 2015.Executed onRespectfully,e6' 2/.,George l-L'GellrichSite Vice PresidentGHG/PSF/bjm

Attachment:

(1) Request for Additional Information Regarding the National Fire ProtectionAssociation Standard 805

Enclosure:

1. Updated pagescc: NRC Project Manager, Calvert CliffsNRC Regional Administrator, Region INRC Resident Inspector, Calvert CliffsS. Gray, MD-DNR ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THENATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805Calvert Cliffs Nuclear Power PlantFebruary 9, 2015 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805By letter dated September 24, 2013 Calvert Cliffs Nuclear Power Plant, LLC (CCNPP),submitted a license amendment request (LAR) for Calvert Cliffs Nuclear Power Plant, Units 1and 2 (Calvert Cliffs) to transition its fire protection licensing basis from Title 10 of the Code ofFederal Regulations (10 CFR) Section 50.48(b) to 10 CFR 50.48(c), National Fire ProtectionAssociation Standard (NFPA) 805, "Performance-Based Standard for Fire Protection for LightWater Reactor Electric Generating Plants," 2001 Edition. The Nuclear Regulatory Commission(NRC) staff is reviewing the application and has determined that the following additionalinformation is needed to complete the review of the LAR:Fire Protection Engineerin,. (FPE) Request for Additional In formation (RAI) 01:Section 3.3.4 of NFPA 805, 2001 Edition, requires that thermal insulation materials, radiationshielding materials, ventilation duct materials, and soundproofing materials be noncombustibleor limited combustible. in Attachment A, "NEI [Nuclear Energy Institute) 04-02 Table B-i -Transition of Fundamental Fire Protection Program & Design Elements," of the LAR, thelicensee stated that the plant "Complies with Clarification" on the basis that the referencedprocedures, specifications, and the Combustible Loading Analysis Database control andaccount for the use of thermal insulation materials, radiation shielding materials, ventilation ductmaterials, and soundproofing materials. The licensee does not state whether these materialsare specified in the documents to be noncombustible or limited combustible. Provide thefollowing information:a. Clarify that the procedure(s), specifications, and database specify that thermal insulationmaterials, radiation shielding materials, ventilation duct materials, and soundp roofingmaterials shall be noncombustible or limited combustible.b. Clarify in the compliance bases whether thermal insulation materials, radiation shieldingmaterials, ventilation duct materials, and soundp roofing materials that are eitherpermanently or temporarily installed in the plant are noncombustible or limitedcombustible.c. If installed materials are not noncombustible or limited combustible, describe how thesematerials are accounted for and managed in the fire protection program.CCNPP RESPONSE FPE RAI 01:Response to be provided by 3/11/15.FPE RAI102:Section 3.4.1(c) of NFPA 805 requires that the fire brigade leader and at least two brigademembers have sufficient training and knowledge of nuclear safety systems to understand theeffects of fire and fire suppressants on nuclear safety performance criteria (NSPC). InSection 1.6.4.1, "Qualifications," of NRC Regulatory Guide (RG) 1.189, "Fire Protection forNuclear Power Plants," Revision 2, September 2009 (ADAMS Accession No. ML 092580550),the NRC staff has acknowledged the following example for the fire brigade leader as sufficient:The brigade leader should be competent to assess the potential safety consequences of afire and advise control room personnel. Such competence by the brigade leader may beevidenced by possession of an operator's license or equivalent knowledge of plantsystems.1 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805In Attachment A, the licensee stated that it complies and references Procedure SA-1-105, FireBrigade Training, Section 4.4.A. 1, which includes the NFPA 805, Section 3.4. 1(c) requirementas a responsibility for the shift manager to assure the fire brigade members have the requisitetraining and knowledge. Provide additional detail regarding the training that is provided to thefire brigade leader and members that addresses their ability to assess the effects of fire and firesuppressants on NSPC.CCNPP RESPONSE FPE RA! 02:The compliance basis of Section 3.4.1(c) has been changed to "Complies with Clarification."Calvert Cliffs Nuclear Power Plant (CCN PP) is utilizing the exception to 3.4.1 (c), which states:"Exception: Sufficient training and knowledge shall be permitted to be provided by anoperations advisor dedicated to industrial fire brigade support."CCNPP administrative procedures and the UFSAR ensure that an operations technical advisor,a licensed operator position, is dedicated to respond with the industrial fire brigade.FPE RA1 03:In the compliance bases in Attachment A for NFPA 805, Sections 3.10.1(2) and 3.10.3, thelcensee refers to a required action in Attachment S, Table S-2, item 18 of the LAR.Attachment S, Table S-2 does not include an Item 18; however, Attachment 5, Table S-2,Item 17 appears to address these elements. Confirm that Attachment 5, Table S-2, Item 17 isthe correct reference for the implementation item or provide the correct implementation item forthe Halon system actions identified in the LAR.CCNPP RESPONSE FPE RAI 03:Attachment 5, Table S-2, Item 17 is the correct reference for the implementation item identifiedin Attachment A for NFPA 805 Sections 3.10.1(2) and 3.10.3. The compliance basis forNFPA 805 Sections 3.10.1(2) and 3.10.3, have been revised to reference Attachment 5,Table S-2, Item 17.FPE RAI 04:Section 3.11.3(2), ,'Fire Barrier Penetrations," of NFPA 805 requires that fire dampers complywith NFPA 90A, "Standard for the Installation of Air-Conditioning and Ventilating Systems." InAttachment A, the licensee requested NRC approval for the use of a performance-basedmethodology described in Electric Power Research Institute (EPRI) TR- 1006756, "FireProtection Surveillance Optimization and Maintenance Guide for Fire Protection Systems andFeatures," to change the surveillance frequencies for fire dampers. Attachment L of the LAB,Approval Request 1, which is related to the use of performance-based methodology describedin EPRI TR-1006756, only includes NFPA 805, Section 3.2.3(1), as the NFPA 805 requirementthat is applicable.Clarify if Attachment L, Approval Request 1, is also applicable to NFPA 805, Section 3. 11.3(2),and revise Approval Request 1 as necessary to accommodate the additional section.2 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805CCNPP RESPONSE FPE RAI 04:The intent of Attachment L, Approval Request 1, is to request NRC approval for the option toutilize the performance-based methodology described in EPRI TR-1006756 for all fire protectioninspection, testing, and maintenance at CCNPP. Therefore, the approval request is applicableto NFPA 805 Section 3.2.3(1).Attachment A, Section 3.11.3(2) has been revised to remove the "Submit for NRC Approval"compliance statement. The additional existing "Complies, with Required Action" compliancestatement remains in Attachment A as damper inspection frequencies may potentially beupdated based on the guidance in EPRI TR-1006756 during or after implementation.FPE RA! 05:In Attachment L, Approval Request 2, the licensee proposed a performance-based approach toevaluate the acceptability of unprotected cables located above the suspended ceilings forcompliance with the requirements of NFPA 805, Section 3.3.5.1. Provide the followinginformation:a. Provide further details that describe the extent of use of extension cords that are locatedabove the suspended ceilings, such as number, length, size, use (e.g., types of electricalloads), and if the extension cords are for permanent or temporary use.b. Describe the administrative controls that are (or will be) in place to maintain the technicalbases for the request (e.g., prevent/limit future placement of ignition sources andcombustible materials, periodic surveillance above the ceiling, etc.).c. Clarify the following:i.If the Nuclear Safety Capability Assessment (NSCA) credited cables that are routed inmetal conduit above the suspended ceiling need to be free from fire damage in order tosupport a nuclear safety function or fire risk evaluation (FRE) for a fire in the fire areasdescribed in this request.ii. The NSPC discussion implies fire damage will not occur because, in part, the cablesare protected in metal conduit or in metal covered trays. Metal conduit and metal traysare not generally sufficient to protect cables from exposure fire damage. Provideadditional discussion and/or details that provide assurance that NSCA credited cablesare not susceptible to damage from extension cords or other potential fire hazards inthe area above the ceiling.d. The licensee appears to conclude that because defense-in-depth (DID) Echelon 1 issatisfied, that Echelons 2 and 3 are also satisfied. The NRC staff notes that DID is basedon a balance of the three echelons. Provide additional details related to how Echelons 2(fire detection and suppression) and 3 (safe shutdown) of the DID concept are maintained.CCNPP RESPONSE FPE RAI 05:05a -Response to be provided 4/13/15.05b -Response to be provided 3/11/15.3 ATTACHMENT (1)REQUEST FOR ADDITIONAL iNFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 80505c -i.The NSCA credited cables that are routed above suspended ceilings were evaluated on afire area basis to determine if their failure would result in a VFDR. Cables routed in metalconduit were not screened out of the analysis or considered to be free from the effects offire in the area; those cables above the suspended ceiling were evaluated as failed in theNSCA. There are some conduits routed above suspended ceilings that contain NSCAcredited cables. A few conduits contained NSCA cables that resulted in VFORs. ThoseVFDRs were evaluated in accordance with NFPA 805, Section 4.2.4.2, performance-based approach -fire risk evaluation with simplifying deterministic assumptions. The riskassessment concluded for each of these VFDRs that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFFA 805 Section 4.2.4, therefore, no furtheraction is required.ii. As stated in the response to part (i) of this response, the NSCA did not credit the metallicconduit as a means to prevent fire induced failure of NSCA-credited cables routed abovesuspended ceilings. The NSCA credited cables that are routed above suspended ceilingswere evaluated on a fire area basis to determine if their failure would result in a VFDR.Cables routed in metal conduit were not screened out of the analysis or considered to befree from the effects of fire in the area; those cables above the suspended ceiling wereevaluated as failed in the NSCA. NSCA cable failures that resulted in a VFDR wereevaluated in accordance with NFPA 805, Section 4.2.4.2, performance-based approach -fire risk evaluation with simplifying deterministic assumptions. NSCA credited cables aresusceptible to damage from extension cords and other potential fire hazards in the areaabove a ceiling and no assurance is given that metal conduit will protect those cables.05d -Response to be provided 4/13/15.FPE RAt 06:In Attachment L, Approval Request 3, the licensee requested the use of procedural guidancethat will allow performance of welding, cutting and other hot work in sprinklered fire areas whilethe suppression system is impaired, as an acceptable performance-based approach to complywith NFP'A 805, Section 3.3.1.3.1. Provide the following information:a. In the bases for the request: the licensee stated that this request is applicable to any firearea containing a sprinkler system, as identified in Attachment C, Table C-2. Discuss thebases for limiting this hot work procedure request to only fire areas that contain requiredfire sprinkler systems identified in Attachment C, Table C-2.b. Describe the hot work administrative controls for the fire areas that contain a suppressionsystem that is no._t identified as a required suppression system in Attachment C, Table C-2,and whether the administrative controls are different than those for fire areas with requiredfire suppression systems.c. In the bases for the request, the licensee stated that permanent combustibles locatedwithin 35 feet of the work area that cannot be removed must be covered with theappropriate style of blanket. Clarify if the "appropriate style of blanket" is a listed orapproved welding curtain, welding blanket, welding pad, or equivalent, as required byNFPA 51B, "Standard for Fire Prevention During Welding, Cutting, and Other Hot Work."4 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805d. Describe any additional actions/controls to be used when hot work is performed in fireareas/zones where one or more sprinkler systems are impaired above and beyond thosetaken for any other hot work activity conducted when sprinklers are in service.CCNPP RESPONSE FPE RAI 06:a. Attachment L, Approval Request 3, is applicable to any fire area containing a sprinklersystem and is not limited to 0nly fire areas that contain required fire sprinkler systems.Attachment C, Table 0-2 lists all sprinkler systems in plant fire areas, regardless ofwhether the system is required. The text of Attachment L, Approval Request 3, has beenrevised to clarify that the request is applicable to all fire areas containing fire sprinklersystems.b. As discussed in (a), above, Approval Request 3 is applicable to all fire areas containingfire sprinkler systems.c. Permanent combustibles located within 35 feet of the work area that cannot be removedmust be covered with an NFPA 51 B compliant blanket. The text of Attachment L,Approval Request 3, has been revised to clarify that an "appropriate style of blanket" is anNEPA 51 B compliant blanket.d. The Technical Requirements Manual (TRM) ensures that appropriate contingencymeasures are in place when TRM sprinkler systems are not in service. Thesecontingency measures, which are above and beyond those taken for any other hot workactivity conducted when sprinklers are in service, may include, but are not limited to,ensuring backup suppression is available (i.e., laying hose from an operable hose stationin an adjacent fire area). This is addressed by one of the bases for Attachment L,Approval Request 3, which states "Back-up fire suppression equipment is available inareas where the fire suppression system is inactive."Note: Editorial changes were made to the text of Attachment L, Approval Request.FPE RAi 07:NRC endorsed guidance NE) 04-02, "Guidance for Implementing a Risk-Informed,Performance-Based Program Under 10 CFR 50.48(c)," states that, where used in Chapter 3,"power block" and "plant" refer to structures that have equipment required for nuclear plantoperations, such as containment, auxiliary building, service building, control building, fuelbuilding, radiological waste, water treatment, turbine building, and intake structure, or structuresthat are identified in the facility's pre-transition licensing basis.Section 4.1.3 and Attachment i, Table I-1, "Definition of Power Block," of the LAR state thatbuildings that are required for nuclear plant operations (i.e., required to meet the nuclear safetyor radioactive release (BAD) performance criteria identified in Sections 1.5.1 and 1.5.2 of NFPA805) are considered within the power block. The licensee reviewed the plant for compliancewith the BAD performance criteria, and the results are documented in Attachment E, whichincludes the following compartments as screened in for BAD review, but are not described aspart of the power block in Attachment I, Table 1-1:* Interim Resin Storage Facility (Lake Davies)*Material Processing Facility5 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE______PROTECTION ASSOCIATION STANDARD 805 ~-* Office and Training Facility* Original Steam Generator Storage Facility,Pre -Assembly Facility (Upper Laydown Area)* Sewage Treatment Plant* Unit 1 Butler Building* Unit 2 Butler Building* Warehouse No. 3o West Road CageDescribe the basis for excluding these structures from the power block, based on the criteriastated in Section 4.1.3, " ... those that contain equipment required to meet the nuclear safetyand HAD criteria ... ,"an consequently, exclusion from the NFPA 805, Chapter 3 elements thatapply to the power block.CCNPP RESPONSE FPE RAI 07:Response to be provided 3/11/15.FPE RAI 08:Section 3. 11.1 of NFPA 805 requires that each major building within the power block beseparated from the others by barriers having a designated fire resistance rating of 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months or byopen space of at least 50 feet or space that meets the requirements of NFPA 80A,"Recommended Practice for Protection of Buildings from Exterior Fire Exposures." InAttachment A, the licensee stated that it "Complies with Clarification" and described that theNorth Service Building and Turbine Building are analyzed as one fire area in the NFPA 805NSCA, and are, therefore, treated as one building from a building separation perspective. Thelicensee also stated that it "Complies with Use of EEEE's" with respect to excluding the 45"-0"elevation of the North Service Building from the power block. The licensee did not discuss thebasis for excluding this specific elevation from the power block in Attachment I.Provide the basis for excluding the 45'-0" elevation of the North Service Building from the powerblock.CCNPP RESPONSE FPE RAI 08:Attachment I has been revised to discuss the basis for excluding the 45' elevation of the NorthService Building from the power block.The 45' elevation is excluded from Fire Area TB/NSB/ACA and the power block as justified byan engineering equivalency evaluation. The bases for acceptability are summarized as follows:* There are no cables or equipment required to achieve nuclear safety performance criteria(NSPC) in the 45' elevation of the North Service Building.* There are no cables or equipment required to achieve NSPC in the Yard within 50 feet ofthe 45' elevation of the North Service Building.A fire originating in the 45' elevation of the North Service Building will not impact cables orequipment required to achieve NSPO in adjacent fire area TB/NSB/ACA.6 A1-TACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805FPE RAI 09:Section 3.4.1l(a) of NFPA 805 requires that a fully staffed, trained, and equipped fire-fightingforce be available at all times to control and extinguish all fires on site. In Attachment A, thelicensee stated that in Section 5.5.B of Procedure SA-1-1O1, "If there are less than 5 brigademembers notify the Control Room."Current NRC guidance, Frequently Asked Question (FAQ) 12-0063, "Fire Brigade Make-Up"(ADAMS Accession No. ML 121980572), discusses conditions where fire brigade complementmay be less than the minimum for a period of time, in order to accommodate unexpectedabsence of on-duty shift members. Further, licensees may claim prior approval if their currenttechnical specifications or fire protection safety evaluation addresses the issue. If prior approvalhas not been granted, then the licensee should seek NRC approval in the NFPA 805 LMR.Pro vide additional detail on the compliance bases related to conditions when there are less than5 fire brigade members onsite.CCNPP RESPONSE FPE RAI 09:Response to be provided 4/13/15.FPE RA! 10:Section 3.11.5 of NEPA 805 requires that Electrical Raceway Fire Barrier Systems (ERFBS)that are required by NFPA 805, Chapter 4, be capable of resisting the fire effects of the hazardsin the area. In Attachment A, the licensee stated that there are no ERFBS credited forcompliance with Chapter 4, and, therefore, there is no compliance applicable to NEPA 805,Section 3.11.5. However, in Attachment B (Attributes 3.4.1.3, 3.4.1.5, 3.4.2.2, and 3.4.2.3) thelicensee described that one of the means of addressing cable impacts of fire damage is toprotect the cables by an ERFBS.a. Clarify if there were any cable resolutions in the NSCA that credit an EREBS to protect theaffected cables to meet NEPA 805, Chapter 4. If yes, then clarify if the ERFBS are incompliance' with NFPA 805, Section 3.11.5.CCNPP RESPONSE FPE RAI 10:LAR Attachment B documents the Nuclear Safety Capability Assessment Methodology Review.Sections 3.4.1.3, 3.4.1.5, 3.4.2.2, and 3.4.2.3 of LAR Attachment B identify that ERFBS may beutilized as an acceptable method to protect cables from fire damage and be credited within theanalysis. The NSCA has not credited any EREBS.Safe Shutdown Analysis (SSA) RA1 01:Attribute 3.2.1.2 of NE/ 00-0 1, Revision 2, includes the assumption that exposure fire damage tomanual valves and piping does not adversely impact their ability to perform their pressureboundary or safe shutdown function, and that any post-fire operation of a rising stem valvelocated in the fire area of concern should be well justified using an engineering evaluation. InAttachment B, the alignment bases for NEI 00-0 1, Attribute 3.2.1.2, states that manual valvesthat are repositioned for credited NFPA 805 recovery actions (RAs) are included in the NFPA805 NSPC equipment list and are subject to assessment of feasibility.7 ATT'ACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FiRE_______ PROTECTION ASSOCIATION STANDARD 805Provide the following information:a. Clarify if any rising stem valves involved in an RA are subjected to fire damage.b. If any of the valves in the fire area of concern being repositioned by an RA are rising stemnvalves, then clarify if an engineering evaluation was performed to evaluate the exposurefire damage to manual valves and piping to determine if the exposure to fire wouldadversely impact their ability to perform their pressure boundary or safe shutdownfunction. If used, describe the method and results obtained from the engineeringevaluation.CCNPP RESPONSE SSA RAI 01:a. All recovery actions as documented in LAR Attachment G were reviewed. There are norecovery actions that credit the manipulation of rising stem valves that have been exposedto the effects of fire.b. An engineering evaluation is not required as there are no recovery actions that require themanipulation of a rising stem valve that has been exposed to fire.SSA RAI102:Attribute 3.3.1.1.4 of NEI 00-01, Revision 2, includes criteria and assumptions for evaluatingpower cables for breaker coordination concerns and includes safe shutdown cables and thosenon-safe shutdown cables that can impact safe shutdown. In Attachment B of the LAR, thealignment bases for NEI 00-0 1, Attribute 3.3.1.1.4, states that the NSCA circuit identification andanalysis should utilize a "building block" approach and include only, as applicable, the powercable from the NSCA component to the upstream electrical power source.Provide the following information:a. Clarify if cables that supply loads not required to meet the NSPC off of the nuclear safetybuses are classified as "required" cables. if non-nuclear safety cables are not included,then provide the justification for not considering the failure of non-nuclear safety cables inmeeting the breaker coordination criteria for protection.b. The alignment basis states that plant modifications have been identified to achieveselective coordination of breakers/fuses and identified as implementation items inAttachment S, Table S-2. Identify the specific modifications that are required to achievethe selective coordination of breakers/fuses.CCNPP RESPONSE SSA RAI 02:Response to be provided 3/11/15.SSA RA1 03:Attribute 3.5.1.3 of NE/ 00-0 1, Revision 2, includes an assumption that circuit contacts areinitially positioned (i.e., open or closed) consistent with the normal mode/position of the"required for hot shutdown" equipment, and that the analyst must consider the position of the"~required for hot shutdown" equipment for each specific shutdown scenario when determiningthe impact that fire damage to a particular circuit may have on the operation of the equipment.In LAB Attachment B, the alignment basis for Attribute 3.5.1.3 states that the circuit analysismay discount spurious operation based on a fire affected cable being routed in a dedicated8 A1FrACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805conduit and the cable being protected from* external sources of voltage (also taking intoconsideration the potential impact from ground equivalent hot shorts).For multi-conductor cables routed in dedicated conduit, provide a description if intra-cable hotshorts (wire-to-wire shorts) are considered as a potential impact of fire damage on requiredposition of the NSCA equipment (i.e., the function of the initial position of circuit contacts are notaffected by intra-cable hot shorts).CCNPP RESPONSE SSA RAt 03:Response to be provided 3/11/15.SSA RAI 04:The nuclear safety goal described in NFPA 805, Section 1.3.1, is to provide reasonableassurance that a fire during any operational mode and plant configuration will not prevent theplant from achieving and maintaining the fuel in a safe and stable condition. In Section 4.2.1.2,the licensee stated that the NSCA will demonstrate that the plant can achieve and maintain safeand stable conditions for at least 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with the minimum shift operating staff. After12 hours, the Emergency Response Organization (ERO) will be available to support "safe andstable" actions to extend hot standby conditions.a. in Section 4.2.1.2, subsection "Methods to Maintain 'Safe and Stable' and Extend HotStandbY Conditions," of the LAR, local manual actions are described to align varioussystems and functions. In Item No. 8, the licensee stated that should alternating current(AC) charging sources be lost, local manual operator action may be required, and thatstation batteries are capable of providing a minimum of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of 125 V direct currentpower to their respective loads during a station blackout without AC charging sources.The licensee further stated that this time allowance credits securing 11NVIT1 1 in thecable spreading room (CSR) within 45 minutes. Clarify if this local manual action iscredited as an RA in any fire area.b. In Section 4.2.1.2, subsection "Assessment of Risk," the licensee stated that the EROprovides sufficient resources for assessment of fire damage and completion of repairs toequipment necessary to maintain hot standby for an extended period, transition to coldshutdown, or return to power operations as dictated by the plant fire event. Describe ifany repair activities are necessary to maintain hot standby for an extended period (safeand stable conditions), including a detailed description of the specific repairs that would beneeded, the success path(s) being restored, and the time frame required to complete therepair.CCNPP RESPONSE SSA RA! 04:Response to be provided 3/11/15.SSA RA! 05:Section 2.4 of RG 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," Revision 1, dated December 2009 (ADAMS Accession No.ML092 730314) describes the treatment of RAs supplemented by guidance provided inNEI 04-02 and FAQ 0 7-0030, "Establishing Recovery Actions" (ADAMS Accession No.ML 110070485). In RG 1.205, the NRC staff clarifies that operation of alternative or dedicatedshutdown controls while the main control room (MCR) remains the command and control9 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805location would primarily be considered an RA because, for such scenarios, the dedicated oralternative controls are not considered primary. Attachment G of the LAR describes the primarycontrol stations (PCS) and identifies RAs performed at the PCS in Fire Area 16 (Unit 1 CSR and1C cable chase) and Fire Area 17 (Unit 2 CSR and 2C cable chase). Provide the followinginformation:a. Clarify if the control room remains the command and control location for a fire in FireAreas 16 and 17, and if so, discuss how the RAs at the PCS are evaluated for compliancewith NFPA 805, Section 4.2.4.b. In Fire Areas 16 and 17, there are RAs at the PCS that are not associated with a variancefrom deterministic requirement (VFDR).o For Fire Area 16, the RAs are:61CHECKRXSD1; 161CONSERVE1; 161SECHTR1 1_13; 161ADV1C43;1611C43CONTROL; and 161RCSTEMP1.* For Fire Area 17, the RAs are:1 71CHECKRXSD2; 1 71CONSERVE2; 1 71SECHTR2 1_23; 1 71AD V2C43;1712C43CONTROL; and 17IRCSTEMP2.Clarify the purpose of performing these RAs, and whether the actions are required to meetthe NSPC required by NFPA 805, Section 1.5.1.c. In Attachment G, Table G-1 of the LAR, disposition of VFDR 16-19-1 credits RAs at thePCS to energize pressurizer backup heater banks 11 and 13; however, another non-VFDR related BA (161SECHTR1 1_ 13) is credited to secure the pressurizer backup heaterbanks 11 and 13. Discuss how the contradicting RAs are evaluated in the NSCA.d. In LAR Attachment G, Table G-l, RAs are credited to disposition VFDRs 16-27-1 and1 7-25-2 to control atmospheric dump valve (ADV) hand valves to support control of theADVs at the PCS locations 1C43 and 2C43, respectively. However, the RAs(161ADVIC43 and 171ADV2C43) to control the ADVs at the PCS location do not have aVFDR associated with them. Discuss the method for crediting RAs to support the VFDRdisposition without crediting the BA at the PCS.CCNPP RESPONSE SSA RAt 05:Response to be provided 3/11/15.SSA RA1 05:Attachment W Tables W-6 and W-7 of the MAR appear to conflict with information described inAttachment C, Table C-i, and Attachment G, Table G-2. Clarify the following discrepancies:a. In Attachment C, Table C-i, Fire Area 34 is identified as transitioning deterministically inUnit 2 (Section 4.2.3.2 of NEPA 805) with no VFDRs identified. However, inAttachment W, Table W-7, Fire Area 34 is identified as transitioning using performance-based methods (Section 4.2.4.2 of NFPA 805), VFDRs are identified, RAs are credited,and the risk of the RA was calculated. Clarify the correct nuclear safety compliancestrategy for Fire Area 34.b. In Attachment C, Table C-1, the following fire areas are identified as transitioningdeterministically with no VFDRs identified. However, Attachment W, Table W-6 (for10 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805Unit 1) and Attachment W, W-7 (for Unit 2) identify that these fire areas have VFDRsidentified. Further, an FRE was performed that calculated a delta core damage frequency(CDF) and delta large early release frequency (LERF) value as follows:Unit 1: 2, 8, 13, 18, i8A, 22, 23, 25, 26, 27, 28, 31, 38, 40, and 2CNMTUnit 2: 3, 4, 6, 14, 15, 19, 19A, 21, 30, 33, 39, and 1CNMTFor each of these fire areas, clarify the correct nuclear safety cornpliance strategy, andjustify the bases for performing an FRE that is not discussed in the NSCA in LARAttachment C, Table C-i, and the bases for crediting RAs that are not included in LARAttachment G, Table G-1.c. in Attachment C, Table C- 1, the following fire areas are identified as transitioning usingperformance-based methods (FRE) to meet the NSPC, and no RAs were credited (eitherfor risk or DID). However, Attachment W, Table W-6 (for Unit 1) and Attachment W, W-7(for Unit 2) identify these fire areas as crediting RAs and the risk of the RA was calculated:Unit 1: 12, 14, 15, 19A, 21, 30, 32, 33, 35, 36, 39, 1 CNMT, and ISUnit 2: 12, 13, 18A, 20, 26, 27, 28, 32, 34, 35, 36, 40, 2CNMT and ISFor each of these fire areas, clarify the correct nuclear safety compliance strategy forthese fire areas and the bases for crediting RAs that are not included in Attachment G,Table G- 1.d. In Attachment C, Table C-i, the following fire areas are identified as transitioning usingdeterministic methods to meet the NSPC, and no RAs were credited (either for risk orDID). However, Attachment W, Table W-6 (for Unit 1) and Attachment W, W-7 (for Unit 2)identifies these fire areas as crediting RAs and the risk of the RA was calculated:Unit 1: 13, 18, 18,A, 22, 23, 25, 26, 27, 28, and 2CNMTUnit 2: 14, 15, 19, 19,4,21, 30, 33, 39 and 1CNMTFor each of these fire areas, clarify the correct nuclear safety compliance strategy andjustify the bases for not including these RAs in Attachment G, Table G-1, if these RAs areactually credited in the NSCA.CCNPP RESPONSE SSA RAI 06:Response to be provided 4/13/15.SSA RAI107:Modifications were identified in Attachment S, Table S-2, that appear to resolve certain VFDRissues. However, the disposition of the certain VFDRs as summarized in Attachment C,Table C-i1, do not describe whether the modification was credited or not. Provide *clarification on*how the modifications described below were addressed in the disposition of the VFDRs listed:a. Attachment S, Table S-2, Item 7, involves modifying control circuits for the PressurizerPower Operated Relief Valves (PORVs), 1(2)ERV402 and 1 (2) ER V404, to prevent thePORVs from spuriously opening. However, VFDRs 1 6-46-1, 24-26-1, 16-47-1, 24 -2 7-1,17-41-2, 24-63-2, 17-42-2, and 24-64-2 involve fire damage to cables which could result inspurious opening of the Pressurizer PORV, and the VFDR dispositions credits an RA forDID.11 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805b. Attachment S, Table S-2, Item 8, involves modifying the control circuits for the auxiliaryfeed water (AEW) steam admission valves 1(2)CV4070 and 1(2)CV4071 to ensureadequate separation such that one set of valves will be available during a fire in either theCSR or switchgear rooms. However, VFDRs (16-22-1, 17-16-2, 16-26-1, and 1 7-26-2)involve fire damage to cables that could cause the loss of control and/or spuriousoperation of i(2)CV4070 and 1 (2)4071, and the VEOR dispositions credit an BA either toreduce risk (VFDRs 16-22-1 and 17-16-2) or for DID (VFDRs 16-26-1 and 1 7-26-2).c. Attachment S, Table S-2, Itemn 11, involves modifying control circuits for the Main SteamIsolation Valves (MS/Vs), 1(2)CV40430P and 1(2)CV404#OP, to ensure at least onesolenoid dump valve can be energized to close the MS/ Vs. However, VFDRs 1 6-31-1,16-32-1, 1 7-23-2, and 1 7-24-2 involve fire damage to cables that could cause a loss ofcontrol and/or spurious operation of the associated MS/V, and the VFDR dispositionscredit an BA for DID (VFDRs 16-31-1, 1 6-32-1, 1 7-23-2, and 1 7-24-2).CCNPP RESPONSE SSA RAI 0}7:Response to be provided 4/113115.SSA RAI 08.:In Attachment K, Licensing Action 5, the licensee requested that a previously approvedexemption, related to dedicated water curtains as being adequate to maintain the 3-hour firerating of barriers, be transitioned to the NFPA 805 program. The licensee described thesprinkler systems located in Room 216A and Room 106 as supplying the sprinkler heads for thededicated water curtains. In the summary of the exemption approved by the NRC in a letterdated March 15, 1984 (ADAMS Accession No. ML0 10430325), the licensee stated that on theCorridor No. 110 side of the hatch, a dedicated sprinkler head will be supplied from the RoomNo. 116 sprinkler system. However, in the Baltimore Gas & Electric Company submittal datedNovember 21, 1983 (ADAMS Accession No. 8311290159), the licensee stated that on thecorridor No. 110 side of the hatch, a dedicated sprinkler head will be supplied from the RoomNo. 106 sprinkler system. The NRC staff also noted that Attachment C, Table C-i, refers toroom numbers in the "Required Fire Protection System and Features," adAttachment C, TableC-2, refers to fire zones. The NRC staff also noted in the discussion for Licensing Action 1 thatroom numbers at the plant may have changed over time. Provide the following information:a. Describe if the fire zone numbers listed in Attachment C, Table C-2, are the same as theroom numbers listed in the fire area summary in Attachment C, Table C-1. Describe if theroom numbers in Attachment C correspond with the room numbers cited in the previouslicensing actions in Attachment K.b. Provide a description of the water curtain arrangement, including the sprinkler systemsthat supply the required sprinkler heads using the current terminology for rooms, fireareas, and/or fire zones such that the staff can fully understand the installation and howthe installation is represented in the various tables in the submittal and the previouslicensing actions.CCNPP RESPONSE SSA RAl 08:a. The "fire zone" numbers listed in Attachment C, Table 0-2 are the same as the "roomnumbers" listed in the fire area summary in Attachment C, Table C-1. The "roomnumbers" in Attachment C also correspond with the "room numbers" cited in the previous12 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805licensing actions in Attachment K. There is a typographical error in Section II of theEnclosure to NRC letter dated March 15, 1984 (ADAMS Accession No. ML01 0430325),which summarizes the exemption request. This is discussed in part (b.), below.b. Attachment C, Table C-i and Table C-2, identify water curtain sprinkler in room 110 asprotecting the hatch to room 216A and the water curtain sprinkler in room 216A asprotecting the hatch to room 110.The room 216A water curtain is supplied by the room 216A sprinkler system as describedin the BG&E exemption request dated November 21, 1983 (ADAMS Accession No.8311290159) and in the summary of the exemption approved by the NRC in the letterdated March 15, 1984 (ADAMS Accession No. ML01 0430325).The room 110 water curtain is supplied by the room 106 sprinkler system as described inthe BG&E exemption request dated November 21, 1983 (ADAMS Accession No.8311290159). There is a typographical error in Section II of the Enclosure to NRC letterdated March 15, 1984 (ADAMS Accession No. ML01 0430325) which inaccuratelydescribes the system as being supplied by the sprinkler system in room 116. The systemhas been confirmed to be supplied by the sprinkler system in room 106.SSA RA1 09:/In Attachment C, Table C-i, under the heading "Fire Suppression Effects on Nuclear SafetyPerformance Criteria," the majority of the fire areas contain the concluding statement, "Firesuppression in this fire area will not impact the ability to achieve the NSPC in accordance withNFPA 805, Sections 4.2.1 and 4.2.4.1.5." NFPA 805, Section 4.2.4.1.5, is associated with thefire modeling performance-based approach, which the licensee stated it did not use inSection 4.5.2.1 of the LAR. In addition, the suppression effects sections for several other fireareas (e.g., 18A, 20, 21, 22, 23, 35, and 36) contain the statement, "There is no suppressioneffect concern for this fire area as the fire area does not contain NSCA equipment," yet the firearea contains VFDRs. Address the following:a. Clarify the basis for discussing the fire suppression effects for a fire modelingperformance-based approach when the fire areas used a risk evaluation performance-based approach.b. Provide additional discussion for those fire areas where VFDRs are identified, but thesuppression effects discussion states there is no NSCA equipment in the fire area.CCNPP RESPONSE SSA RAI 09:Response to be provided 3/11/15..SSA RAI 10:In Section 4.5.2.2, the licensee stated that there are no VFDF~s that involved performance-based evaluations related to wrapped or embedded cables. However, in Attachment C,Table C-i, Fire Areas 18, 19, 35, 36, and TB/NSB/ACA are performance-based fire areas andcredit EEEE, "ECP- 13-000359 -"Generic Letter (GL) 86-10 Evaluation of Embedded Conduit inthe Turbine Building and Barrier Thickness of the Floor/Ceiling Barrier between AB-4/AB-5 and517/518," which justifies the acceptability of conduits embedded in the Turbine Building floorslab (elevation 279), the floor/ceiling slab between stairwells AB-4 and AB-5, and the horizontal13 ATrACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805cable chases (Rooms 517 and 518). Clarify if the disposition of the VFDRs in Fire Areas 18, 19,35, 36, and TB/NSB/ACA credit the embedment as evaluated in the EEEE.CCNPP RESPONSE SSA RAi 10:Response to be provided 3/11/15.:SSA RAI 11:in Attachment C, Table C-1, the licensee identified Marinite boards as fire protection featuresthat are credited for "S" (required for Chapter 4 separation criteria) and "R" (required for risksignificance) to protect cables for a fire in Fire Area 1CNMT (Unit 1 Containment) and 2CNMT(Unit 2 Containment). Provide the following information:a. Describe the extent that Marinite boards are credited for Chapter 4 separation ("S") andfor risk significance ("R'f) in the Unit 1 and Unit 2 Containments. In addition, describe thedesign and plant configuration of the Marinite boards and the nuclear safety functions thatthe passive fire protection features are protecting.b. Provide previous NRR staff approval (if any) for the use of Marinite boards in containmentto demonstrate meeting the requirements of Appendix B,Section III.G.2, which can becredited to meet the requirements of NEPA 805, Section 4.2.3.4, or evaluate acceptabilityusing a performance-based analysis approach in accordance with NFPA 805,Section 4.2.4.CCNPP RESPONSE SSA RA! 11:Response to be provided 4/13/15.SSA RAI 12:In Attachment C, Table C-2, the licensee makes reference to "Unit 1 Containment (App-BPurposes Only)" and "Unit 2 Containment (App-B Purposes Only)," for fire protection systemsand features. The fire protection systems and features are identified as required for "S"(Re quired for Chapter 4 Separation Criteria), "R" (Required for Risk Significance), and/or "D"(maintain an adequate balance of DID in a change evaluation or FRE).Clarify the meaning of 'Appendix-B Purposes Only" and if these fire protection systems andfeatures are credited with respect to compliance with NFPA 805, Chapter 4.CCNPP RESPONSE SSA RAI 12:The containment areas were divided into rooms under the Appendix R program. The currentdescriptions for these rooms, as identified in plant documents, contain "(App-R Purposes Only)"within their name. The phrase "(App R Purposes Only)" will be removed from plantdocumentation upon implementation of NFPA 805. The phrase can be found in the followingsections of the LAR:* In attachment C, Table C-1 -NEI 04-02 Table B-3 Fire Area Transition* In attachment C, Table 0-2 -Summary of NFPA 805 Compliance Basis and RequiredFire Protection Systems and Features* In attachment E, NEI 04-02 Radioactive Release Transition14 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE____________PROTECTION ASSOCIATION STANDARD 805-The fire protection systems and features are credited with respect to compliance with NFPA 805Chapter 4. The containment fire areas will remain subdivided as described in plant documents.SSA RAI 13:Provide the following pertaining to non-power operations (NPO) discussions provided inSection 4.3 and Attachment D:a. Section 4.3.2 and Attachment D state that incorporation of the recommendations from the"KSF [key safety function] pinch point" evaluations into appropriate plant procedures priorto implementation will be done to ensure the requirements of NFPA 805 are met. Identifyand describe the changes to outage management procedures, risk management tools,and any other document resulting from incorporation of KSFs identified as part ofNFPA 805 transition. Include changes to any administrative procedures such as "Controlof Combustibles."b. For those components that had not previously been analyzed in support of the at-poweranalysis or whose functional requirements may have been different for the non-poweranalysis, cable selection was performed in accordance with approved project procedures.Provide a list of the additional components and a fist of those at-power components thathave a different functional requirement for NPO. Describe the difference between the at-power safe shutdown function and the NPO function. Include with this list a generaldescription by system indicating why components would be selected for NPO and not beincluded in the at-power analysis.c. Section 4.3.1 and Attachments D and H- state that the NPO analysis was performed inaccordance with FAQ 0 7-00410, "Non-Power Operations Clarifications (ADAMS AccessionNo. ML 082200528). However, the LAR did not provide the results of the KSF pinch pointanalysis. Provide a list of KSF pinch points by fire area that were identified in the NPO firearea reviews using FAQ 0 7-0040, including a summary level identification of unavailablepaths in each fire area. Describe how these locations will be identified to the plant staff forimplementation.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 of anyactions being credited to minimize the impact of fire-induced spurious actuations on poweroperated valves (e.g., air-operated valves and motor-operated valves) during NPO(e.g., pre-fire rack-out, actuation of or pinning of valves, and isolation of air supplies).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 KSFs areachieved via RAs or for which instrumentation not already included in the at-poweranalysis is needed to support RAs required to maintain safe and stable conditions.Identify those RAs and instrumentation relied upon in NPO and describe how RAfeasibility is evaluated. Include in the description whether these variables have been orwill be factored into operator procedures supporting these actions.15 ATTACHMENT (1)REOUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE__________ PROTECTION ASSOCIATION STANDARD 805CCNPP RESPONSE SSA RAI 13:13a -Plant procedure updates are to be completed during NFPA 805 implementation. Theseupdates will include procedures such as NO-1-103, Conduct of Lower Mode Operations, toincorporate Key Safety Function (KSF) Pinch Point Analysis as identified in LARAttachment S, Table S-3 IMP-4. Changes will include the following:* Limiting/prohibiting hot work in select Fire Zones.* Detection/Suppression Systems should be verified to be functional, (not tagged out etc.).SLimiting/Prohibiting the hazard of combustible materials.a Using alternate equipment and/or the equipment's position whenever removing power.*Appropriate compensatory measures required during periods of increased vulnerability.=Activities that may impact KSFs should be limited and strictly controlled to mitigate losses.* Consider the hazards from the introductions of combustible materials and sources of fireprecursors.* Limiting work during periods of High Risk Evolution (HRE) conditions.° Ensure HRE are identified in a manner consistent with NUMARC 91-06 and FAQ07-0040.13b -Report NFPA 805-00008, Section 4 provides an overview of the Non-Power Operations(NPO) model development methodology, including cable selection. Section 7 of this reportdocuments systems included and excluded. Attachment A to this report provides detaileddescription of the NPO model. Attachment C to this report contains the NPO Equipment Listincluding required/credited function. NPO only components appear in this attachment but notin NFPA 805-00006, Nuclear Safety Capability Assessment (NSCA) Attachment 7-5, theNSCA Equipment List. Functional differences can be identified by comparing the requiredNSCA and NPO positions which are identified in these reports and also in a plant database.Cable selection packages for components credited in the NPO evaluation that have adifferent function from the function required by the NSCA, were reviewed, evaluated andupdated as required, for all credited functions. Differences in equipment and functions aretypically attributable to the difference in plant operating state.Some example systems where a change in state or different equipment selection may occurinclude:° Process Monitoring -Different instruments are required due to differences in plantoperating state and differences in credited systems (e.g. Shutdown Cooling).* Shutdown Cooling (Low Pressure Safety Injection) -Credited for decay heat removal(DHR) KSF in NPO, not credited in NSCA.* Shutdown Cooling Isolation Valves -Required closed High Low Pressure interface forthe NSCA, required open for DHR KSF in NPO.* High Pressure Safety Injection -Credited for Inventory KSF in NPO, not credited inNSCA.* Auxiliary Feedwater System -Required operable for DHR in NSCA, not credited inNPO.13c -Report NFPA 805-00008, Attachment B documents the results of the pinch point analysis.This attachment identifies the key safety functions (KSFs) that are evaluated and the status16 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805 __of each path that could be used to accomplish the KSF for each fire area. As identified inLAR Attachment D, there were 51 fire areas for which the evaluation identified one or morepinch points for Unit 1 and 46 fire areas for which the evaluation identified one or more pinchpoints for Unit 2. Thirty-five fire areas were identified to have at least one pinch point for bothunits. Report NFPA 805-00008, Attachment B will be used as a reference document insupport of site procedure updates, as discussed in response to item (a) of this RAI, and willbe available during outage planning and HRE fire risk mitigation reviews.13d -Response to be provided 3/11/15.13e -Response to be provided 3/11/15.13f -Recovery actions have not been credited as the sole means of mitigating KSF pinchpoints. However, recovery actions have not been excluded as a method of mitigating fireimpact to KSFs. Recovery actions have been evaluated for several failure modes includingloss of HVAC systems, loss of Instrument Air and loss of control room indicators (where localor backup indication is available). These recovery actions were evaluated using existingplant procedural guidance which will be reviewed and updated as necessary during NFPA805 implementation. Any recovery actions that will be implemented during a HRE will beevaluated for feasibility in a manner consistent with NSCA credited recovery actions.SSA RAI 14:Describe if any RAs require the cross-tie of Unit 1 and Unit 2 systems to achieve the NSPC.Provide the following information:a. Describe whether these cross-connecting RAs require staff from both units. If so,describe how the feasibility analysis reflects the Unit 1 and Unit 2 staffing, communication,and operational interface.b. Describe the operational impacts (by fire), if any, on the unaffected unit created by cross-tying these systems. Describe whether Technical Specification 3.0.3 is entered once thecross-tie with the opposite unit has been completed for fire safe shutdown.CCNPP RESPONSE SSA RAI 14:The only recovery actions that credit a cross-tie between Unit 1 and Unit 2 systems to achievethe NSPC are cross-connecting air systems. Recovery actions to cross-tie the Unit 1Instrument Air system to the Unit 2 Plant Air system are credited for fires in Fire Areas 19, 20and 34.. Recovery actions to cross-tie the Unit 2 Instrument Air system to the Unit 1 Plant Airsystem are credited for fires in Fire Areas 18, 22 and 25. In each area a recovery action wascredited for the VFDR to reduce the risk due to fire in that area.These recovery actions are updates to Table G-1 which currently credits N2 recharge.Additional changes to Table G-1 are anticipated as a result of other RAIs being resolved atdifferent RAI milestone response requirements. LAR Attachment G, Table G-1, RecoveryActions and Activities Occurring at the Primary Control Station(s) will be updated and a markupprovided with the 120 day submittal (4/13/15) that is required to support responses for additionalRAIs.17 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FiREPROTECTION ASSOCIATION STANDARD 805a. Recovery action to cross-tie plant air systems requires staff from both units. The fireimpacted unit staff will direct/request the supporting unit to perform the requiredalignment. The supporting unit staff will then operate the necessary equipment undertheir cognizance and report back to the fire impacted unit operators. The feasibilityanalysis identifies recovery actions, including sub-steps, by unit and operator to ensureadequate staffing. Communications between the units is specifically directed by plantprocedures (fire AOPs) and this process will be maintained by updated procedures.b. There are no recovery actions which credit unit cross-tie which require entry intoTechnical Specification LCO 3.0.3.SSA RAI 15:In Attachment B of the LAR, the alignment basis discussion for Attribute 3.2.1.2 provides thefollowing statement on possible fire damage to instrument air tubing that includes copper tubingwith soldered joints that are susceptible to separation during a fire and could cause the loss ofinstrument air to components:These affects were evaluated on an area basis to determine if the instrument air systempressure could be maintained. Calculation CA07971 demonstrates that the instrument airsystem can maintain system pressure with a 1 inch line pipe rupture.Calculation CA07971 states, "Evaluation Of Maximum Air Line Break Size in Which NominalInstrument Air Pressure Can Be Maintained at 50 psig." The NRC staff noted apparentdiscrepancies in the use and recovery of instrument air as described in Attachment C. Providethe following:a. Provide justification that 50 psig of instrument air pressure will not prevent instrument airoperated valves from changing position.b. Provide justification for limiting the size of the line to 1" soldered joints being susceptibleto separation during a fire. Describe the soldered joints used in the plant instrument airsystem. For any soldered joints larger than 1 ", describe how they were treated in theNSCA and Fire Probabilistic Risk Assessment (PRA).C. For several fire areas in Attachment C (such as Fire Areas 18, 19, 20, 21, and 22), thelicensee stated in the method of accomplishment for the vital auxiliaries performance goalthat instrument air may be recoverable from the opposite unit plant air system. However,the VFDRs associated with the fire areas (such as VFDRs 18-16-2, 19-01-1, 20-02-1,2 1-02-1, and 22-05-2) state that plant air from the opposite unit cannot be used becauseof failure of 1CV2061 or 2CV2061, and the VFDR? disposition credits an RA that involvesaligning backup nitrogen to the affected unit control valves. Clarify the discrepancybetween the method described in the subject fire areas for achieving the performancegoal, the VFDRs that state this method is not available, and the RAs cited in LABAttachment G for resolution of the VFDRs.d. In Attachment C, the discussion of fire suppression effects on the NSPC for Fire Areas 39and 40 addresses the impact of suppression damage to redundant instrument aircompressors and the saltwater air system, and states that the AFW air accumulators canbe charged from the nitrogen system with an RA. However, the disposition of VFDRs39-01-1 and 40-01-2, which address fire damage to the respective unit's instrument airsystem, stated that the VFDR has been evaluated with no further action required. Inaddition, the RA to align the nitrogen system to the AFW air accumulators is not discussed18 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE__________ PROTECTIONASSOCIATION STANDARD 805 _in LAR Attachment G for these fire areas. Clarify the apparent discrepancy between theeffects of fire damage and suppression damage on the instrument air system and saltwater air compressors (SWAG) with regard to the need for an RA. If an RA is necessaryto mitigate the suppression effects on the instrument air compressors and SWAG, thendescribe the feasibility and additional risk of the RA.CCNPP RESPONSE SSA RAI 15:15a -Response to be provided 4/1 3/1 5.15b -Soldered joints are only used at endpoint/load connections within the instrument airsystem. There are no soldered joints equal to or larger than 1" in the system. Any joints inthe system of this size are pipe fittings. The parts of the system that could fail are the "soft"'components connecting the instrument air distribution system to their loads. Each load thatcould be fire affected for a given fire area was reviewed. The cumulative impact of thesefailures was evaluated to determine if the system would blowdown due to the rupture beinggreater than the makeup capacity of the operable air compressors. This analysis was doneon a deterministic basis and in each case, where system blowdown could occur, it wasidentified with a VFDR.15c -Response to be provided 4/13/1 5.i15d -Response to be provided 4/1 3/1 5.SSA RAI 16:In Attachment C, the licensee stated in the summary of vital auxiliaries for Fire Area 178 thatthe control room and CSR heating, ventilating, and air conditioning (HVAC) is not availablewithout an RA and referenced VFDR 178-01-0. However, the disposition discussion for VFDR178-01-0 states that no further actions are required based on the performance-based analysisfor the VFDR, and no RAs required for risk or DID were identified in Attachment G. Clarify thebases for the discrepancy between the description of the vital auxiliaries' discussion and theVFDR disposition.CCNPP RESPONSE SSA RAI 16:The deterministic NSCA evaluation identified that the Control Room and Cable Spreading RoomHVAC System could be affected for a fire in Fire Area 1 7B. This failure was documented withVFDR 17B-01-0. The VFDR was evaluated in accordance with NEPA 805, Section 4.2.4.2,performance-based approach -fire risk evaluation with simplifying deterministic assumptions.The results of this fire risk evaluation determined that the risk, safety margin, and defense-in-depth, meet the acceptance criteria of NFPA 805 Section 4.2.4, with no further action required.A recovery action is not required for this VFDR. The Vital Auxiliary section of Fire Area 17B,Attachment C will be updated to state, 'Control Room and Cable Spreading Room HVAC maynot be available."SSA RAI 17:In .Attachment G, there are numerous RAs to provide portable fans for temporary cooling ofswitchgear rooms for Unit 1 Fire Areas 11, 16, 17, 18, and 20, and for Unit 2 Fire Areas 22, 25,19 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 80534, and yard. Plant procedures indicate the use of portable generators to power the fans ifnormal power is not available. Provide the following additional in formation:a. Describe the location of the portable generators and the location of NSCA structures,systems, and components (SSCs), if any, in the vicinity of these location(s), In yourdescription, include a summary of the procedure guidance for the use of portable gasgenerators and how the BA aligns with each of the feasibility criteria of FAQ 07-0030(i.e., training, procedures, drills, etc.).b. Describe the type of fuel and quantity associated with the portable generators and theavailability and the location(s) of sufficient fuel sources to support maintaining safe andstable conditions for the time period required.c. Provide justification that refueling the generators does not present a fire exposure hazardto NSCA SSCs.d. Describe the installation of temporary power cables, connections to distribution panels,and any disruptions to fire area boundaries.e. Describe the method (e.g., the analyzed ventilation path configuration) of providingtemporary cooling when portable fans are used for these RAs.CCNPP RESPONSE SSA RA1 17:Response to be provided 4/13/15.RAD RAI 01:The radioactive material (RAM) described in the CENG [Constellation Energy Nuclear Group]Calculation No. CA07953 provides a quantification of the maximum amount of RAM that may bestored in various areas. Provide information, if'any, on site procedures that are (or wiil be)established to limit the amount of RAM in storage containers to the levels identified in theanalyses (e.g., West Road Cage area, Warehouse #3, Pre-Assemb/y Facility, and UpperLaydown Area).CCNPP RESPONSE RAD RAi 01:Response to be provided 3/11/15.RAD RAi 02:Provide information, if any, on site procedures that establish operational controls to restrict theopening of storage containers in open, uncontained areas (e.g., West Road Cage area,Warehouse #3, Pre-Assemb/y Facility, and Upper Laydown Area).CCNPP RESPONSE RAD RAI 02:Response to be provided 3/11/15.RAD RAI 03:In the Upper Laydown Area, there are "sealed" Sealand containers, casks, and othercontainers. Describe what is meant by "~sealed" (e.g., are the containers locked and access isnot allowed, and do site procedures prevent the opening of these containers?). Also, describe20 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE________PROTECTION ASSOCIATION STANDARD 805how potential effluent will be contained based on the "sealing" of containers and concluding thatthere will be negligible RAD.CCNPP RESPONSE RAD RAI 03:Response to be provided 3/11/15.RAD RA1 04:Describe any compensatory actions that may be taken during fire suppression activities tominimize RAD (e.g., diking of liquid effluent, use of storm drain covers, radioactive monitoring,or use of other gaseous effluent controls (e.g., use of eductors, effluent filtration)).CCNPP RESPONSE RAD RAI 04:Response to be provided 3/11/15.Fire Modeling (FM) RA1 01:Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall beacceptable to the NRC. The NRC staff noted that fire modeling comprised the following:*The algebraic equations implemented in Fire Dynamics Tools (FDTs) were used tocharacterize flame radiation (heat flux), flame height, plume temperature, ceiling jettemperature, and hot gas layer (HGL) temperature, and the latter in the multi-compartment analysis (MCA).*Fire Dynamics Simulator (EDS) was used to assess MCR habitability, to calculatetemperatures and heat fluxes for damage assessment to critical targets in selectedcompartments, calculate the flame height and how that affected certain targets, andcalculate temperature rise for the purposes of estimating smoke detector activation.*The Thermally-Induced Electrical Failure model, as part of EDS, was used as a secondarycheck on the temperature and heat flux calculations using EDS for zone of influence (ZOI)purposes.Section 4.5.1.2, "Fire PRA" of the LAR states that fire modeling was performed as part of theFire PRA (FPRA) development (NFPA 805, Section 4.2.4.2). Reference is made toAttachment J, "Fire Modeling V& V,"3 for a discussion of the acceptability of the fire models thatwere used.Regarding the acceptability of the PRA approach, methods, and data:a. Identify whether any fire modeling tools and methods have been used in the developmentof the LAR that are not discussed in Attachment J. In addition, identify any fire modelingtools and methods that are discussed in Attachment J that were not used in the firemodeling analyses performed at the plant.b. It is discussed in the detailed fire modeling analysis that, "the FDTs are not setup forsecondary ignition or for the effects of suppression systems on a fire scenario." Thisimplies that secondary combustibles were not considered for any fire modeling analysis atthe plant, except those using FDS.21 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805c. Provide justification for ignoring the effects of flame spread and fire propagation insecondary combustibles (for example, cable trays) and the corresponding heat releaserate (HRR) on the calculated ZOl and HGL temperature.d. Provide in formation on how non-cable intervening combustibles were identified andaccounted for in the fire modeling analyses.e. Typically, during maintenance or measurement activities in the plant, electrical cabinetdoors are opened for a certain period of time. Explain what administrative controls are inplace to minimize the likelihood of fires involving such a cabinet, and describe howcabinets with temporary open doors were treated in the fire modeling analyses.f. Describe the criteria that were used to decide whether a cable tray in the vicinity of anelectrical cabinet will ignite following a high energy arcing fault (HEAF) event in thecabinet. Explain how the ignited area was determined and subsequent fire propagationwas calculated. If applicable, describe the effect of tray covers and fire-resistant wraps onHEA F-induced cable tray ignition and subsequent fire propagation.g. Provide justification for the assumed fire areas and elevations that were used in thetransient ZOI calculations. Explain how the model assumptions in terms of location andHRR of transient combustibles in a fire area or zone will not be violated during and post-transition.h. Explain how wall and corner effects were accounted for in the fire modeling calculations,or provide justification if these effects were not considered.i.Specifically regarding the use of FDS in the MCR abandonment calculations:I. It appears that the ceiling height of the MCR used in the calculations is rather high(- 17ft.). Explain how the MCR dimensions specified in the FDS in put files wereestablished, and confirm that they are consistent with the actual dimensions of thecontrol room. In addition, if a false ceiling is present to separate the interstitial spaceabove the operator and back panel areas, provide justification for ignoring it in thecontrol room abandonment calculations.ii. Explain if the doors of the MCR were assumed to be closed or open at all times, orwere assumed to be open at a specified time. Discuss the impact of this assumptionon the calculated abandonment times. Describe the additional leakage paths thatwere specified in the FDS input files, and provide the technical basis for the assumednatural vent areas.iii. The abandonment calculations consider two mechanical ventilation modes: HVACinoperative and HVAC in smoke purge mode. Explain why the normal HVAC modewas not considered in the analysis, and why the two modes that were considered arebounding.iv. The MCR abandonment calculations for a specified ignition source appear to includeFDS runs for 10 HRR bins. Appendix-E of NUREG/CR-6850, "EPRi/NRC-RES FirePRA Methodology for Nuclear Power Facilities: Volume 1: Summary and Overview,"September 2005 (ADAMS Accession No. ML0525800 75) uses a 15-bin discretization.Explain why only 10 bins were used, and describe how the 10-bin discretization wasestablished.v. Describe the technical basis for choosing the location of the ignition source in theelectrical cabinet and transient fire scenarios that were modeled in FDS, and confirm22 A1TrACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805that locations in the opera tar area and the back panel area were considered for bothtypes of ignition sources. Provide technical justification for not considering firescenarios with the ignition source against a wall or in a corner.vi. Explain how the area and elevation of electrical cabinet and transient fires weredetermined, and demonstrate that the assumed areas and elevations are consistentwith plant conditions or lead to conservative estimates of the abandonment times.vii. Provide justification for not considering scenarios that involve secondarycombustibles in the MCR abandonment calculations.viii. Explain how the HRRs for electrical cabinets were determined and whether thevalues are consistent with the type(s) of cabinets present in the MCR at the plant.ix. Provide technical justification for not considering electrical cabinet fires thatpropagate to adjacent cabinets.x. Provide the technical basis for the material properties that were specified in FDS forthe cables inside the cabinets in the MCR. Provide confirmation that the assumedsoot yield and heat of combustion values (the latter either explicitly or implicitlythrough the specified fuel composition) lead to conservative estimates of the sootgeneration rate.xi. Describe the transient fire growth rate(s) used in the control room abandonmentcalculations and provide the technical basis for the assumed time(s) to peak HRR.xii. Provide the technical basis for the material properties that were specified in FDS forthe transient combustibles in the MCR. Provide confirmation that the assumed sootyield and heat of combustion values (the latter either explicitly or implicitly through thespecified fuel composition) lead to conservative estimates of the soot generation rate.xiii. Describe the habitability conditions that were used to determine the time to MCRabandonment. FDS "devices" (temperature and optical density) were placed at aheight of 6 feet and at four different locations in the MCR. Describe the basis forchoosing these locations and demonstrate that these locations are eitherrepresentative of where operators are expected to be, or lead to conservativeabandonment time estimates. Confirm that heat flux sensors were not specified and,if so, provide technical justification for using temperature sensors as a surrogate forheat flux sensors.xiv. Variations in the input parameters such as ambient temperature, soot yield of the fuel,fire base height, etc., affect the output of FDS calculations. The abandonmentanalyses for the MCR were performed using a single set of input parameters for eachscenario. Demonstrate that the FDS calculations obtained using this set of inputparameters provide conservative or bounding results. Alternatively, demonstrate thatthe abandonment times for a given scenario are not sensitive to variations within theuncertainty of the input parameters.xv. Explain how the results of the MCR? abandonment time calculations were used in theFPRA.j. Specifically regarding the MCA:i. Describe the criteria that were used to screen multi-compartment scenarios based onthe size of the exposing and exposed compartments.23 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STAN DARD 805ii. Explain how the methods described in Chapter 2 of NUREG-1805, "Fire DynamicsTools (FDTs): Quantitative Fire Hazard Analysis Methods for the U.S. NuclearRegulatory Commission Fire Protection Inspection Program," December 2004(ADAMS Accession No. ML043290075) were used in the calculations to screen anignition source based on insufficient HRR to generate a HGL condition in theexposing compartment. In addition, clarify which FDTs were used for the HGLcalculations.iii. In the MCA scenario analysis, explain the technical basis of modeling the ZOI as avertical cylinder with the radius equal to 0.2 times the ceiling height in scenarioswhere the fire occurs near the opening between the two compartments and damagesitems on both sides within its ZOI.iv. Some of the EDT calculations make the following assumption: "It is assumed that theforced ventilation of air flow rate is distributed among the interconnectedcompartments, especially corridors, based on the volume of the compartments."Provide technical justification for this assumption.v. The screening process based on the ZOl specifies that if there are cable trays,conduits, or targets on the exposed side of the barrier within the ZOI, which may notbe the same as those inside the exposing compartment, the scenario should beanalyzed further. Provide details about this additional analysis.k. Specifically regarding the use of FDS in the CSR (physical analysis units (PA Us) 306 and302) calculations:i.It is stated that engineering judgment is used to assess that the delay in smokedetector activation, which is associated with cross-train logic that is not possible toincorporate in FDS, would be in the range of 2 to 10 seconds. Provide technicaljustification for this estimate.ii. The FDS "devices" (temperature and heat flux) were placed at different locationsaround the switchgear rooms. Describe the basis for choosing these locations.iii. The analysis highlights the location of possible electrical cabinet fires that wereconsidered. Provide technical justification for selecting these specific fire locations ordemonstrate that these locations lead to bounding or conservative estimates.iv. A number of transient fires were postulated in the CSRs, but the documentationindicates that the walkdown identified no transient combustibles and there were nostorage areas for more permanent combustibles in the fire areas, Provide justificationfor selection of the transient fire areas and indicate if this selection is dependent onany administrative controls of transient combustibles in the CSRs.v. The HRR used for the cabinet fires indicates that the cabinet doors were assumed tobe closed. Provide justification for this assumption (e.g., on the basis of the actualplant configuration or operational condition).vi. As stated in FM RAI 1.b, it is expected that secondary combustibles (ignition, flamespread, and cable tray fire propagation) would be part of the FDS analysis for theCSRs. Clarify how secondary combustibles were considered in the FDS analysis ofthe CSRs, and if they were not considered, provide justification for their omission.I. During the walkdown of the MOR, several observations were made, which requireadditional information:24 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805I.The main horseshoe and back panel cabinet configurations consist of open cabinetswith a steel mesh open top with the open sides facing each other across a narrowaisle. The EDS analysis utilizes an HRR case from Appendix G of NUREG-CR 6850,which assumes closed cabinets. Provide justification for not using an HRR caseapplicable to open cabinets or update the analysis with the appropriate HRR.ii. During the discussion about the open cabinets, it was also discussed that the currentanalysis does not consider the potential for fire spread across the aisle (i.e., within thehorseshoe) from the front to back or vice versa. Provide justification for notconsidering this potential fire spread or update the analysis to include this scenario.iii. During the walkdown of the MCR, several combustible items, which could beconsidered transient fire sources, were observed that could potentially have an/-HRRof greater than 317 kW. Examples include the kitchen area, the upholstered furniturein the shift manager's office and space below the shift manager's office, andphotocopiers. Provide additional information that can justify that the transient firesource selected in the FDS analysis is conservative and bounding.CCNPP RESPONSE FM RAI 01:Response to be provided 4/13/15.011.i -Response to be provided 3/11/1 5.FM RA! 02:The ASME/ANS Standard RA-Sa-20 09, "Standard for Level 1/Large Early Release FrequencyProbabilistic Risk Assessment for Nuclear Power Plant Applications," Part 4, requires damagethresholds be established to support the FPRA. Thermal impact(s) must be considered indetermining the potential for thermal damage of SSCs and appropriate temperature and criticalheat flux criteria must be used in the analysis.a. Describe how the installed cabling in the power block was characterized, specifically withregard to the critical damage threshold temperatures and critical heat fluxes for the rmosetand thermoplastic cables as described in NUREG/CR-6850. If thermoplastic cables arepresent, explain how raceways with a mixture of thermoset and thermoplastic cables weretreated in terms of damage thresholds.b. Explain how the damage thresholds for non-cable components (i.e., pumps, valves,electrical cabinets, etc.) were determined. Identify any non-cable components that wereassigned damage thresholds different from those for thermoset and thermoplastic cables,and provide a technical justification for these damage thresholds.c. Explain how exposed temperature-sensitive equipment was treated, and provide atechnical justification for the damage criteria that were used.CCNPP RESPONSE FM RA! 02:02a -Response to be provided 4/13/15.02b -Response to be provided 3/11/15.02c -Response to be provided 4/13/15.25 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805FM RAI 03:Section 2. 7.3.2 of NFPA 805 states that each calculational model or numerical method usedshall be *verified and validated through comparison to test results or comparison to otheracceptable models.Section 4.5.1.2 of the LAR states that fire modeling was performed as part of the FPRAdevelopment (NFPA 805, Section 4.2.4.2). Reference is made to Attachment J for a discussionof the verification and validation (V& V) of the fire models that were used. Furthermore,Section 4.7.3 of the LAR states that "Calculation models and numerical methods used insupport of complance with 10 CFR 50.48(c) were verified and validated as required bySection 2.7.3.2 of NFPA 805."For any tool or method identified in the response to FM RAI 1.a above, provide the V& V basis ifnot already explicitly provided in the LAR (for example, in Attachment J). Provide technicaldetails to demonstrate that these models were applied within the validated range of inputparameters, or justify the app lication of the model outside the validated range in the V& V basisdocuments.CCNPP RESPONSE FM RAI 03:Response to be provided 3/11/15.FM RAI104:Section 2.7.3.3 of NFPA 805 states that acceptable engineering methods and numerical modelsshall only be used for applications to the extent these methods have been subject toverifications and validation. These engineering methods shall only be applied within the scope,limitations, and assumptions prescribed for that method.Section 4.7.3 of the LAR states that, "Engineering methods and numerical models used insupport of compliance with 10 CFR 50.48(c) were applied appropriately as required bySection 2.7.3.3 of NFPA 805."Regarding the limitations of use, the NRC staff notes that algebraic models cannot be usedoutside the range of conditions covered by the experiments on which the model is based.NUREG-1805 includes a section on assumptions and limitations that provides guidance to theuser in terms of proper and improper use for each FDT.Identify uses, if any, of FDS and the FDTs outside the limits of applicabilty of the model, and forthose cases, explain how the use of FDS and the FDTs was justified.CCNPP RESPONSE FM RAI 04:Response to be provided 4/13/15.FM RAI 05:Section 4.5. 1.2 of the MAR states that fire modeling was performed as part of the FPRAdevelopment (NFPA 805, Section 4.2.4.2). The NRC staff notes this requires that quailified firemodeling and PRA personnel work together. Furthermore, Section 4. 7.3 of the LAB states thefollowing:26 A1TACHM ENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FiREPROTECTION ASSOCIATION STANDARD 805Cognizant 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.For personnel performing fire modeling for FPRA development and evaluation, CCNPP[Calvert Cliffs Nuclear Power Plant] develops and maintains quailification requirements forindividuals assigned various tasks. Position specific guides were developed to identify anddocument required training and mentoring to ensure individuals are appropriately qualifiedper the requirements of NFPA 805, Section 2.7.3.4, to perform assigned work.Qualification cards provide evidence that Design Engineering and PRA personnel have theappropriate training and technical expertise to perform assigned work, including the use ofengineering analyses and numerical models.Qualification requirements are contained in procedure CNG-TR-1.01-1014 (Reference 6. 47).CCNPP will maintain qualification requirements for the performance of NFPA 805 relatedtasks. Position specific qualification cards identify and document required training andmentoring to ensure cognizant individuals are appropriately qualified to perform assignedwork per the requirements of NFPA 805, Section 2. 7.3.4.Regarding qualifications of users of engineering analyses and numerical models (i.e., firemodeling techniques):a. Describe the requirements to qualify personnel for performing fire modeling calculations inthe NFPA 805 transition.b. Describe the process for ensuring that fire modeling personnel have the appropriatequalifications not only before the transition, but also during and following the transition.c. When fire modeling is performed in support of the FPRA, describe how propercommunication between the fire modeling and FPRA personnel is ensured.CCNPP RESPONSE FM RAI 05:05a -Fire modeling calculations were performed by engineers who meet the qualificationrequirements of Section 2.7.3.4 of NEPA 805. The qualification process through December2014 followed the guidance of ACAD 98-004, "Guidelines for Training and Qualification ofEngineering Personnel," and the CENG procedure on "Conduct of Training." All thoseperforming Fire Modeling for the Fire PRA were qualified and their qualifications weredocumented in the CENG training database. This qualification includes basic fire modelingtechniques as well as Fire PRA techniques. The CENG PRA Engineering Supervisorreviewed experience and education for all fire modeling work. Those performing detailed firemodeling analysis using tools such as CFAST (Consolidated Model of Fire and SmokeTransport) or FDS (Fire Dynamics Simulator) were required to have the relevantqualifications and experience in fire modeling to perform the analysis.In the case of the initial fire modeling, the vendor provided the credentials of the firemodelers, which were reviewed and approved by Risk Management Supervision. Duringand following transition, the existing engineering staff will continue to be knowledgeable infire modeling techniques, including interpreting and maintaining the fire modeling database.If new fire modeling personnel are needed in the future, their credentials will also bereviewed and approved by Exelon supervision. Currently the Risk Management organizationhas transitioned to Exelon qualification processes which include the Fire PRA qualification.27 ATTACHM ENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805Engineering Supervisors have responsibility for "Verifying qualifications prior to assigningpersonnel to perform job performance requirement independently." This requires reviewingthe qualifications in the training server to verify the qualification is current.05b -Response to be provided 3/11/15.05c -Response to be provided 3/11/15.FM RAI 06:Section 4.7.3 of the LAR states that, "Uncertainty analyses were performed as required bySection 2.7.3.5 of NFPA 805 and the results were considered in the context of the application.This is of particular interest in fire modeling and FPRA development."Regarding the uncertainty analysis for fire modeling:a. Describe how the uncertainty associated with the fire mode! input parameters wasaccounted for in the fire modeling analyses.b. Describe how the "model" and "completeness" uncertainties were accounted for in the firemodeling analyses.CCNPP RESPONSE FM RAI 0]6:Response to be provided 4/13/15.PRA RAI 01 -Fire Event Facts and Observations:Section 2.4.3.3 of NFPA 805 states that the probabilistic safety assessment (also referred to asPRA) approach, methods, and data shall be acceptable to the authority having jurisdiction,which is the NRC. RG 1.2 05 identifies NUREG/CR-6850 as documenting a methodology forconducting an FPRA and endorses, with exceptions and clarifications, NE! 04-02, Revision 2, asproviding methods acceptable to the NRC staff for adopting a fire protection program consistentwith NEFA 805. RG 1.200 describes a peer review process utilizing an associated ASME/ANSstandard (currently ASME/ANS-RA-Sa-2009) as one acceptable approach for determining thetechnical adequacy of the PRA, once acceptable consensus approaches or models have beenestablished for evaluations that could influence the regulatory decision. The primary result of apeer review are the facts and observations (F& Os) recorded by the peer review and thesubsequent resolution of these F&Os.Clarify the following dispositions to fire F&Os and Supporting Requirement (SR) assessmentsidentified in Attachment V of the LAR that have the potential to impact the FPRA results and donot appear to be fully resolved:a) PRM-83-01: The disposition to F&O PRM-83-01 appears to indicate that the EPRA wasupdated to address events involving a fire induced loss of MCR HVAC, which the peerreview suggests has a conditional core damage probability (CCDP) of 1.0, by increasingthe likelhood of functional failures in lieu of assuming their occurrence. Justify thefunctional failures modeled by the FPRA to address this loss of MCR HVAC. In addition,explain how the FPRA evaluates the degradation of equipment due to elevatedtemperatures caused by loss of HVAC as an increase in equipment failure rates, andprovide a technical basis for doing so.28 A1TACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805b) FSS-A5-01: This F&O states that some PAUs are further divided into "sub-PA Us" andappears to indicate that there is no explicit process for evaluating the fire spread acrosssub-PA U boundaries, which, as the peer review noted, are not defined by physicalbarriers. The disposition to this F&O, however, does not discuss such a process, and byreferencing a sensitivity analysis limited to a number of "representative" PAUs, suggeststhat this apparent deviation from acceptable methods has not been fully addressed for allPAUs for which sub-PA Us have been defined. Explain how the fire effects across non-physical sub-PA U boundaries are identified and evaluated. Discuss how this approach isconsistent with or coniservatively bounds acceptable methods.c) FSS-G4-01: The disposition to this F&O indicates that the MCA did not postulate apropagation scenario if doing so would require failure of a penetration seal. Thelicensee's analysis (CO-FSS-08) suggests that a similar approach may have also beenfollowed for other barrier types (e.g., walls). As a result, identify each barrier type forwhich propagation scenarios were not postulated, and provide quantitative justification(e.g., an evaluation demonstrating that MCA scenarios involving barrier failure are lowrisk, even considering the risk associated with the multi-compartment fire) for notaddressing propagation. As an alternative, provide updated risk results as part of theintegrated analysis requested in PRA RAI 03, summing the generic barrier failureprobabilities for each type of barrier present between communicating compartments,consistent with NUREG/CR-6850.d) FSS-G5-01: The disposition to this F&O indicates that unreliability values were applied toall normally open, self-closing dampers and doors; however, the disposition neitherprovides a basis for the values applied nor mentions active elements discussed elsewhere(e.g., water curtains in F&O PP-B5-0 1). Summarize the types of active fire barrierelements credited in the EPRA, and provide quantitative justification for their unreliabilityand unavailability.e) HRA-B2-01: The disposition to this F&O indicates that "adverse" operator actions, whichinclude actions to de-energize electrical busses as a means to address spuriousoperations, are modeled in the FPRA by assuming all equipment disabled by the action isfailed (i.e., the action is successful). Although the licensee's analysis (Section 2.2 ofCO-HRA -001) indicates that this assumption is conservative, the basis for this conclusionis unclear if the action is taken to reduce risk. In light of this:i. Provide justification for the assumption that modeling "adverse" actions as successfulis conservative. Note that guidance in NUREG-1921 offers considerations forevaluating fault clearing strategies in the FPRA human reliability analysis (HRA).ii. Clarify how "adverse" actions are addressed by the FPRA HRA dependency analysis,given that these actions are modeled by failing associated equipment directly within thePRA logic model.iii. Explain the statement in Attachment G that "[nione of the recovery actions were foundto have an adverse impact on the EPRA." In doing so, clarify how "adverse" riskimpact was defined. Note that FAQ 07-0030 states that "[i]f activities (recovery actionsor other actions in the post-fire operational guidance) are determined to have anadverse risk impact, they should be resolved during NFPA 80.5 implementation via analternate strategy that eliminates the need for the action in the NSCA. "f) CS-B1-01: The licensee's analysis (Appendix F of ECP-13-O00321, "Common PowerSupply and Common Enclosure Study') identifies several MCC 208/120 Volts alternating29 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805current load breakers that were not coordinated with their respective feed breakers. Thedisposition to this F&O indicates that these 120V panel breaker coordination issues are tobe addressed by plant modification; however, Attachment S does not appear to containsuch a modification. Identify the Attachment S modification(s) being credited to resolvethe 120V panel breaker coordination issues identified in the disposition to this F&O.CCNPP RESPONSE PRA RAI 01:01a -Modeling of the loss of control room HVAC impact was developed for internal events priorto the NFPA 805 application. The Calvert Cliffs control room and cable spreading roomsshare the common control room HVAC system. A GOTHIC thermal hydraulic analysis of theheat-up following a loss of control room HVAC determined the maximum temperature in therooms is 1 12°F. The design temperature limit is 104°F and the normal control roomtemperature is considered to b~e 720F. Given the maximum temperatures are slightly abovethe design temperature, the failure rates of the equipment were increased rather thanassuming a complete failure.The basis for the failure rate increases is IEEE 500, "Guide to the Collection andPresentation of Electrical, Electronic, Sensing Component, and Mechanical EquipmentReliability Data for Nuclear-Power Generating Stations." IEEE 500 listed a range of potentialfailure rate increases for various equipment types. The maximum recommend increase wasselected from among all equipment types in the control room and cable spreading rooms.Since all of the control room and cable spreading room controls and instrumentation aresupplied from 125VDC and/or 120VAC buses, the failure likelihood of these power supplieswas increased by the maximum recommended increase over the mission time of theventilation loss.01ib -Response to be provided 4/1 3/1 5.01c -Response to be provided 4/13/15.O1d -Normally open fire dampers are not considered active fire barriers in the CCNPP FPRA.The CCNPP FPRA does not model any normally open doors with closures which initiate doorclosure due to fire.The following failure probabilities were used for fire barriers that were credited in the CCNPPFPRA:For fire dampers the probability of failure 2.70E-03 was used based on the suggestedvalues in NUREG/CR-6850 Table 11-3.SFor doors, including watertight doors, the probability of failure 7.40E-03 was used basedon the suggested values in NUREG/CR-6850 Table 11-3. It may be noted that thewatertight doors included in the scenarios are used on a regular basis to enter and exitthe compartments. Also, the probability values used is assumed to include thepossibility of finding a door being propped open given a fire in the exposingcompartment.* For credited, installed water curtains, the fire barrier failure probability was based on anon-suppression probability (NSP) derived per the guidance in NUREG/CR-685030 AI-FACHM ENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805Appendix P, as suggested in NUREG/CR-6850 Section 11.5.4.4. Installed watercurtains were credited in the CCNPP FPRA for the following potential multi-compartmentfires:o Fires that could potentially propagate from PAU 101 (Unit 2 ECCS Pump Room) toPAU 1 20E (Unit 2 Containment Recirculation Pipe Tunnel -East)o Fires that could potentially propagate from PAU 102 (Unit 2 ECCS Pump. Room) toPAU 120W (Unit 2 Containment Recirculation Pipe Tunnel -West)o Fires that could propagate from PAU 118 (Unit 1 #12 EGOS Pump Room) to PAU122E (Unit 1 Containment Recirculation Pipe Tunnel -East)o Fires that could propagate from PAU 119 (Unit 1 #11 ECCS Pump Room) to PAU122E (Unit 1 Containment Recirculation Pipe Tunnel -East)Plant specific failure data was not collected to verify these probability values; however,procedures are implemented that ensure integrity of the active fire barriers that are creditedin the CCNPP fire protection program, including the active fire barriers that have beencredited in the CCNPP FPRA. The active fire barriers that have been credited in the CCNPPEPRA are, therefore, inspected periodically per plant procedures. Also, the active firebarriers that have been credited in the CCNPP FPRA that are relied upon to maintain safety /separation have established compensatory measures that are put in place whenever anissue is discovered with the credited feature.01e -i.There are both positive and negative aspects to "adverse" operator actions. On thepositive side, a successful "adverse" operator action will preclude a spurious actuationthat could otherwise have negative consequences. On the negative side, an "adverse"operator action disables equipment that may be credited in the CCNPP FPRA. Whenthe "adverse" operator action is assumed to be successful, the negative impact mustalso be assumed (i.e., the associated mitigation function credited in the CCNPP FPRAmodel must be assumed to be failed). When an "adverse" operator action is credited, adetailed human reliability analysis (HRA) quantification is required. This was themethodology employed in the CCNPP FPRA HRA. As an example, when the CCNPPFPRA HRA credited a procedure to de-energize a valve such as a PORV to prevent aspurious opening which can lead to a loss of coolant accident (LOCA) scenario, suchcredit was only applied following a detailed HRA quantification, it also always assumedthe negative impact that the valve was not available to energize to support a feed andbleed type function (the negative impact is incorporated by, setting the associated basicevent, failure to open the pilot operated relief valve (PORV) in this case, to TRUE in thequantification, such that the risk of core damage or large early release frequency (LERF)increases).i.The CCNPP FPRA will model all equipment disabled by the "adverse' action as beingfailed (i.e., the action is successful and, as a consequence, equipment is disabled);therefore, this is not evaluated in the dependency evaluation. If an "adverse" action iscredited to prevent spurious actuation, then a detailed human reliability analysis (HRA) isdeveloped. All of the detailed HRAs are evaluated in the dependency evaluation.31 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805iii. A recovery action is considered to have an adverse impact on the CONPP FPRA if itcauses more than a negligible increase in core damage frequency (CDF) or large, earlyrelease frequency (LERF). Through implementation of Item 15 in LAR Table S-3,CCNPP will ensure that there is not an adverse impact on the CON PP FPRA.Implementation Item 15 in LAR Attachment S table S-3 includes updating the AbnormalOperating Procedures (AOPs) for severe fires for the recovery actions evaluated. Aspart of this effort, actions are either functionally based actions or direct procedure stepsthat are always required. Functional actions are only implemented as required tosupport a functional loss. For example, if 4kV Bus 21 cannot be re-powered, then action11 I0PEN4KVBKRS in Table G-1 would be implemented. In other cases, the actions areonly conditional on some fire-related condition such as the MCR being abandoned. Thenegative aspects of these actions will always be assumed to occur when the fire-relatedcondition occurs. Although some of the directly implemented actions may have bothnegative and positive impacts, overall recovery actions are only credited when they arejudged to be risk beneficial. In support of Implementation Item 15, the actions that areultimately credited in the AOPs for severe fires will be risk beneficial or will include thequantification of the adverse impact of the actions in the FPRA quantification.All risk adverse actions will be removed from the AOPs for severe fires except thoserequired due to operational concerns. Risk adverse actions that are required due tooperational concerns will be appropriately modeled in the CON PP FPRA.01f -While Appendix F of ECP-13-000321, "Common Power Supply and Common EnclosureStudy," does state that several MOO 208/120 VAC load circuit breakers were not coordinatedwith their respective feed circuit breakers; a later, more detailed, analysis documented inECP-1 3-000776 concluded that those same power supplies did not have any coordinationissues. A plant modification is not, therefore, required.PRA RAi 02 -Internal Event F&Os: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 an FPRA and endorses, with exceptions and clarifications, NEI 04-02,Re vision 2, as providing methods acceptable to the staff for adopting a fire protection programconsistent with NFPA 805. RG 1.200 describes a peer review process utilizing an associatedASME/ANS standard (currently ASME/ANS-RA-Sa-2009) as one acceptable approach fordetermining the technical adequacy of the PRA once acceptable consensus approaches ormodels have been established. The primary results of a peer revieW are the F&Os recorded bythe peer review and the subsequent resolution of these F&Os.Clarify the following dispositions to internal events F&Os and SR assessments identified inAttachment U of the LAR that have the potential to impact the EPRA results and do not appearto be fully resolved:a) 4-5: This F&O indicates that the alignment strategy assumed by the PRA for the OC dieselgenerator (DG) is not appropriately justified and may be non-consen'ative. While thedisposition to this F&O clarifies how alignment of the CC DG is modeled in the PRA, ajustification for this treatment is not provided. Provide a technical and/or procedural basisfor the alignment strategy assumed in the PRA for the CC DG, and indicate whether anyoperator interviews were conducted to support the analysis.32 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE_________PROTECTION ASSOCIATION STANDARD 805b) 6-23: This F&O indicates that some joint human error probabilities (HEPs) applied withinthe internal events PRA (IEPRA) may not accurately reflect the sequential timing ofassociated operator actions. While the disposition appears to address the specificexample referenced by the F&O, it is not clear that the broader issue has been fullyresolved in the fire PRA, particularly noting that the status of this F&O in Table U-i isidentified as "open."i. Explain how the HRA methods used by the EPRA for developing HEP and joint HEPvalues are consistent with or conservatively bound NRC-accepted guidance inNUREG/CR-6850 or NUREG- 1921. Alternatively, provide updated risk results as partof the aggregate change-in-risk analysis requested in PRAN RAI 03 applying HEPand joint HEP values developed using NRC-accepted guidance.ii. NUREG-1921 indicates and NUREG-1792 (Table 2-1) states that joint HEP valuesshould not be below 1. OE-05. Confirm that each joint HEP value used in the FPRAbelow 1.0E-05 includes its own justification that demonstrates the inapplicability of theNUREG- 1792 lower value guideline. Provide an estimate of the number of these jointHEPs below 1.0E-05 and at least two different types of justification.CCNPP RESPONSE PRA RAI 02:a) The alignment of the 0C Diesel Generator (DG) is not fixed. in the AOPs for severe fires, the00 DG is aligned to both 4kV Buses 11 and 24. Depending on the circumstance, Operationspersonnel will align the 00 DG to the location where the most equipment can be restored.4kV Buses 11 and 24 support a motor driven Auxiliary Feedwater (AFW) pump and MCRheating, ventilation and air conditioning (HVAC). 4kV Buses 14 and 21 support a similar setof loads. As such, in many circumstances those buses would be the logical choice. There isnothing, however, to prevent a re-alignment if the equipment being powered from the 00 DGfails or is not otherwise satisfying operational needs. To prevent excessive modelcomplexity, a fixed alignment strategy is used which is conservative compared to the realityof the flexible alignment strategy.Additional operator interviews regarding alignment of the 00 DG will be conducted.Summaries of these interviews will be added to the CCNPP FPRA documentation. TheCCNPP FPRA documentation will also be revised to capture simulator observationsregarding alignment of the 00 DG. As revisions are drafted for the AOPs for severe fires, theimpacts on the 0C DG alignment modeling will be considered to ensure appropriate modelingfor the expected post transition configuration. These changes will also be discussed duringinterviews with operators to ensure realistic modeling is performed.The revised documentation containing the additional operator interviews described above willbe will be generated in conjunction with the update of CCNPP FPRA analysis documentationsupporting RAI PRA-03.b)i. As noted in Attachment U of the LAR, human failure event (HFE) timelines were reviewed.Some events in the CCNPP FPRA human reliability assessment (HRA) were split intomultiple HFEs, where appropriate, to account for the different scenarios and to ensure thatthe sequential timing of the associated operator actions is appropriate. The CCNPP FPRAHFEs were then assessed as part of the dependency analysis using the EPRI's HRA33 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805Calculator. This is considered an acceptable HRA method for developing HEP and joint HEPvalues.To confirm the appropriateness of the review mentioned in Attachment U of the LAR anadditional step was taken: A review all CCNPP FPRA HEPs with a Tdeiay of zero wasundertaken. Tdelay represents the period between the start time of an event and the time theoperator acknowledges the cue. All HEPs with a Tdelay of zero were reviewed to determine ifthese HEPs were discussed during interviews with CCNPP Operations personnel. Operatorinterviews provide valuable insights and understanding of plant-specific crew responses andfeasibility. It was determined that all HEPs with a Tdelay of zero were discussed duringoperator interviews conducted either during the Internal Events PRA's Peer Review or theFPRA's Peer Review, or as part of the original Internal Events PRA's HRA. it should benoted that, while all HEPs were discussed during these interviews, only a very small fractionof the HEPs were found to have a Tdelayof zero that might not be appropriate.Changes to the Tdalay could increase or decrease the risk. As the number of potentiallyaffected HEPs in the CCNPP FPRA is small as compared to the total number of HEPs in theCCNPP FPRA, any such changes to the Tdelay would likely have a negligible impact on theCCNPP FPRA results.b)ii. None of the individual HEPs included in the CCNPP FPRA have a value less than1 .OOE-05. The CCNPP FPRA evaluated 2700 joint human error probabilities (JHEPs); 2259of these 2700 JHEPs are below 1 .O0E-05.The documentation provided in the EPRI HRA Calculator justifies each unique HEP value,including the unique values used for all JHEPs. The evaluations of the individual HEPsinclude common cognitive actions where appropriate. JHEPs are only developed when thereis not a common cognitive failure mode.Justification examples of JHEPs CombinationU1_1770 and CombinationU1_1065 aredescribed below:CombinationU1_1770- 4.23E-090FW00HFMPZ9-FR -1 .9E-02 -Operations fails to control Main Feedwater flow post-tripto prevent Steam Generator overfill given fire-induced plant triplrn~act: SG overfill fails the running turbine driven Auxiliary Feedwater pumpLocation: Main Control Room*CSTOHF-DEPLETION-FR -6.00E-05 -Operations fails to detect Condensate StorageTank level dropping during fire (common cognitive)Impact: Failure of AFW pumpsLocation: MCR*CVCOHFOTA8HRS-FR -7.1E-03 -OTCC -All AFW / MFW failed after CST depletionduring fireImpact: Failure of once through core cooling and all AFW start actionsLocation: MOR34 ATrACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805All of the actions in combination U1_1770 are treated independently largely due to timeseparation. Operations fails to secure the MFW pumps prior to a SG overfill (FW00HFMPZ9-FR) this fails the running TO AFW pump. The minimum time to CST depletion is six hoursfollowing the trip. Although this is the time used in the HRA evaluation with a TO AFW pumpfailed, this would be longer. Due to this large time separation and different cues (SG waterlevel and CST water level), these actions are considered independent. If Operations fails torecognize that the condensate storage tank is depleting (CSTOHF-DEPLETION-), then all therunning AFW pumps fail. Operations could start the other unit's AFW pump or align forOTOC (CVCOHFOTA8HRS-FR). There is at least two hours available to align following aCST depletion AFW failure. A CST depletion failure in combination with an AFW failurewould provide even more time. Due to the large time separation and different cues (SGwater level and CST water level), this is also considered to be independent.CombinationUl_1065- 1.38E-07*AFW0HFCCSGDEC-FR -1.2E-03 -Operations fails to diagnose SG level decreasingduring fire (common cognitive)Impact: Failure of OTOC and AFW start actionsLocation: MCR*CA00HFN2C8-FRI -7.80E-04 -Operations fails to start both Salt Water AirCompressors (SWACs), in Emergency Operating Procedure (EOP) 8, no loss of airannunciators, dual unit trip during fireImpact: Loss of air to AFW flow control valves. AFW delivers full flow to the SGs.Location: MCR*AFWOHFHXB-FR -5.6E-3 -Operators fail to control AFW flow during fire, no CR flowsupport, EOP-8Impact: SG overfill fails the running TD AFW pumpLocation: AFW Pump RoomAction AFW0HFHXB-FR and AFW0HFCCSGDEC-FR are considered independent whileactions CA00HFN2C8-FRI and AFWOHFHXB-FR are considered a medium dependency (perTHERP, maximum credit for a medium action is 0.14). Operations fail to start both SWACsand provide long term air for AFW before the AFW accumulators deplete (CA00HFN2C8-FR I). Once air is lost, Operations has a limited time to prevent overfill (AFWOHFHXB-FR); assuch, this action has a medium dependency with CAO0HFN2C8-FRI. Following a late overfillevent caused by AFW accumulator depletion, Operations has over two hours to re-establishAFW flow. Due to the large time separation, these actions are considered independent.It should be noted that for these joint events to occur, not only does the whole Operationscrew need to fail, but the whole emergency response organization must fail as well. Foractions to be independent there needs to be at least 60 minutes of time separation betweenthe cues for the Operations actions as well as no common cognitive function. Any fire thatprogresses to core damage must affect multiple redundant groups of safety relatedequipment. The Calvert Emergency Action Level procedure requires us to declare an "Alert"when:FIRE or EXPLOSION resulting in EITHER:VISIBLE DAMAGE to ANY SAFETY-RELATED35 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805STRUCTURE, SYSTEM, OR COMPONENT within ANYTable H-i areaORControl Room indication of degraded performance of ANYSAFETY-RELATED STRUCTURE, SYSTEM, ORCOMPONENT within ANY Table H-i areaTable H-i areas include: MCR, Containment, Auxiliary Building, Diesel Generator Rooms,Intake Structure, 1A/OC Diesel Generator Buildings, Refueling Water Tank (RWT), RWTRooms, CST No. 12, Fuel Oil Storage Tank (FOST) No. 2 and AFW Pump Rooms.As a fire that progresses to core damage always affects safety related equipment in theseareas, a core damage or large early release fire will always result in the activation of theemergency response organization. The Technical Support Center is required to beoperational within 60 minutes of activation.PRA RAI 03 -Inteqrated Analysis: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.RG 1.174 provides quantitative guidelfines on ODE and LERF, identifies acceptable changes tothese frequencies that result from proposed changes to the plant's licensing basis, anddescribes a general framework to determine the acceptability of risk-informed changes. TheNRC staff's review of the information in the LAR has identified additional information that isrequired to fully characterize the risk estimates.The PRA methods currently under review in the LAB include:* PRA RAI 01.a regarding loss of MOR HVAC* PRA RAI 01.b regarding division of PAUs into "sub-PA Use'*PRA RA! 01.c regarding treatment of propagation in the MOA* PRA RA! 01. d regarding unreliability and unavailability of active barriers* PRA RAI 01.e regarding adverse operator actions* PRA RAI 01. f regarding 120V panel breaker coordination issues* PRA RAI 02.a regarding alignment of 00 diesel generator* PRA BAI 02.b regarding HRA methods, including sequential timing of operator actions* PRA RAI 04 regarding placement of transient fires* PRA RAI 05 regarding transient influence factors* PRA RA! 06 regarding reduced transient HRR* PRA BA! 07 regarding self-ignited cable fires and those caused by welding and cutting* PRA RAI 08 regarding treatment of junction boxes* PRA RAI 09 regarding treatment of sensitive electronicso PRA BA! 10 regarding circuit failure probabilities* PRA BA! 11 regarding counting and treatment of Bin 15 electrical cabinets* PRA BAI 12 regarding treatment of HEAF* PRA BAI 13 regarding MOB modeling* PRA BAI 14 regarding credit for MOR abandonment actions* PRA BA! 15 regarding MOB abandonment on loss of control36 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805_____* PRA RAI 16 regarding application of the state-of-knowledge correlation (SOKC)* PRA RA! 18 regarding ,4ICDF, ALERF and additional risk of RAs* PRA RA!123 regarding other deviations from acceptable methodsProvide the following:a) Results of an aggregate analysis that provide the integrated impact on the fire risk(i.e., the total transition CDF, LERF, ,4CDF, ALERF, and additional risk of RAs) ofreplacing specific methods identified above with alternative methods that are acceptableto the NRC. In this aggregate analysis, for those cases where the individual issues havea synergistic impact on the results, a simultaneous analysis must be performed. For*those cases where no synergy exists, a one-at-a-time analysis may be done. For thosecases that have a negligible impact, a qualitative evaluation may be done. It should benoted that this list may change depending on NRC's review of the responses to otherRAIs in this document.b) For each method (i.e., each bullet) above, explain how the issue will be addressed in1) the final aggregate analysis results provided in support of the LAR, and 2) the PRA thatwill be used at the beginning of the self-approval of post-transition changes. In addition,provide a process to ensure that all changes will be made, that a focused-scope peerreview will be performed changes that are PRA upgrades as defined in the PRAstandard, and that any findings will be resolved before self-approval of post-transitionchanges.c) in the response, explain how RG 1.205 risk acceptance guidelines are satisfied for theaggregate analysis. Additionally, discuss the likelihood that the risk increase in anyindividual fire area would exceed the acceptance guidelines, and if so, why exceeding theguidelines should be acceptable. If applicable, include a description of any newmodifications or operator actions being credited to reduce delta risk as well as adiscussion of the associated impacts to the fire protection program.d) If any unacceptable methods identified above will be retained in the PRA and will be usedto estimate the change in risk of post-transition changes to support self-approval, explainhow the quantification results for each future change will account for the use of thesemethods.CCNPP RESPONSE PRA RAI 03:Response to be provided 4/13/15.PRA RAl 04 -Transient Fire Placement at Pinch Points: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 an FPRA and endorses, with exceptions and clarifications, NE/ 04-02,Revision 2, as providing methods acceptable to the staff for adopting a fire protection programconsistent with NFPA 805. Methods that have not been determined to be acceptable by theNRC staff, or acceptable methods that appear to have been applied differently than described,require additional justification to allow the NRC staff to complete its review of the proposedmethod.The NRC staff could not identify in the LAR or licensee's analysis a description of how "pinchpoints" for transient fires were treated in the FPRA. Per NUREG/CR-6850, Section 11.5.1.6,37 A'1-ACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805transient fires should, at a minimum, be placed in locations within the plant PAUs where CCDPsare highest for that PAU (i.e., at "poinch points"). Pinch points include locations of redundanttrains or the vicinity of other potentially risk-relevant equipment. Cable congestion is typical forareas like the CSR, so placement of transient fire at pinch points in those locations is important.Hot work should be assumed to occur in locations where hot work is possible, even ifimprobable, keeping in mind the same philosophy.a) Clarify how "pinch points" were identified and modeled for general transient fires andtransient fires due to hot work.b) Describe how general transient fires and transient fires due to hot work are distributedwithin the PAUs at Calvert Cliffs. In particular, identify the criteria used to determinewhere such ignition sources are placed within the PA Us.CCNPP RESPONSE PRA RAI 04:Response to be provided 4/13/15.PRA RAI 05 -Transient influencinq Factors:Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall beacceptable to the NRC. The 1.205 identifies NUREG/CR-6850 as documenting a methodologyfor conducting an FPRA and endorses, with exceptions and clarifications, NEI 04-02,Revision 2, as providing methods acceptable to the NRC staff for adopting a fire protectionprogram consistent with NFPA 805. Methods that have not been determined to be acceptableby the NRC staff, or acceptable methods that appear to have been applied differently thandescribed, require additional justification to allowv the NRC staff to complete its review of theproposed method.Appendix H of the LAR does not indicate that FAQ 12-0064, "Hot Work/Transient FireFrequency influence Factors," dated January 17, 2013 (ADAMS Accession No. ML 12346A488),was used in preparation of the EPRA. According to this FAQ, transient influence factor may notbe assigned a ranking value of 0, unless associated activities and/or entrance during poweroperation are precluded by design and/or operation. The licensee's analysis (Table C-2 ofCO-/GN-O01) indicates, however, that a large number of PAUs are assigned ranking values of 0for one or more of the transient influence factors. As a result, clarify whether ranking valuesassigned to transient influencing factors were developed consistent with the guidance inNUREG/CR-6850 and FAQ 12-0064, in particular Section 6.5.7.2, and if not, providejustification. if justification cannot be provided, then provide treatment of transient influencefactors consistent with NRC guidance in the integrated analysis provided in response to PRARA/ 03.CCNPP RESPONSE PRA RAI 05:Response to be provided 4/13/15.PRA RAI 06 -Reduced Transient Heat Release Rates: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 an FPRA and endorses, with exceptions and clarifications, NE/ 04-02,Revision 2, as providing methods acceptable to the NRC staff for adopting a fire protectionprogram consistent with NFPA 805. Methods that have not been determined to be acceptable38 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805by the NRC staff, or acceptable methods that appear to have been applied differently thandescribed require additional justification, to allow the NRC staff to complete its review of theproposed method.It appears that reductions below the NUREG/CR-6850 98th percentile HRR of 317 kilowatt (kW)for transient fires may have been credited in the FPRA. In particular, the licensee's analysis(e.g., Section 6.5.4 of Addendum 1 to CO-FSS-O04) indicates that a 142 kW (75th percentile)HRR transient fire was postulated in the switchgear rooms. As a result, discuss the key factorsused to justify any reduced HRR below 317 kW, per the guidance endorsed by the June 21,2012, memo from Joseph Glitter to Biff Bradley, "Recent Fire PRA Methods Review PanelDecisions and EPRI 1022993, 'Evaluation of Peak Heat Release Rates in Electrical CabinetFires"' (ADAMS Accession No. ML 12171A583). In doing so:a) Identify all PA~ls for which a reduction in the HRR below 317 kW for transient fires iscredited.b) For each location where a reduced HRR is credited, describe the administrative controlsthat justify the reduced HRR, including how location-specific attributes and considerationsare addressed.c) Provide the results of a review of records related to violations of transient combustible andhot work controls, including how this review informs the development of administrativecontrols credited, in part, to justify an HRR lower than 317 kW.CCNPP RESPONSE PRA RAI 06:Response to be provided 4/113/15.PRA RA! 07- Self-Igqnited and Caused by Welding and Cuttinu,: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 an FPRA and endorses, with exceptions and clarifications, NEI 04-02,Revision 2, as providing methods acceptable to the NRC staff for adopting a fire protectionprogram consistent with NFPA 805. In letter dated July 12, 2006, to NEI (ADAMS AccessionNo. ML061660105), the NRC established the ongoing FAQ process where official agencypositions regarding acceptable methods can be documented until they can be included inrevisions to RG 1.205 or NEI 04-02.Appendix H of the LAR does not indicate that FAQ 13-0005, "Cable Fires Special Cases: Self-Ignited and Caused by Welding and Cutting," dated June 26, 2013 (ADAMS Accession No.ML 13322B260), was used in preparation of the FPRA. Explain whether the treatment of self-ignited fires and fires caused by welding and cutting is consistent with FAQ 13-0005, and if not,provide justification. If justification cannot be provided, then provide treatment of self-ignitedfires and fires caused by welding and cutting consistent with NRC guidance in the integratedanalysis provided in response to PRA RAI 03.CCNPP RESPONSE PRA RAi 07:Response to be provided 3/11/15.39 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805PRA RAI 08 -Junction Boxes: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 an EPRA and endorses, with exceptions and clarifications, NE! 04-02,Revision 2, as providing methods acceptable to the NRC staff for adopting a fire protectionprogram consistent with NFPA 805. In letter dated July 12, 2006, to NEi from SunilWeerakkody (ADAMS Accession No. ML061660 105), 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.Appendix H of the LAR does not indicate that FAQ 13-0006, "Modeling Junction Box Scenariosin an Fire PRA," dated May 6, 2013 (ADAMS Accession No. ML13149A527), was used inpreparation of the FPRA. Explain whether the treatment of junction box fires is consistent withFAQ 13-0006, and if not, provide justification. If justification cannot be provided, then providetreatment of junction box fires consistent with NRC guidance in the integrated analysis providedin response to PRA RA! 03.CCNPP RESPONSE PRA RAI 08:Response to be provided 3/11/15.PRA RAI 09 -Sensitive Electronics: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 an FPRA and endorses, with exceptions and clarifications, NEI 04-02,Revision 2, as providing methods acceptable to the NRC staff for adopting a fire protectionprogram consistent with NFPA 805. Methods that have not been determined to be acceptableby the NRC staff, or acceptable methods that appear to have been applied differently thandescribed, require additional justification to allow the NRC staff to complete its review of theproposed method.The NRC staff could not identify in the LAR or licensee's analysis a description of how potentialfire damage to sensitive electronics was modeled. Though the treatment of sensitive electronicsmay be consistent with recent guidance on the modeling of sensitive electronics, Appendix H ofthe LAR does not cite FAQ 13-0004, "Clarifications Regarding Treatment of SensitiveElectronics," dated December 3, 2013 (ADAMS Accession No. ML13322A085), as one of theFAQ guidance documents used to support the FPRA. Describe the treatment of sensitiveelectronics for the FPRA and explain whether it is consistent with the guidance in FAQ 13-0004,including the caveats about configurations that can invalidate the approach (i.e., sensitiveelectronic mounted on the surface of cabinets and the presence of louvers or vents). If theapproach is not consistent with FAQ 13-0004, justify the approach, or replace the currentapproach with an acceptable approach in the integrated analysis performed in response to PRARAI 03.CCNPP RESPONSE PRA RAI 09:Response to be provided 4/13/15.40 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE______ PROTECTION ASSOCIATION STANDARD 805 ______PRA RAI 10- Conditional Probabilities of Spurious 0perations: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.RO 1.174 provides quantitative guidelines on CDF, LERF, and identifies acceptable changes tothese frequencies that result from proposed changes to the plant's licensing basis anddescribes a general framework to determine the acceptability of risk-informed changes. TheNRC staff's review of the in formation in the LAB has identified additional information that isrequired to fully characterize the risk estimates.Attachment V of the LAB indicates that application of circuit failure probabilities was limited tocircuits without control power transformers and further clarifies that the probabilities appliedyield conservative risk and delta risk estimates relative to the July 1, 2013, interim guidance(ADAMS Accession No. ML13165A214). However, new guidance on using conditionalprobabilities of spurious operation for control circuits was recently issued by the NRC inSection 7 of NUREG/CR-7150, Volume 2. This guidance included a) replacement of theconditional hot short probability tables in NUREG/CR-6850 for Option #1 with new circuit failureprobabilities for single brepk and double break control circuits, b) Option #2 in NUREG/CR-6850is not an adequate methodi and should not be used, c) replacement of the probability of spuriousoperation duration figure in FAQ 08-005 1 for AC control circuits, d) aggregate values for circuitfailure probabilities should be used unless it is demonstrated that a cable is only susceptible to asingle failure mode, e) incorporation of the uncertainty values for the circuit failure probabilitiesand spurious operation duration in the SOKC for developing the mean CDF!LERF, andf) recommendations on the hot short probabilities to use for other cable configurations, includingpanel wiring, trunk cables, and instrument cables. Provide an assessment of the assumptionsused in the Calvert Cliffs FPRA relative to the updated guidance in NUREG/CR-7 150,Volume 2, specifically addressing each of the above items. If the FPRA assumptions are notbounded by the new guidance, provide a justification for each difference, or provide updated riskresults as part of the aggregate change-in-risk analysis requested in PRA RAI 03, utilizing theguidance in NUREG/CR-7150.CCNPP RESPONSE PRA RA1 10:Response to be provided 4/13/15.PRA RAi 11 -Counting and Treatment of Bin 15 Electrical Cabinets:Section 2.4.3.3 of NEPA 805 states that the PRA approach, methods, and data shall beacceptable to the NRC. RO 1.205 identifies NUREG/CR-6850 as documenting a methodologyfor conducting an FPRA and endorses, with exceptions and clarifications, NE! 04-02,Re vision 2, as providing methods acceptable to the NRC staff for adopting a fire protectionprogram consistent with NFPA 805. Methods that have not been determined to be acceptableby the NRC staff, or acceptable methods that appear to have been applied differently thandescribed, require additional justification to allow the NRC staff to complete its review of theproposed method.The licensee's analysis (Section 2.2.1 of CO-FSS-O02) appears to indicate that the EPRAevaluates the potential for propagation of electrical cabinet fires based solely on the text inAppendix 0 (Section G.3.3) to NUREG/CR-6850; however, portions of this text were either41 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805clarified or disregarded in Chapter 8 of Supplement 1 of NUREG/CR-6850. In fight of thisobservation, address the following:a) Per Section 6.5.6 of NUREG/CR-6850, fires originating from within "well-sealed electricalcabinets that have robustly secured doors (and/or access panels) and that house onlycircuits below 440V'" do not meet the definition of potentially challenging fires and,therefore, should be excluded from the counting process for Bin 15. By counting thesecabinets as ignition sources within Bin 15, the frequencies applied to other cabinets areinappropriately reduced. Clarify that this guidance is being applied. If not, then addressthe impact as part of the integrated analysis performed in response to PRA RAI 03.b) Clarify if the criteria used to evaluate whether electrical cabinets below 440V are "wellsealed" are consistent with guidance in Chapter 8 of Supplement 1 of NUREG/CR-6850.If not, then address the impact as part of the integrated analysis performed in response toPRA RAI 03.c) All cabinets having circuits of 440V or greater should be counted for purposes of Bin 15frequency apportionment based on the guidance in Section 6.5.6 of NUREG/CR-6850.Clarify that this guidance is being applied. If not, then address the impact as part of theintegrated analysis performed in response to PRA RAi 03.d) Por those cabinets that house circuits of 440V or greater, propagation of fire outside theignition source should be evaluated based on guidance in Chapter 6 of NUREG/CR-6850,which states that "an arcing fault could compromise panel integrity (an arcing fault couldburn through the panel sides, but this should not be confused with the high energy arcingfault type fires)." Describe how fire propagation outside of cabinets greater than 440V isevaluated (including those that are considered "well-sealed'). If propagation is notevaluated, then address the impact as part of the integrated analysis performed inresponse to PRA RA! 03.CCNPP RESPONSE PRA RAI 11:11 a -Response to be provided 4/13/15.1 lb -Response to be provided 3/11/15.1 lc -Response to be provided 3/11/15.1lid -Response to be provided 3/11/15.PRA RAI 12- High,,Energyv Arcingq Faults: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 an FPRA and endorses, with exceptions and clarifications, NEI 04-02,Revision 2, as providing methods acceptable to the NRC staff for adopting a fire protectionprogram consistent with NFPA 805. Methods that have not been determined to be acceptableby the NRC staff, or acceptable methods that appear to have been applied differently thandescribed, require additional justification to allow the NRC staff to complete its review of theproposed method.The NRC staff could not identify in the LAR or licensee's analysis a description of how HEAFwere modeled. The licensee's analysis (e.g., Appendix B to CO-FO-CO1) appears to indicate42 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805that HEAF ignition sources are combined with other ignition sources (e.g., transients) to formfire scenarios. Per Appendix P of NUREG/CR-6850, however, HEAF events and other types offires have different non-suppression probability curves. In addition, the NRC staff'sinterpretation of the NUREG/CR-6850 guidance is that the growth of a fire subsequent to aHEAF event, unlike other types of fires, instantaneously starts at a non-zero HRR because ofthe intensity of the initial heat release from the HEAF:. As a result, provide a detailedjustification of the FPRA's treatment of HEAF events and the ensuing fire that includes adiscussion of conservatisms and non-conservatisms relative to the accepted methods andassesses the associated impacts on the fire total and delta risk results. Alternatively, replacethe current approach with an acceptable approach in the integrated analysis performed inresponse to PRA RAI 03. Note that the response should address the treatment of all HEAFscenarios, including in the HGL analysis and MCA.CCNPP RESPONSE PRA RAI 12:Response to be provided 4/13/15.PRA RAI 13 -MCR Modelinq: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.RG 1.174 provides quantitative guidelines on CDF, LERE, and identifies acceptable changes tothese frequencies that result from proposed changes to the plant's licensing basis anddescribes a general framework to determine the acceptability of risk-informed changes. TheNRC staff's review of the information in the LAR has identified additional information that isrequired to fully characterize the risk estimates.The licensee's analysis (Section 11.1 of CO-FSS-007) appears to assume that all of the wiringinside MCR control panels is qualified, even though unqualified wiring is known to be present aswell. Describe how the presence of both qualified and unqualified wiring is incorporated into theNUREG/CR-6850 Appendix L evaluation. Alternatively, provide treatment of qualification that isconsistent with or bounds the actual MCR configuration in the integrated analysis provided inresponse to PRA RAI 03.CCNPP RESPONSE PRA RAI 13:Response to be provided 4/13/15.PRA RAI 14 -Credit for MCR Abandonment Actions: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 NEPA 805 basedprogram, and all future plant changes to the program, shall be acceptable to the NRC.RG 1.174 provides quantitative guidelines on CDF, LERF, and identifies acceptable changesto. these frequencies that result from proposed changes to the plant's licensing basis anddescribes a general framework to determine the acceptability of risk-informed changes. TheNRC staff's review of the information in the LAR has identified additional information that isrequired to fully characterize the risk estimates.43 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE________PROTECTION ASSOCIATION STANDARD 805Tables W-2 through W-5 of the LAR and the licensee's analysis (Section 9.0 of CO-FSS-007)appear to represent MCR abandonment on loss of habitability as a single scenario with unitspecific CCDP and conditional large early release probability (CLERP) values. However, theNRC staff could not identify in the LAR or the licensee's analysis the method(s) used to obtainthese values./In light of this:a) Describe how MCR abandonment was modeled for loss of habitability in both the post-transition and the compliant plant. Include identification of the actions required to executesafe alternate shutdown and how they are modeled in the FPRA, including actions thatmust be performed before leaving the MCR. Also, include an explanation of how theCCDPs and CLERPs are estimated for fires that lead to MCR abandonment.b) Explain how the CCDPs and CLERPa estimated for fires that lead to abandonment due toloss of habitability address various possible fire-induced failures. Specifically, provide adiscussion of how the following scenarios are addressed:i. Scenarios where fire fails only a few functions aside from forcing MOB abandonmentand successful alternate shutdown is straightforward;ii. Scenarios where fire could cause some recoverable functional failures or spuriousoperations that complicate the shutdown, but successful alternate shutdown is likely;and,iii. Scenarios where the fire-induced failures cause great difficulty for shutdown by failingmultiple functions and/or complex spurious operations that make successfulshutdown unlikely.c) Explanation of the timing considerations (i.e., total time available, time until cues arereached, manipulation time, and time for decision-making) made to characterize scenariosin Part (b). Include in the explanation the basis for any assumptions made about timing.d) Discussion of how the probability associated with failure to transfer control to the AuxiliaryShutdown Panel is taken into account in Part (b).CGNPP RESPONSE PRA RAI 14:Response to be provided 4/13/15.PRA RAt 15-= MCR Abandonment on Loss of Control:Section 2.4.3.3 of NEPA 805 states that the PRA approach, methods, and data shall beacceptable to the NRC. Section 2.4.4.1 of NEPA 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.RG 1.174 provides quantitative guidelines on CDF, LERF, and identifies acceptable changes tothese frequencies that result from proposed changes to the plant's licensing basis anddescribes a general framework to determine the acceptability of risk-informed changes. TheNRC staff's review of the information in the LAR has identified additional information that isrequired to fully characterize the risk estimates.LAR Table G-1 identifies several PCS actions for non-MCR fire areas (Fire Areas 16 and 17),which encompass, in part, the Unit 1 and Unit 2 CSRs. Additionally, the licensee's analysis(Table 6 of CO-HRA -001) appears to credit actions to transfer control from the MCR to theauxiliary shutdown panel for fires in the CSR. In light of this:44 A1-TACHM ENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805a) Clarify whether the above fire areas (or other non-MCR areas) contain fire scenarios forwhich primary command and control is not retained in the MCR (i.e., the MOR isabandoned), and if so, explain how this decision was reached.b) If primary command and control is retained in the MCR, then RG 1.205 states, "Operationof dedicated or alternative shutdown controls while the MCR remains the command andcontrol location would normally be considered a recovery action." If actions taken at thePCS are not considered RAs for scenarios in which primary command and control areretained in the MCR, assess the impact of treating such actions consistent with RG 1.205on both the delta risk and additional risk of RAs as part of the integrated analysisperformed in response to PRA RAI 03. Additionally, discuss the results of the feasibilityand reliability evaluation of any new RAs in accordance with FAQ 07-0030.c) For scenarios in which primary command and control is not retained in the MCR and isinstead transferred to the PCS, the actions taken at the PCS are not RAs, and the MOR isassumed to be abandoned on loss of control (or function). Describe these scenarios,discussing how actions taken prior to and after MCR abandonment are modeled in theFPRA and its HRA. Additionally, explain the cues that result in the decision to abandonand their timing, identify the instruments being relied upon to make the abandonmentdecision, discuss whether the identified instruments are protected, and discuss how failureto transfer control to the PCS is taken into account.CCNPP RESPONSE PRA RAI 15:Response to be provided 4/113/15.PRA RAi 16 -State-of-Knowledgqe Correlation: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 NEPA 805 basedprogram, and all future plant changes to the program, shall be acceptable to the NRC.RG 1.174 provides quantitative guidelines on CDF, LERF, and identifies acceptable changes tothese frequencies that result from proposed changes to the plant's licensing basis anddescribes a general framework to determine the acceptability of risk-informed changes. TheNRC staff's review of the information in the LAR has identified additional information that isrequired to fully characterize the risk estimates.Section 4.7.3 of the L4R explains that the sources of uncertainty in the FPRA were identified,and specific parameters were analyzed, for sensitivity in support of the NFPA 805 FRE process.It is further explained that during the FRE process, the uncertainty and sensitivity associatedwith specific FPRA parameters were considerations in the evaluation of the change in riskrelative to the applicable acceptance thresholds. Based on these explanations, it appears thatthe risk results presented in Attachment W of the LAR are point estimates and do not includeparameter uncertainty. Explain how the SOKC was taken into account in the FPRAquantification, including fire ignition frequencies, circuit failure likelihood and hot short duration,and non-suppression probabilities. If the SOKC for these parameters was not addressed in theFPRA quantification, then include the impact of the SOKC for these parameters in the integratedanalysis performed in response to PRA RA! 03.45 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805CCNPP RESPONSE PRA RAI 16:The risk information presented in Attachment W is based on point estimate calculations. This isconsidered a reasonable approximation to the equivalent mean calculation. The state-of-knowledge correlation (SOKO) only causes the mean to be larger when multiple correlatedparameters appear in a single cutset. The vast majority of the cutsets contain only uncorrelatedparameters (e.g. fire scenario frequency, a single HRA failure likelihood, and a non-suppressionprobability). There are cutsets that contain multiple operator actions, but these events arealready evaluated as a joint failure probability which addresses the correlation issue. There arealso cutsets which could contain multiple circuit failure likelihood values. The CCNPP FPRAuncertainty notebook (C0-UNC-001 Revision 1) indicates only a minor change in the risk whencircuit failure likelihoods are not credited (see sensitivity analysis summary below, for Unit 1(Unit 2 similar):Unit 1 Sensitivity Impact IU-.00USensitivity AnalysisFigure 15: Impact of Sensitivity In Terms of Change to Unit 1 CDFGiven the above, the use of point estimates in the delta risk calculations is considered to beappropriate.The uncertainty analysis documentation in C0-UNC-001 will be updated to include a comparisonof the mean to the point estimate, and a sensitivity study on the circuit failure likelihood.46 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805PRA RAt 17- Sensitivity Analysis on FAQ 08-0048 Fire Bin Frequencies: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 an FPRA and endorses, with exceptions and clarifications, NE! 04-02,Revision 2, as providing methods acceptable to the NBC staff for adopting a fire protectionprogram consistent with NEPA 805. Methods that have not been determined to be acceptableby the NRC staff, or acceptable methods that appear to have been applied differently thandescribed, require additional justification to allow the NBC staff to complete its review of theproposed method.The licensee's analysis appears to indicate that generic fire ignition frequencies were basedupon those provided in Supplement 1 to NUBEG/CR-6850. Chapter 10 of this supplement,however, states that a sensitivity analysis should be performed when using the fire ignitionfrequencies in the supplement instead of those provided in Table 6-1 of NUREG/CR-6850. Aspart of the response to PRA BAI 03, provide the results (i.e., CDF, LERF, AICDF and ALERF) ofa sensitivity analysis that evaluates the impact of using the supplement frequencies, consistentwith Chapter 10 of Supplement 1 to NUREG/CR-6850. If RG 1.17 4 risk acceptance guidelinesare exceeded, (1) discuss which ones are exceeded, (2) describe the fire protection or relatedmeasures that will be taken to provide additional DID, and (3) discuss conservatisms in theanalysis and the risk significance of these conservatisms.CCNPP RESPONSE PRA RAI 17:Response to be provided 4/13/15.PRA RAI 18 -Calculation of VFDRAzCDF andAzLERF: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.RG 1.174 provides quantitative guidelfines on CDF, LERF, and identifies acceptable changes tothese frequencies that result from proposed changes to the plant's licensing basis anddescribes a general framework to determine the acceptability of risk-informed changes. TheNBC staff's review of the information in the LAB has identified additional information that isrequired to fully characterize the risk estimates.Section W.2. 1 of the LAB provides some description of how the change-in-risk and theadditional risk of RAs associated with VFDRs is determined, but not enough detail to make theapproach completely understood. As a result, provide the following:a) A detailed definition of both the post-transition and compllant plant models used tocalculate the reported change-in-risk, including any special calculations for the MCR andother abandonment areas (if applicable). Include description of the model adjustmentsmade to remove VFDRs from the Compliant plant model, such as adding events or logic,or use of surrogate events. Also, provide an explanation of how VFDB- and non-VFDR-related modifications are addressed for both the post-transition and compllant plantmodels.47 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805b) Justification for the assumption in the licensee's analysis (Section 8.0 of EPM ReportB2215-008-024) that the risk associated with the post-transition p/ant model is consideredequivalent to that of the compliant plant model for scenarios requiring MCR abandonment.c) A description of how the reported additional risk of RAs was calculated, including anyspecial calculations performed for the MCR and other abandonment areas (if applicable).If non-VFDR-rela ted modifications are credited to reduce delta risk, equating theadditional risk of RAs (as discussed in W.2. 1) to the sum of the delta risks of the VFDRsthat are resolved by crediting an RA may be non-conservative. In this case, the additionalrisk of these RAs should be re-calculated consistent with FAQ 07-0030 as part of theintegrated analysis performed in response to PRA BAI 03.d) A summary of the types of VFDRs that were identified but not modeled in the EPRA.Include any qualitative rationale for excluding these from the change-in-risk calculations.e) A clarification of whether they DID RAs listed in Attachment G of the LAB are quantified inthe EPRA. Also, explain whether credit for such DID RAs is necessary for the change-in-risk to be acceptable.CCNPP RESPONSE PRA RAI 18:18a -Variances from deterministic requirements (VFDRs) were removed from the CCNPPFPRA compliant plant model by setting the VFDR related cables or basic events to false (norandom failure for basic events set to false). Basib events are set to false when equipment inthe room would be damaged by fire for those components whose loss causes the VFDR.The delta risk was obtained by quantifying this compliant case and comparing with the basemodel risk. All modifications, VFDR and non-VFDR related, were including in both the post-transition and compliant plant models.There were no special considerations for the MCR regarding the VFDR calculations.Further, in the submittal, fires in the MCR are the only fires that forced a complete MCRabandonment. The VEDR delta risk approach as described above was used in all areasMCR and non-MCR, abandonment and non-abandonment.1 8b -Response to be provided 4/13/15.18c -The compliant plant will be evaluated by setting the root cause failures for each VFDR(cables or basic events) to FALSE (i.e., by simulating a deterministically compliant version ofthe fire area). Risk will then be calculated and the difference between the base model andthe compliant case will be the delta risk. The delta risk of recovery action will be obtained bysubtracting the baseline risk from the calculated risk with the HRA successful (i.e., thedifference between the base model risk with the HRA set to zero, or equivalent compliantcontrol room HEP value, and the base model risk). This delta risk can be represented as:Delta Risk of Recovery Action = Rrecov base -Rrecov complWhere:Rrecov base = the baseline fire risk of associated scenario with credit for a recoveryModifications not associated with VFDRs are not credited to reduce delta risk.48 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805The results of the compliant plant evaluation described above will be will be submitted inconjunction with the response to RAI PRA-03.18d -Variances from deterministic requirements were identified for the loss of required powersupplies; loss of support systems, including HVAC; loss of pumps for credited flow paths;failures of valves for credited flow paths; and, spurious operations in both credited and non-credited paths which could impact the NSPC.VFDRs were excluded from the change in risk calculations if the component or componentscovered by the VFDR were screened out during the CCNPP FPRA's component selectionprocess. Examples include VFDRs covering non-FPRA credited instrumentation (e.g.,Reactor Coolant System temperature indication) and pressurizer heaters.18e -The instances in which recovery actions are specified as Risk!/ Defense In Depth (DID)(i.e., credited for both risk and DID), the actions will either be retained as Risk actions onlyand will be included in the CCNPP FPRA risk quantification or they will be re-classified asDID only and will not be credited in the FPRA risk quantification. This will ensure that all DIDactions are credited as additional actions for which credit is not included in the CCNPP FPRAquantification.PRA RA! 19 -Attachment W Inconsistencies:Several inconsistencies were noted within Attachment W as well as between its tables andthose in Attachments C and G for particular fire areas. In light of this:a) Provide clarification on the following inconsistencies, and discuss their significance to therisk results reported in Tables W-6 and W-7:i.In Table W-6, Unit 1 Fire Areas 2, 8, 13, 18, 18A, 22, 23, 25, 26, 27, 28, 31, 38, 40,and 2CNMT are indicated as Deterministically Compliant (4.2.3.2); however, they areindicated as having VFDRs (i.e., there is a "Yes"' under the "VFDR" column andsometimes under the "RAs" column) as well as very small risk values (i.e., Fire Area18) or epsilon for ACDF/ALERF. Similarly, in Table W-7, Unit 2 Fire Areas 3, 4, 6, 14,15, 19, 19A, 21, 30, 33, 39, and 1CNMT are noted as Deterministically Compliant(4.2.3.2); however, they are indicated as having VFDRs and very small risk values(i.e., Fire Areas 19 and 30) or epsilon for ACDF/A4LERF. Attachment C does notidentify any of the above deterministic fire areas as having VFDRs. Furthermore,while for most of these fire areas the AICDF/ALERF and additional risk of RAs isreported to be epsilon, actual (very small) numerical values are reported forA4CDF/ALERF for Unit 1 Fire Area 18 and for Unit 2 Fire Areas 19 and 30, and actual(very small) numerical values are reported for additional risk of RAs for Unit 1 FireArea 23.ii. In Table W-6, Unit 1 Fire Areas 12, 14, 15, 19A, 21, 30, 32, 33, 35, 36, 39, 1CNMT,and IS are indicated as Performance-Based (4.2.4.2) and as having an RA credited inthe FPRA (i.e.. there is a "Yes"' under the "RAs" column); however, no RAs aredescribed in the VFDR dispositions presented in Attachment C or listed in AttachmentG for these areas. Similarly, Unit 2 Fire Areas 12, 13, 18A, 20, 26, 27, 28, 32, 34, 35,36, 40, 2CNMT, and IS are indicated as Performance-Based (4.2.4.2) and identify a"Yes"' under RA; however, no RAs were described in the VFDR dispositionspresented in Attachment C or listed in Attachment G for these areas. Furthermore,49 ATT-ACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805while for most of these fire areas the additional risk of FAs is reported to be epsilon,actual (very small) numerical values are reported for Unit 1 Fire Areas 21 and 36 andfor Unit 2 Fire Area 13.iii. The LERF (9.49E-O8/year (yr) reported in Table W-4 for scenario PAU CC-1A-C(Complete Burn of Vertical Cable Chase 1A) is greater than the total LERF(4.04E-OS/yr) reported in Table W-6 for Fire Area 20 (Cable Chase 1A). For twoscenarios reported in Table W-4 (PAU 230E-C and PAU 230W-C,), which representfires in Unit 1 Containment, the summation of their LERF (1.99E-O7/yr) is greater thanthe total LERE (1.91 E-O7/yr) reported in Table W-6 for Fire Area 1CNMT (Unit 1Containment). These inconsistencies also exist between Tables W-5 and W-7 for thesame scenarios in Unit 2.iv. The Table W- 1 Unit 1 fire LERE of 3.2E-O6/(chemical reactor (rx)-yr) does not matchthe corresponding value reported in Table W-6. Similarly, the Table W-1 Unit 2 fireLERF of 4.4E-O6/(rx-yr) does not match the corresponding value reported in TableW-7.b) Describe what is meant by the use of "e," or epsilon, in columns for Fire Area CDF/LERF,ACDF/AlLERF, and additional risk of RAs. Address if epsilon is defined by a specific cut-off value(s). Also, clarify how an actual value for LERE can be reported while epsilon isreported for the corresponding CDF (i.e., Unit 1 Fire Area 24 for additional risk of RAs,*Unit 2 Fire Areas 8 and 10 for CDF/LERFandA CDF/LERF).c) Describe what is meant by the use of "N/A" in columns for Fire Area CDF/LERF,ACDF/A1LERF, and additional risk of RAs. In doing so, clarify the basis for not reportingFire Area CDF/LERF values (or epsilon) for Unit 1 and Unit 2 Fire Areas 44, A B-i, AB-3,ABFL, DGB1, DGB2, and TBFL.d) Tables W-6 and W-7 include a risk reduction credit for internal events that is described ina footnote to these tables as covering random failures and internal floods. This riskreduction credit is used to offset the increase in fire risk reported in these tables. Explainhow the risk reduction from internal events reported in these tables is calculated.CCNPP RESPONSE PRA RAI 19:1 9a -Response to be provided 4/13/15.1 9b -Epsilon is not defined by a specific cutoff. It is used to indicate that the risk contributionfrom that element is negligible. A review of the usage of epsilon will be done in conjunctionwith the final quantification and submittal of RAI 3 Attachment W tables to eliminateinconsistencies in the use of this term in the LAR tables.1 9c -Calvert Cliffs Nuclear Power Plant LAR Attachment W, Tables W-6 and W-7 utilized "N/A"in the columns for "Fire Area CDF/LERF, ACDF/ALERF" and "Additional Risk of RAs" for FireAreas where there were no VFDRs and/or no risk assessments. Attachment W will berevised to provide an explanation of this use of "N/A" in the final Tables W-6 and W-7 whichwill be submitted in conjunction with the response to RAI PRA-03.1 9d -Response to be provided 4/13/1550 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FlIREPROTECTION ASSOCIATION STANDARD 805-PRA RAI120- Implementation Item Impact on Risk Estimates:Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall beacceptable to the NRC. Section 2.4.4.1 of NFPA 805 further states that the change in publichealth risk arising from transition from the current fire protection program to an NFPA 805 basedprogram, and all future plant changes to the program, shall be acceptable to the NRC.AG 1.174 provides quantitative guidelines on CDF, LERF, and identifies acceptable changes tothese frequencies that result from proposed changes to the plant's licensing basis anddescribes a general/framework to determine the acceptability of risk-informed changes. TheNRC staff's review of the information in the LAR has identified additional information that isrequired to fully characterize the risk estimates.Table S-3, Implementation Item 12 of the EAR commits to updating the EPRA and verifying therisk results after "risk related" plant modifications have been incorporated. However, it isunclear to which modifications the implementation item refers. Update Implementation Item 12to reflect completion of both the Table S-2 modifications and Table S-3 implementation itemsbefore this verification.CCNPP RESPONSE PRA RAI 20:Any as-built changes that affect the CCNPP FPRA model will include an evaluation of whetherthe acceptance criteria and the delta risk criteria are still satisfied.We propose revising Implementation Item 12 to read as follows:Incorporate as built risk related modifications, implementation items and any other additionalrefinements that may be needed into the Fire PRA and Internal Events Model and verify the riskresults are not appreciably changed. The credited modifications will be tracked through thedesign input and the engineering configuration control process. The PRA model will be updatedas necessary to reflect the final change package, and impacts to the risk estimates will beverified. As the actual engineering implementation of each modification is developed in concertwith Fire PRA evaluations of the proposed change, if the Fire PRA indicates that the as-builtchange in risk would not meet the acceptance criteria as described in LAR Section 4.5.2.2, themodification under development or its representation in the Fire PRA will be refined to ensurethat the acceptance criteria are satisfied. In addition, CCNPP will verify the validity of thereported change-in-risk estimates of Attachment W following completion of both PRA-creditedmodifications and implementation items. If this verification determines that the risk metrics havechanged such that the RG 1.205 acceptance guidelines are not met, the NRC will be notifiedand additional analytical efforts, and/or procedure changes, and/or plant modifications will bemade to assure the RG 1.205 risk acceptance criteria are met.This implementation item is an on-going action initiated within the 180 day timeframe forcompletion of implementation items but only complete after completion of modificationimplementation per Table S-2.We also propose revising the last paragraph of LAR Section 4.8.2 to read as follows:The FPRA model represents the as-built, as-operated and maintained plant as it will beconfigured at the completion of the transition to NFPA 805. The FPRA model includes credit forthe planned implementation of modifications identified in Attachment S, Table S-2. Following51 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE____ PROTECTIONASSOCIATION STANDARD 805installation of modifications and the as-built installation details, additional refinementssurrounding the modifications may need to be incorporated into the FPRA model (the FPRA willverify the validity of the reported change-in-risk on as-built conditions after the modifications arecompleted). However, these changes are not expected to be significant. No other significantplant changes are outstanding with respect to their inclusion in the FPRA model (seeAttachment S, Table S-3, IMP-12). CCNPP will verify the validity of the reported change-in-riskestimates of Attachment W following completion of both PRA-credited modifications andimplementation items. If this verification determines that the risk metrics have changed suchthat the RG 1.205 acceptance guidelines are not met, the NRC will be notified and additionalanalytical efforts, and/or procedure changes, and/or plant modifications will be made to assurethe RG 1.205 risk acceptance criteria are met.We propose revising the first bullet of LAR Section 5.5 to read as follows:Implementation of new NFPA 805 fire protection program to include procedure changes,process updates, and training to affected plant personnel. This will occur 180O daysfollowing the issuance of an approved SER from the NRC unless that date falls within ascheduled refueling outage. Then, implementation will occur 60 days after startup fromthat scheduled refueling outage. See Attachment S, Table S-3. It should be noted thatimplementation item IMP-12 is associated with incorporation of the NPFA 805modifications and the completion of this implementation item is an on-going actioninitiated within the 180 day timeframe for completion of implementation items but onlycomplete after completion of modification implementation per Table S-2.these procedurechanges, process updates, and training represent implementation items that are on-goingactions initiated within the 180 day timeframe for completion, but will only be completeafter completion of modification implementation per Table S-2.We propose revising Item (3) on Page M-5 of LAR Attachment M to read as follows:(3) The licensee shall implement the items listed in Enclosure 1, Attachment S, Table S-3,"Implementation Items," from license amendment request dated ____within 1J80 days afterNRC approval unless that implementation date falls within a scheduled refueling outage. Then,implementation will occur 60 days after startup from that scheduled refueling outage. It shouldbe noted that implementation item IMP-i12 is associated with incorporation of the NPFA 805modifications and the completion of this implementation item is an on-going action initiatedwithin the 180 day timeframe for completion of implementation items but only complete aftercompletion of modification implementation per Table S-2.1tshould be noted that theseimplementation items are on-going actions initiated within the 180 day timeframe for completionof implementation items but only complete after completion of modification implementation perEnclosure 1, Attachment 5, Table S-3.Markups of the submitted LAR pages for each of the above proposed revisions are enclosed.PRA RAI 21 -Internal Events Peer Review:Section 2.4.3.3 of NEPA 805 states that the PRA approach, methods, and data shall beacceptable to the NRC. The RG 1.205 identifies NUREG/CR-6850 as documenting amethodology for conducting an FPRA and endorses, with exceptions and clarifications,NE! 04-02, Revision 2, as providing methods acceptable to the NRC staff for adopting a fireprotection program consistent with NFPA 805. The RG 1.200 describes a peer review process52 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805utilizing an associated ASME/ANS standard (currently ASME/ANS-RA-Sa-2009) as oneacceptable approach for determining the technical adequacy of the PRA once acceptableconsensus approaches or models have been established.Attachment U of the LAR indicates that the full-scope IEPRA peer review was performed againstASME/ANS PRA Standard, RA-S-2OO8a. In light of this observation, if RG 1.200, Revision 2,and ASME/ANS PRA Standard, RA-Sa-2009, were not used as the basis for the peer review ofthe IEPRA, then discuss whether any differences between SRs were evaluated and whetherthey had any impact on the application.CCNPP RESPONSE PRA RA! 21:The peer review of the CCNPP internal events PRA was completed against RegulatoryGuide 1.200 Revision 2 and American Society of Mechanical Engineers (ASME) / AmericanNuclear Society (ANS) PRA Standard RA-Sa-2009.PRA RA1 22- PRA Up~qrades:Section 2.4.3.3 of NEPA 805 states that the PRA approach, methods, and data shall beacceptable to the NRC. The RG 1.205 identifies NUREG/CR-6850 as documenting amethodology for conducting an FPRA and endorses, with exceptions and clarifications,NEI 04-02, Revision 2, as providing methods acceptable to the NRC staff for adopting a fireprotection program consistent with NEPA 805. The RG 1.200 describes a peer review processutilizing an associated ASME/ANS standard (currently ASME/ANS-RA-Sa-2009) as oneacceptable approach for determining the technical adequacy of the PRA, once acceptableconsensus approaches or models have been established.The LAR does not indicate whether any changes made to the IEPRA or FPRA since their mostrecent full-scope peer reviews are consistent with the definition of a "PRA upgrade" inASME/ANS-RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency for NuclearPower Plant Applications," as endorsed by RG 1.200, Revision 2. In light of this, identify anysuch changes. If a focused-scope peer review has not been performed for the identifiedchanges, describe what actions will be implemented to address this issue. If a focused-scopepeer review has been performed, confirm whether it was done consistent with the guidance inASME/ANS-RA-Sa-2 009, as endorsed by RG 1.200, and provide any findings and theirresolutions.CCNPP RESPONSE PRA RAl 22:No "PRA upgrades" (as defined in ASME/ANS-RA-Sa-2009, Regulatory Guide 1.200 Revision 2and closed out Frequently Asked Questions) have been performed, or are planned to beperformed, to the CCNPP Internal Events PRA model or the CCNPP FPRA model. As such, afocused scope peer review is not required. If in the course of developing the RAI responses achange to the FPRA is identified which constitutes a PRA upgrade, a focused scope peerreview will be initiated in accordance with PRA analysis upgrade procedure requirements inExelon Training and Reference Material (T&RM) guidance documents ER-AA-600-1015 (FPIEPRA Model Update) and ER-AA-600-1 061 (Fire PRA Model Update and Control).PRA RA1 23- Deviations from Acceptable Methods: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 public53 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805health risk arising from transition from the current fire protection program to an NFPA 805 basedprogram, and all future plant changes to the program, shall be acceptable to the NRC. TheRG 1.174 provides quantitative guidelines on CDF, LERF, and identifies acceptable changes tothese frequencies that result from proposed changes to the plant's licensing basis anddescribes a general framework to determine the acceptability of risk-informed changes. TheNRC staff's review of the in formation in the LAR has identified additional information that isrequired to fully characterize the risk estimates.Section 4.5.1.2 of the LAR states that the FPRA model uses "a methodology consistent with theguidance provided in NUREG/CR-6850 and subsequent clarifications documented in responsesto NFPA 805 FAQs" and that "[n]o unre viewed methods or deviations from NUREG/CR-6850were utilized in the EPRA model development." Indicate if any other methods were employedthat deviate from other NRC-accepted guidance (e.g., subsequent clarifications documented inFAQs, interim guidance documents, etc.). If so, describe and justify any proposed method thatdeviates from NRC guidance, or replace the proposed method with an accepted method. Also,include the proposed method as a method "currently under review" as part of the integratedanalysis in the response to PRA RAI 03.CCNPP RESPONSE PRA RA! 23:Only mutually agreed upon methods will be used in preparation of the final CCNNP FPRAsupporting the NEPA 805 LAR and the responses to the associated RAIs. The peer review ofthe CCNPP FPRA that was completed against Regulatory Guide 1.200 Revision 2 andAmerican Society of Mechanical Engineers (ASME) I American Nuclear Society (ANS) PRAStandard RA-Sa-2009 did not find any unapproved methods. It is our intent to continue to useonly mutually agreed upon methods for any future work done in support of the CCNPP FPRA.and the LAR. Agreement on acceptable approaches based on the interpretation of acceptablemethods will be resolved through the RAI process.PRA RAE 24 -Defense-in-Dep~th and Safety Margqin: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 NEPA 805 basedprogram, and all future plant changes to the program, shall be acceptable to the NRC. TheRG 1.174 provides quantitative guidelines on CDF, LERF, and identifies acceptable changes tothese frequencies that result from proposed changes to the plant's licensing basis anddescribes a general framework to determine the acceptability of risk-in formed changes. TheNRC staff's review of the information in the LAR has identified additional information that isrequired to fully characterize the risk estimates.LAR Section 4.5.2.2 provides a high-level description of how the impact of transition to NFPA805 impacts DID and safety margin was reviewed, including using the criteria from Section 5.3.5of NE/ 04-02 and from RO 1.205. However, no explanation is provided of how specifically thecriteria in these documents were utilized and/or applied in these assessments.a) Provide further explanation of the method(s) or criteria used to determine when asubstantial imbalance between DID echelons existed in the FREs, and identify the typesof plant improvements made in response to this assessment.b) Provide further discussion of the approach in applying the NEt 04-02, "Guidance forImplementing a Risk-Informed, Performance-Based Fire Protection Program Under 1054 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805CFR 50.48(c)," Revision 2 (ADAMS Accession No. ML081 130188) criteria for assessingsafety margin in the FREs.CCNPP RESPONSE PRA RAI 24:24aThe method used in the CON PP .FPRA Fire Risk Evaluations (FREs) to determine when asubstantial imbalance between DID echelons existed was based on the guidance in NEI 04-02,Revision 2. Specifically, a review was performed of each DID echelon on a fire area basisbased on the following considerations:Echelon 1 (Prevent fires from starting):Combustible and hot work controls are fundamental elements of DID and as such are always inplace. The issue considered during FRE was whether this element needs to be strengthened tooffset a weakness in another echelon, thereby providing a reasonable balance.Considerations included:* Creating a new Transient Combustible Free Area;* Creating a new Hot Work Restriction Area; and/or,* Modifying an existing Transient Combustible Free Area or Hot Work Restriction Area.The fire scenarios involved in the FRE quantitative calculation were reviewed to determine ifadditional controls should be added.The remaining elements of DID were reviewed to ensure an overreliance was not placed onprogrammatic activities for weaknesses in plant design.Echelon 2 (Rapidly detect, control, and extinguish promptly those fires that do occur therebylimiting fire damage):Automatic suppression and/or detection may or may not exist in the fire area in question. Theissue considered during the FRE was whether installed suppression and/or detection wasrequired for DID or whether suppression and/or detection needed to be strengthened to offset aweakness in another echelon, thereby providing a reasonable balance.Considerations included:Risk Insights:* If existing VFDRs were never affected in a potentially risk significant fire scenario, manualsuppression capability was generally considered adequate and no additional systemsrequired.Recovery Actions:* If the fire area required recovery actions, typically detection and manual suppressioncapability were considered required. Additionally, requiring automatic suppression wasconsidered.55 A'1-ACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FiREPROTECTION ASSOCIATION STANDARD 805* If the fire area contained neither suppression, nor detection; and a recovery action wasrequired, the addition of automatic detection and/or suppression systems wereconsidered.Firefighting Activities:* If firefighting activities in the fire area were expected to be challenging (either due to thenature of the fire scenario or accessibility to the fire location), then the addition of bothsuppression and detection were considered, if absent.Fire Scenarios:* If fire scenarios credited fire detection and/or fire suppression systems, then these werealready considered to form an integral part of DID.Echelon 3 (Provide adequate level of fire protection for systems and structures so that a fire willnot prevent essential safety functions from being performed):If fires occur and they are not rapidly detected and promptly extinguished, then the third echelonof DID would be relied upon. The issue considered during the FRE was whether existingseparation was adequate (or over relied on) and whether additional measures (e.g.,supplemental barriers, fire rated cable, or recovery actions) were required to offset a weaknessin another echelon, thereby providing a reasonable balance.Considerations included:Risk Insights:* If existing VFDRs were not affected in a "potentially risk significant" fire scenario, internalfire area separation was considered adequate and no additional reliance on recoveryactions was considered necessary.* If existing VFDRs were affected in a risk significant fire scenario, internal fire areaseparation may not be adequate and reliance on a recovery action, supplemental barrier,or other modification was considered.*If the consequence associated with existing VFDRs were considered high(e.g., CCDP>I E-01 or by qualitative Safe Shutdown (SSD) assessment), regardless of-whether it is in a risk significant fire scenario, a recovery action, supplemental barriers, orother modification was considered.*There are known modeling differences between a FPRA and NSCA due to differentsuccess criteria, end states, etc. Although a VFDR may be associated with a function thatis not considered a significant contribution to core damage frequency, in some cases, theVFDR may have been considered important enough to the NSCA to retain a recoveryaction as credited for DID, but not required for risk.The fire scenarios involved in the FRE quantitative calculation were reviewed to determine thefires evaluated and the consequence in the area to best determine options for this element ofDID.Each fire area was evaluated for the need to incorporate DID enhancements to provideassurance that plant performance goals can be achieved and maintained. Documentation of56 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805these DID enhancements was tied directly to applicable VFDR dispositions in the respective firearea FRE.DID enhancements that were specifically implemented consisted of implementing transientcombustible free and hot work restricted areas; credit for fire detection systems; creditsuppression systems; and credit for DID recovery actions. These DID enhancements wereimplemented based on the above considerations.24bIn accordance with NEI 04-02, Revision 2, the maintenance of adequate Safety Margin wasassessed by the consideration categories of analyses used by the CCNPP FPRA FREs. Safetymargins were considered to be maintained if:* Codes and standards or their alternatives accepted for use by the NRC are met; and,* Safety analysis acceptance criteria in the licensing basis (e.g., UFSAR, supportinganalyses, etc.) are met, or provide sufficient margin to account for analysis and datauncertainty.The requirements related to safety margins for the FREs were evaluated for each specificanalysis type. These analyses can be grouped into four categories. These categories are:1. Fire Modeling;2. Plant System Performance;3. PRA Logic Model; and,4. Miscellaneous.As described in the CCNPP FREs, Safety Margins were maintained as follows:1) Fire Modelinga) For all fire areas assessed, the "bounding risk assessment" or "combined analysisapproach" was used as follows:i) Where the "bounding risk assessment" was used per NE! 04-02, Section 5.3.4.2, theanalysis conservatively assumed that target set damage occurred for postulated fireevents which resulted in whole room burn up. Fire modeling was not performed insupport of the change evaluations for such areas and results were therefore basedon whole area burn up. As such, the results are considered bounding.ii) Where the "combined analysis approach" was used per NEi 04-02, Section 5.3.4.3,fire modeling was performed in support of the transition within the CCNPP FPRAusing codes and standards developed by industry and NRC staff which have beenverified and validated in authoritative publications.b) The Risk-Informed, Performance-Based (RIPB) processes used were based upon NFPA805, 2001 edition, as endorsed by the NRC in 10 CFR 50.48(c).2) Plant System Performancea) The FRE process was performed in accordance with NEI 04-02, Revision 2. RegulatoryGuide 1.205 endorses portions of NEI 04-02, Revision 2, where it has been found toprovide methods acceptable to the NRC for implementing NFPA 805 and complying with10 CFR 50.48(c).57 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE________ PROTECTION ASSOCIATION STANDARD 805__b) Fire protection systems and features determined to be required by NFPA 805 Chapter 4were confirmed to meet the requirements of NFPA 805 Chapter 3 and their associatedreferenced codes and listings, or provided with acceptable alternatives using processesaccepted for use by the NRC.3) FPRA Logic Modela) The CCNPP FPRA was developed in accordance with NUREG/CR-6850, which was developedjointly between the NRC and EPRI.b) The CCNPP FPRA has undergone an industry peer review, in order to ensure that itmeets the appropriate quality standards of ASME / ANS Joint Standard RA Sa 2009,"Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment forNuclear Power Plant Applications," dated February 2, 2009.c) In accordance with the requirements of 10 CFR 50.48(c)(iii), the CCNPP FPRA results,including the sequences for the scenarios of concern, have been reviewed and it wasverified that the results do not rely solely on feed and bleed as the fire-protected safeshutdown path for maintaining reactor coolant inventory, pressure control, and decayheat removal capability for the fire areas considered.4) Miscellaneousa) No analyses that were not addressed by the above categories were identified.Example of a Typical Safety Margin Review as Contained in a FRE for a Fire Area With One orMore VFDRsIn accordance with NEI 04-02, the maintenance of adequate safety margin is assessed by theconsideration categories of analyses utilized by this FRE.Safety margins are considered to be maintained if:° Codes and Standards or their alternatives accepted for use by the NRC are met.AND°Safety analyses acceptance criteria in the licensing basis (e.g., FSAR, supportinganalyses) are met, or provides sufficient margin to account for analysis and datauncertaintyThe following summarizes the bases for ensuring the maintenance of safety margins:* The risk-informed, performance based processes utilized are based on NFPA 805, 2001edition, endorsed by the NRC in 10 CFR 50.48(c).* The FRE process is in accordance with NEI 04-02, Revision 2, which is endorsed by theNRC in RG 1.205, Revision 1.* The FPRA is developed in accordance with NUREG/CR-6850, which was developedjointly between the NRC and EPRI.°The FPRA has undergone an industry peer review, in order to ensure the FPRA meets theappropriate quality standards of ASME/ANS Joint Standard RA-Sa-2009.o The "combined analysis approach" is used during transition (NEi 04-02, Section 5.3.4.3);therefore, MEFS/LFS is not analyzed separately from the FPRA results, or58 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTIONASSOCIATIONSTANDARD 805 ______* The "bounding risk evaluation" is used during transition (NEI 04-02, Section 5.3.4.3);therefore, MEFS/LFS is not analyzed separately from the FPRA results.* The CCNPP internal events PRA model received a formal industry peer review conductedin accordance with applicable NEI guidelines. The full-scope peer review was performedin 2010. The Pressurized Water Reactor Owners Group (PWROG) peer review coveredall aspects of the CCNPP PRA model and administrative processes used to maintain themodel. All findings which could impact the FPRA results have been dispositioned, and theCCNPP PRA model has been revised as appropriate. Some findings not relevant toFPRA, such as those related to internal flooding, are pending. Refer to LAR Attachment U-Internal Events PRA Quality -for the final disposition of internal events F&Os as relatedto FPRA.* Fire protection systems and features determined to be required by NFPA 805 Chapter 4have been confirmed to meet the requirements of NFPA 805 Chapter 3 and theirassociated referenced codes and listings, or provided with acceptable alternatives usingprocesses accepted for use by the NRC (i.e., FAQ 06-0008, FAQ 06-0004, 07-0033).* Fire modeling performed in support of the transition has been performed within the FPRAutilizing codes and standards developed by industry and NRC staff which have beenverified and validated in authoritative publications, such as NUREG-1 824, "Verificationand Validation of Selected Fire Models for Nuclear Power Plant Applications." In general,the fire modeling performed in support of the FREs has been performed usingconservative methods and input parameters that are based upon NUREG/CR-6850 asdocumented in the detailed fire modeling notebook. While this is generally not ideal in thecontext of best estimate probabilistic risk analysis, it is a pragmatic approach given thecurrent state of knowledge regarding the uncertainties related to the application of the firemodeling tools and associated input parameters for specific plant configurations.* In accordance with the requirements of 10 CFR 50.48(c)(iii), the FPRA results, includingcutsets for the scenarios of concern, have been reviewed and it was verified that theresults do not rely solely on feed and bleed as the fire-protected safe shutdown path formaintaining reactor coolant inventory, pressure control, and decay heat removal capabilityfor the fire areas.59 ENCLOSURE1IUPDATED PAGESCalvert Cliffs Nuclear Power PlantFebruary 9, 2015 Constellation Energy Nuclear Group Attachment A -NEI 04-02 Table B-1 -Transition of Fundamental FP Program and Design ElementsNFPA 805 Ch. 3 Ref. RequirementelGuldance Compliance Statement Compliance Basis Reference Document3.4.1 (c)During every shift, the brigadeleader and at least two brigademembers shall have sufficienttraining and knowledge ofnuclear safety systems tounderstand the effects of fireand fire suppressants on nuclearsafety performance criteria.Exception: Sufficient training andknowledge shall be permitted tobe provided by an operationsadvisor dedicated to industrialfire brigade support.The industrial fire brigade shallbe notified immediately uponverification of a fire.Complies withClarificationAdo.3l ......... t. CCNPP complies byutilizing the exception to this section of NFPA805. CCNPP adminstrative procedures andthe UFSAR ensure that an OperationsTechnical Advisor (OTA), a licensed operatorposition, is dedicated to industrial fire brigadesupport.4A4Procedure NO-1 -200,Control of Shift Activities. Rev.055001/ Section 5.1 .AUpdated Final Safety AnaysisReport (UFSAR). Rev. 471ISection 9.9.5FPERAI023.4 1(d)CompliesNo Additional ClarificationProcedure ERPIP-3.0, ImmediateActions, Rev. 05101 /Attachment16, Section 2Procedure SA-1-101, Fire Fighting,Rev. 00600/1 Section 5.4.A and5.4.B3.4.1(e)Each industrial fire brigademember shall pass an annualphysical examination todetermine that he or she canperform the strenuous activityrequired during manualfirefighting operations. Thephysical examination shalldetermine the ability of eachmember to use respiratoryprotection equipment.CompliesNo Additional ClarificationProcedure CNG-MD-1 .01-3000,Physical Examination Process forEmployees and Contractors, Rev.00300 / Section 5.6.A.4Procedure SA-1-105, Fire BrigadeTraining, Rev. 00500 /Sections5.5.A and 5.5.A.4CCNPP Page A-42CCNPPPage A-42 Constellation Ene~rgy Nuclear Group Attachment A -NEI 04-02 Table B-I -Transition, of Fundamental FP Program and Design ElementsNFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document3.10.1If an automatic total floading andlocal application gaseous firesuppression system is requiredto meet the performance ordeterministic requirements ofChapter 4, then the system shallbe designed and installed inaccordance with the followingapplicable NFPA codes:NFPA 12, Standard on CarbonDioxide Extinguishing SystemsN/AN/A -General Statement; No TechnicalRequirementsNIAN/A3. 10. 1(1)N/ACarbon dioxide systems are not used at CCNPP.3.,10.,1(2)NFPA 12A, Standard on Haion Complies, with Required1301 Fire Extinguishing Systems ActionThe halon systems are in compliance with NFPA12A-1971 Edition as identified in report 51-9124415-~000. The halon systems are set for automaticoperation, and manual operation is accomplished bythe manual release station(s) located at the roomexit. Manual release stations are connected to thePyrotronics System 3 that is provided with batteryback-up in case of power outage per Document ES-2012-000156-001.Letter ES-2012-0001 56-001Evaluate Halon System ManualAction and Determine If CurrentSet Up and NRC CorrespondenceDeem It Acceptable, datedJanuary 31, 2013 /AllNFPA 12A, H-alon 1301Extinguishing Systems, 1971Edition / AllFAQ 08-0054, DemonstratingCompliance with Chapter 4 ofNFPA 805, Rev. 1 / AllReport 51-9124415-000, CCNPPCode Compliance Reviews, Rev.000 / Appendix C and Section 4.3See Attachment S, Table S-2, Item 44 17.IFPERA1 03ICIINII IIIIIPIIIgIeII II IIIICCNPPPage A-94 Constellation Energy Nuclear Group Attachment A -NEI 04-02 Table B-I -Transition of Fundamental FP Program and Design ElementsNFPA 805 Ch. 3 Ref. RequirementslGuldance Compliance Stlatment Compliance Basis Reference Document3.10.3Ventilation system design shalltake into account preventionfrom over-pressurization duringagent injection, adequatesealing to prevent loss of agent,and confinement of radioactivecontaminants.Complies, with RequiredActionDrawings 12263-0003, 12276-0006 and60714SH0004 show door control wiring/circuitsconnected to the halon system control panel.Procedures STP-M-699-1, STP-M-699-2, STP-M-698-1, STP-M-698-2, FTE-72, and FTE-73 outlinethe testing protocol for control functions related to ofthe integrity of the room enclosures. Confinement ofradioactive contaminants is not a concern as thereare no halon systems installed in radiological areas.Procedures STP-M-699-1, STP-M-699-2, STP-M-698-1 and STP-M-698-2 all state the followingpurpose in Section 1:"A. Verify the Halon systems for the Unit 2 CableSpreading Room (CSR) would actuate manually andautomatically upon actuation signal.B. Verify that upon a simulated actuation, theassociated ventilation dampers actuate.C. Verify that associated fire door monitor switchesare operable."Procedures STP-M-699-1, STP-M-699-2, STP-M-698-1, STP-M-698-2, FTE-72 and FTE-73 all statethe following purpose in Section 6 provides theprocedure for performance testing of the halonsystem(s) including verdfying door monitors.Item 48 of Appendix C of Report 9124415-000indicates: "The results of discharge testingsummarized in the referenced letters confirm that theenclosure strength is adequate such that pressurebuild-up is not a concern."See Attachment S, Table S-2, Item 418 17.Drawing 12263-0003, Halon 1301Fire Supression System, Rev. 11 /AllDrawing 12276-0006, Halon 1301Fire Suppression System, Rev. 6 IAllDrawing 60714SH0004, Halon1301 Fire Suppression System,Rev. 11 /AllProcedure FTE-72, Unit 1 DASRoom Halon System FunctionalTest, Rev. 00500 / Section 6.0Procedure FTE-73, Unit 2 DASRoom Halon System FunctionalTest, Rev. 00500 / Section 6.0Procedure STP M-698-1,Functional Test of Halon Systemfor the Unit 1 Cable SpreadingRoom, Rev. 0101 / Section 1.0 and6.0Procedure STP M-698-2,Functional Test of Halon Systemfor the Unit 2 Cable SpreadingRoom, Rev. 0201 / Section 1.0 and6.0Procedure STP M-699-t,Functional Test of Halon Systemfor the Unit 1 Switchgear Rooms,Rev. 00600 / Section 1.0 and 6.0Procedure STP M-699-2,Functional Test of Halon Systemfor the Unit 2 Switchgear Roams,Rev. 0501 /Section 1.0 and 6.0Report 51-9124415-000, CCNPPCode Compliance Reviews, Rev.000 / Appendix C, Item 48FPERAI 03CCNPP Page A-98CCNPPPage A-96 Constellation Energy Nuclear Group Attachment A -NEI 04.02 Table B-I -Transition of Fundamental FP Program and Design ElementsNFPA 805 Ch. 3 Ref. RequirementslGuidance Compliance Statement Compliance Basis Reference DocumentComplies, with Required Implementation items are identified below. NoneActionIMPLEMENTATION ITEMS ISee Attachment S. Table S-3IMP-2 Performance-based surveillance frequencies may be updated based on the guidance of Electric Power Research Institute (EPRI) TechnicalReport TR-1006756, "Fire Protection Surveillance Optimization and Maintenance Guide for Fire Protection Systems and Features." As aminimum, surveillance frequencies for fire dampers will be reviewed against the EPRI guidance and updated if necessary.Appreve&.....rt -'m-nte ar.c. freqm..... nPenetrations. Clarification d~~~~Potcors n fu.~ire amc pers are bou ndedb FA8 dto eto .... n .Penetrations. Clarification d~~oorsA 8 and fieFaPer ar1 onddb0FP10 Edto Sections8.2.3.2.1randrs2.NFPA 90A.CCNPP Page A-115FPERAI 04 ConstellaUon Energy Nuclear GroupAttachment L- NFPA 80S Chapter 3 Requirements for ApprovalConselltio Enegy uclar roup Attchmnt -NPA 05 haptr 3Req~reentsforAppovadetection above the ceilings in these areas. However, the NSCA-credited cables thatare routed through these above-ceiling areas are protected by metal conduit. Themetal conduits are not credited to prevent fire induced failure. ( FPE RAI 5C )This variance is technically acceptable based on the following:Based on walkdowns and above-ceiling surveys in these areas, no ignition sourceswere observed above the suspended ceilings except for extension cords which arepotentially susceptible to self-ignition. Exposed wiring above these ceilings wasobserved to be low-voltage communication and data type "network" cables which arenot prone to heat-generating overload faults. No other fixed ignition sources (i.e. fans,fan motors, etc.) were observed above the ceilings.Industry experience has shown that in the unlikely event of a self-ignited cable tray fire,the fire is not expected to spread beyond the cable tray of fire origin. The EPRI fireevents database shows that self-ignitable tray fires have only led to localized failures ina small number of cables within a single raceway. No event has led to sustained openflaming fires, or damage to cables beyond the initially impacted raceway.The extension cords above the ceilings in question are not bundled with cables or othercombustibles, nor are they routed in cable trays. There is even less likelihood that aself-ignited extension cord fire will lead to a sustained open flaming fire, due to a lack ofcombustible material in the vicinity of the extension cords. The only other significantcombustible material observed above the ceilings was ventilation duct wrap insulation.Documentation of this material identifies that the duct wrap insulation has a flamespread rating of less than 25. The duct wrap insulation will therefore not supportsustained combustion or fire growth. In the unlikely event of fire originating in theexposed non-plenum cable, fire will not spread to the duct wrap insulation.Attachment 3 of CNG-FES-007(Reference 6.38) states: uMinimize wiring abovesuspended ceilings. Where installed, electrical wiring shall be listed for plenum use,routed in armored cable, routed in metallic conduit, or routed in cable trays with solidmetal top and bottom covers." This procedure is in place to ensure that futurecompliance with this NFPA 805 requirement is achieved.Per Drawing 60739SH0001 (Reference L-2), ACA ventilation is served by one supplyunit (RTU-1) and two independent exhaust units (access control exhaust fans 11 and12).In the Fire Area 11 portion of the ACA, supply and exhaust registers in the ceiling areducted to and from these units, as shown in Drawing 60597 (Reference L-3). Theabove-ceiling space is therefore not used as an air plenum.On the Turbine Building side of the ACA, supply registers in the ceiling are ducted to thesupply unit, but some exhaust registers in the ceiling are not ducted, as shown inDrawing 60597 (Reference L-3). Exhaust air is pulled from the ceiling plenum into ductsthat lead to the Unit 2 Main Exhaust Plenum where it is exhausted by the main plantexhaust fan 21 or 22. Per Drawings 61085SH00057 (Referecne L-4); 61085SH0058(reference L-5); and 63085SH003B (Reference L-6); the ACA exhaust fans areinterlocked with the Main Plant Exhaust Fans as well as the ACA supply unit RTU-1.Per Drawings 60739SH0001 (Reference L-7) and 60722SH0001 (Reference L-8), theCCP Pag L-CCNPPPage L-9 Constellation Energy Nuclear GroupAttachment L- NFPA 805 Chapter 3 Requiroments for ApprovalApproval Request 3NFPA 805 Section 3.3.1.3.1 states:"A hot work safety procedure shall be developed, implemented, and periodicallyupdated as necessary in accordance with NFPA 51B, Standard for Fire PreventionDuring Welding, Cutting, and Other Hot Work, and NFPA 241, Standard forSafeguarding Construction, Alteration, and Demolition Operations."NFPA 51B, 1999 Edition, Section 3-2 states, in part:"Hot work shall not be allowed in the following areas:(b) In sprinklered buildings while such protection is impaired."CCNPP hot work procedures "ro doo....... in. Pr.ood.u.ro .. SA. 1 100 (Rofr-no L.13). S.A.-4.-00 allows hot work to be performed in plant areas that are protected byIautomatic sprinkler systems while such systems are impaired, contingent on firemarshal or Engineering Programs Unit approval. Administrative procedures 8A--1-4.00is are therefore not in compliance with the hot work requirements of section 3-2(b) ofI EditorialNFPA 51 B. CCNPP requests NRC approval for performance of hot work in sprinkleredbuildings while such systems are impaired as an acceptable variance from therequirements of NFPA 805 Chapter 3 requirements.Basis for Request:While expected to be a very uncommon occurrence, CCNPP anticipates that there maybe occasions where hot work is necessary in sprinklered plant areas while suchsystems are temporarily impaired. Any fire area containing a sprinkler systelm--as FPEidont.ifiod in. ^At..c...ont C, -rablo C- 2) of thie ....iio reo.... , is subject to the provisions RIOof this request. Attachment C, Table C-2 identifies all sprinkler systems in plant fire areas.The administrative procedures that are in place to limit combustibles and control hot workWith the exception of section 3-2(b) of NFPA 5IB, the procedure employed for hot workis a rigorous one and in compliance with the applicable requirements of the 1999Edition of NFPA 51B and the 2000 Edition of NFPA 241 (codes of record). A summaryof the key elements of the procedure is provided below.* A permit application for hot work is submitted to the fire marshal.* The fire marshal assigns a number to the permit, reviews the permit, and conductsand inspection of the area prior to commencing work.* A hot work permit hazard analysis checklist is successfully completed before startingwork.*The Operations group is notified prior to all hot work. This notification is requiredonce per shift.*A hot work fire watch, with the appropriate fire extinguisher for the type and size forthe hazard, is required to be present during the hot work activity and must remain inCCNPP Pge L-1CCNPPPage L-12 Constellation Energy Nuclear GroupAttachment L -NFPA 805 Chapter 3 Requlremente for ApprovalCoselaionrgy Nula ..u.Atcmn.. NP 0 hper3Rq~eetsfrApoathe immediate work area for a minimum of 30 minutes following completion of thehot work activity.* Back-up fire suppression equipment is available in areas where the fire suppressionsystem is inactive.*Combustibles located within 35 feet of the work area are removed prior to hot workoperations. For permanent combustibles that cannot be removed, they must becovre.wthth appro.iit ...... of .. an NFPA 5IB compliant blanket. FPE* Equipment is checked prior to the activity to ensure it is in good working condition.*If hot work is required in an area in which nuclear safety compensatory actions are inplace, completion of a form approved by the system manager, shiftmanager/operations maintenance coordinator, fire marshal, and responsiblemaintenance group supervisor is required per Prcdu... , A ... 102 (Ronfornco ... Editra46) administrative procedures.I*Hot work procedures are in compliance with all other applicable NFPA 51B andNFPA 241 requirements, including those related to management, permit-authorizingindividual, hot work operator training, fire watch (and training) alarm activation, hotwork areas, hot work permits, hot tapping, and fire prevention (precautions regardingcombustibles, inadvertent sprinkler discharge, etc.):T-hese-~Administrative procedures demonstrate the highest standard of care in fire Editodialprevention measures for hot work activities. The rigorous approval, documentation,training, hazard analysis, precautions, lack of combustibles, manual suppression,training, and vigilance ensure that the occurrence of a fire during hot work operations isvery unlikely. The risk of a fire growing uncontrolled beyond the incipient stage due tohot work is therefore not considered a credible scenario.Acceptance Criteria Evaluation:Nuclear Safety and RadIological Release Performance Criteria:Although the hot work requirements in administrative procedures SA---4-00 do not Editorialcomply with section 3-2(b) of NFPA 51iB, there are strict procedures in place to limit thecombustibles, control the hot work within the area, and provide a fire watch to promptlyextinguish any fires that do occur. Therefore, there is no impact on the nuclear safetyperformance criteria.The use of thev..... currant. proc.edurer, administrative procedures to perform hot work Editorialactivities has no impact on the radiological release performance criteria. Theradiological release performance criteria are satisfied based on the determination oflimiting radioactive release (Attachment E), ,.,hio..,h,..;. .,i"" no ..-c " ,.,c ^ b , S rA n 110.. EditorialSafety Margin and Defense-In-Depth:There are procedures in place to limit the combustibles and control the hot workwithin the area. Since fire prevention and manual suppression is maintained perCCNPPPage L-13 Constellation Energy Nuclear GroupAttachment L -NFPA 805 Chapter 3 Requirements for Approvaladministrative procedures ,$A-1.-4.OQ, the safety margin inherent in the analysis Editorialfor the fire event has been preserved.The three echelons of defense-in-depth are:(1) To prevent fires from starting (combustible/hot work controls)(2) Rapidly detect, control and extinguish fires that do occur, thereby limitingdamage (fire detection systems, automatic fire suppression, manual firesuppression, pre-fire plans)(3) Provide adequate level of fire protection for systems and structures so that a firewill not prevent essential safety functions from being performed (fire barriers, firerated cable, success path remains free of fire damage, recovery actions)Per NFPA 805 Section 1.2, defense-in-depth is achieved when an adequate balance ofeach of these elements is provided.Echelons 1, 2, and 3 are met through the limiting of combustibles, control of hot work,and availability of fire watch (i.e., manual suppression), through administrativeprocedures $A-4-4.1.0. The hot work procedures therefore do not compromise manual dorafire suppression functions or post-frie nuclear safety capability. Since a balance of theelements is provided, defense-in-depth is achieved.Conclusion:NRC approval is requested for the allowance of hot work activities in buildings withimpaired sprinkler systems in accordance with administrative procedures $A-4-400, Editorialcontrary to the requirements of Section 3.2(b) of NFPA 5IB, 1999 Edition. Based on theanalysis above, the level of risk encountered by maintaining this current practice isacceptable, and the approach is considered acceptable because it:(A) Satisfies the performance goals, performance objectives, and performancecriteria specified in NFPA 805 related to nuclear safety and radiologicalrelease;(B) Maintains safety margins; and(C) Maintains fire protection defense-in-depth (fire prevention, fire detection, firesuppression, mitigation, and post-frie nuclear safety capability).CCNPP Page L-14CCNPPPage L-14 Constellation Energy Nuclear GroupAttachment I -Definition of Power BlockCoselto nryNceaTru-tahetI-Deiiino oe lcThe term "power block" is clarified in Section K.2 of NEI 04-02 as "structures that haveequipment required for nuclear plant operations, such as containment, auxiliary building,service building, control building, fuel building, radiological waste, water treatment,turbine building, and intake structure, or structures that are identified in the facility'scurrent license basis." The determination of which buildings are required for nuclearplant operations (i.e., required to meet the nuclear safety or radioactive releaseperformance criteria identified in sections 1.5.1 and 1.5.2 of NFPA 805), and thusconsidered within the power block, is identified below.For the purposes of establishing the structures included in the CCNPP Fire Protectionprogram in accordance with 10 CFR 50.48(c) and NFPA 805, the buildings andstructures listed in the following table are considered to be part of the power blockbased on the application of the preceding evaluation criteria.Table I-1 -Power Block DefinitionPower Block Structures Fire Area(s)1, 2, 3, 4, 5, 6,7, 8,9, 10, 11, 12, 13, 14,15, 16, 16A, 16B, 16C, 17, 17A, 17B, 17C,18, 18A, 19, 19A, 20, 21, 22, 23, 24, 25,Auxiliary Building 26, 27, 28, 29, 30, 31, 32, 33, 34, 35, 36,37, 38, 39, 40, 41, 42, 43, 44, AB-1, AB-2,AB-3, AB-4, AB-5, ABFL, KVMAL1A Emergency Diesel Generator DB(EDG) Building ___________________0C Station Blackout (SBO) Diesel DBGenerator Building___________________Reactor Enclosure -Unit No. 1 1CNMTReactor Enclosure -Unit No. 2 2CNMTTurbine Building/ North ServiceBuilding (NSB 12' & 27' Elevations*) TB/NSB/ACA, TBFLIntake Structure IS13.8 kV Switchgear House Unit 113.8 kV Switchgear House Unit 2Condensate Storage Tank No. 12EnclosureYARDFire Protection Pump HouseNo. 2 Fuel Oil Storage Tank No. 21BuildingPretreated-Well Water House* Note: The 45' elevation of the North Service Building is excluded from Fire Area TB/NSB/ACA and the power block as justified by Engineering Equivalency EvaluationECP-1 3-000357. The bases for acceptability are summarized as follows:IFPEP Al 08FPERAI 08CCNPPPage I-2 Constellation Energy Nuclear GroupAttachment I -Definition of Power BlockCoselto nryNcea ru tahetI-Deii-no oe lc-There are no cables or equipment required to achieve nuclear safety performance FPEcriteria (NSPC) in the 45' elevation of the North Service Building. RAI 08-There are no cables or equipment required to achieve NSPC in the Yard within 50 feetof the 45' elevation of the North Service Building.-A fire originating in the 45' elevation of the North Service Building will not impact cablesor equipment required to achieve NSPC in adjacent fire area TBINSB/ACA.CCNPP ageI1-CCNPPPage 1.3 S u ly. eao IIIIIIIIIIIIIIIInfr ai -W th l UI deI 10 C II 2IIII III0Conutlahtlon Energy Nuclear GroupTable C-I -NEI 04-02 Table B-3 Fire Area TransitionWithhold under 10 CFR 2.390CC P Page C,-,,,,l 1, III II ,,,tPage C-SU Security-Related Information -Withhold Under 10 CFR 2.390Aftaahmsn C -NE! 114-2 Tabl B-3 Fire Aere TranuitlonConatelalonl Enemy Nude. GroupTable C-2 -Summary of NFPA 805 Conmpilance Basis and Required Firs Protecton Systems and FeaturesWithhold under 10 CFR 2.390CGNPP Page C.U6Page Security.Related Information -Withhold Under 10 CFR 2.390CitfstnEfuul Nuceamr Group Attachment C -NW1044G2 Table 5.3 Fir Are TrmdoTable C-2 -Suummary of NFPA 805 Compliance Basis and Required Fire Protection Systems and FeaturesWithhold under 10 CFR 2.390CCNPii i ii iii iiiiiiiiiiiiiaiiiiiiiiiiiI IIIII4nliiCCNlq3Pagan C41D7 Security-Related Information. Withhold Under 10 CFR 2.390AtahetC -NE104402 Table B-3 Fire Area TransitionConsisilslon Enugy Nuclear GroupTable C-2 -Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and FeaturesWithhold under 10 CFR 2.390CCNPP Pigs C-USCCNPPPage C.60B Security-Related Information -Withhold Under 10 CFR 2.390Atahmn C -NE1 04-02 Table B54 Fire Area Trminua~Conmtelmtli EnmYw Nueler GrouoI " I I II I II IIIII II IIIITable C-2 -Summary of NFPA 805 Compliance Basis and Required Fire Proeto Systems and FeaturesWithhold under 10 CFR 2.390CCNPP Page C.699CCNPPPage C-6H Constellation EmwW Nuclear Group Attachment E Radioactive Release TransitionNfl 04-02 Radioactive Release TransitionFind Room Pr-Fr PlnNo , n Englnwmlng Controls Training and Pro-FIrm PlanArea Room Descrpt~on and Title In Uquld Gaseous Rmilw Results Conclusonssuch as SA-1-I01 "FireFighting', SA-1 -1 05 "FireBrigade Training" (andassociated fire drills), theODCM, the Radiation SafetyManual, CNG-TR-1 .01 -1 025"Radiation Protection TrainingProgram', RP-2-100Radioactive MaterialsManagement', and RP-2-101"Radioactive Waste" { discusscontaining, monitoring, andreleasing of gaseous effluents.1 CNMT 230E Unit I SA-FFSM-A559, Yes See Room 230 See Room 230 See Room 230 See Room 230Corntainmenl Auxiliary BuildingjpR- Fire Fighting-weo. Strategy ManualICNMT 230N Unit I SA-FFSM-AB69. Yes See Room 230 See Room 230 See Room 230 See Room 230Containment Auxiliary Building..Fire Fighting..aeum- Strategy Manual1CNMT 230S Unit I SA-FFSM-AB69, Yes See Room 230 See Room 230 See Room 230 See Room 230Containment Auwdllary Building--p*- Fire Fighting..Ppuo. Strategy Manual1CNMT 230W Unit 1 SA-FFSM-AB69, Yes See Room 230 See Room 230 See Room 230 See Room 230Containment Auxiliary Building.-f Fire Fighting0.-upi.- Strategy Manual2CNMT 121 Unit 2 SA-FFSM-AB, Yes During Plant Operation: Floor The Containment Ventilation Training materials reinforce Based on the availability ofContainment Auxiliary Building drains are mouted to the System consists of several the use of FFSM's and the engineered controls for bothTendon Gallery Fire Fighting monitored Reactor Coolant subsystems to fulfill the overall Radiation Safety Manual for smoke and fire suppressionCCNIPP Pag E-371 Nudmr GroupA#admmnt E RIMime TranMtkmNE1 0402 Radioactive Release TransitionFire Room Pre.Flrm Plan No. Screeined Engineerng Controls Training and Pro.Fire PlanArs Room Deesrption and Tllo In Liquid Gaseous Rugl Results Conclusionsassociated fire drills), theODCM. the Radiation SafetyManuaml, CNG-TR-1 .01-1025"Radiation Protection TrainingProgram', RP-2-1 00"Radioactive Materialsand RP-2-101-Radioactive WasteManagement" disusscontaining, monitoring, andreleasing of gaseous effluents.2CNMT 229E Unit 2 SA-FFSM-AB69, Yes See Room 229 See Room 229 See Room 229 See Room 229Containment Auxiliary Building(p.- Fire FightingPweee Strategy Manual2CNMT 229N Unit 2 SA-FFSM-AB69, Yes See Room 229 See Room 229 See Room 229 See Room 229Containment Auxiliary Building-4pn." Fire Fighting-.pse. Strategy Manual2CNMT 229S Unit 2 SA-FFSM-AB69, Yes See Room 229 See Room 229 See Room 229 See Room 229Auxiliary Building"Fre FightingS Strategy Manual2CNMT 229W Unit 2 SA-FFSM-AB69, Yes See Room 229 See Room 229 See Room 229 See Room 229Containment Auxiliary Building-t.R.. Fire Fighting.-Pess- Strategy ManualDGBI DB001 Stair Na. 1 N/A No N/A N/A N/A N/A -Screened Out(EL-35)DGB1 DB002 Oil Separator NIA No N/A N/A N/A N/A -Screened OutRoom (EL-35)C G 11111ae111118E.M1 Security-Relatod Information -Withhold Under 10 CFR 2.390Conelellamt. Energy Nuclear Group Afalnn C -NW154-02 Table 843 Fire Area TruneonTable C-I -NEI 04.02 Table B.3 Fire Area TrnanitionWithhold under 10 CFR 2.390CCNPP Peg. V4CCNPPPage C-SS8 Security-Related Information -Withhold Under 10 CFR 2.390Attachment C -NEI 04-02 Table B-3 Fire Area TransitionConstellation Energy Nuclear GroupTable C-I -NE1 04-02 Table B-3 Fire Area TransitionWithhold under 10 CFR 2.390CCNPP Page C-245CCNPPPage C-245 Constellation Energy Nuclear GroupAttachment M -License Condition ChangesCos elltn EnryNula Grop ttchen U -cns odiinChn(3) The licensee shall implement the items listed in Enclosure 1, Attachment 8,Table S-3, "Implementation Items,TM from license amendment request dated ___within 180 days after NRC approval unless that implementation date falls within ascheduled refueling outage. Then, implementation will occur 60 days after startupfrom that scheduled refueling outage. I6.nsert(4) The licensee shall maintain appropriate e=lnsatory measures in place untilcompletion of the modifications listed above.jIt should be noted that implementation item IMP-I12 is associatedImplementation Date jwith incorporation of the NPFA 805 modifications and the RzApd 3, 218completion of'this implementation item is an on-going action RAIApi 0 08Iinitiated within the 180 day timeframe for completion of20implementation items but only complete after completion of 2modification implementation per Table S-2.CCNPP Page M-5CCNPPPage M45 Constellation Enemv Nuclear Grouo4.0 Compliance with NFPA 805 RequirementsC..... lla.i.n....... Nu ea ...... .. 4. o[ln ew t NFP 8 T R,,,,mntThe FPRA model represents the as-built, as-operated and maintained plant as it will beconfigured at the completion of the transition to NFPA 805. The FPRA model includescredit for the planned implementation of modifications identified in Attachment S, TableS-2. Following installation of modifications and the as-built installation details, additionalrefinements surrounding the modifications may need to be incorporated into the FPRAmodel (the FPRA will verity the validity of the reported change-in-risk on as-builtconditions after the modifications are completed). However, these changes are notexpected to be significant. oohr ..,... ..-.go .. r... ou,.....,n. ....r--ott tor ................. in' t ,b , FP.A. no.. Bo ts.mntB Tbo, IM......... ... ..... Insert4.8.3 Supplemental Information -Other LUcensee Specific IssuesNone. CCNPP will verify the validityof the reported change-in-riskestimates of Attachment Wfollowing completion of bothPRA-credited modificationsand implementation items. Ifthis verification determinesthat the risk metrics have P~changed such that the RG RATI1.205 acceptance guidelines 20* .1 20are not met, inc NiKt Will nenotified and additionalanalytical efforts, and/orprocedure changes, and/orplant modifications will bemade to assure the RG 1.205risk acceptance criteria aremet.CCNPP Page 59CCNPPPage 59 Constellation Energy Nuclear Group5.0 Regulatory EvaluationCoselaio TneyNula Gru ".0 R uaToyvautn51.22(c). That evaluation is discussed in Attachment R. The evaluation confirms thatthis LAR meets the criteria set forth in 10 CFR 51 .22(c)(9) for categorical exclusion fromthe need for an environmental impact assessment or statement.5.4 Revision to the UFSARAfter the approval of the LAR, in accordance with 10 CFR 50.71 (e), the CCNPP UFSARwill be revised. The format and content will be consistent with NEI 04-02 FAQ 12-0062.6.5 Transition Implementation ScheduleThe following schedule for transitioning CCNPP to the new fire protection licensingbasis requires NRC approval of the LAR in accordance with the following schedule:Implementation of new NFPA 805 fire protection program to include procedurechanges, process updates, and training to affected plant personnel. This willoccur 180 days following the issuance of an approved SER from the NRC unlessthat date falls within a scheduled refueling outage. Then, implementation willoccur 60 days after startup from that scheduled refueling outage. SeeAttachment S, Table S-3. Insert*Modifications will be complete April 30, 2018. This date assumes SERapproval within two years from uARtbmittal. Appropriate compensatorymeasures will be maintained until modifica'l¢ are complete. See Attachment8, Table S-2. ________________________[it should be noted that implementation Item IMP-12 is associated jwith incorporation of the NPFA 805 modifications and thejcompletion of this implementation item is an on-going action RATIinitiated within the 180 day timeframe for completion of 2 0implementation items but only complete after completion ofmodification implementation per Table S-2.CCNPP Page 66CCNPPPage 66 Constellation Energy Nuclear GroupAttachment U -Internal Events PRA QualityConseaionii iiiiii Enrg Nula Gru tahetU-ItenrEet ultIn accordance with RG 1.205 position 4.3:"The licensee should submit the documentation described in Section 4.2 ofRegulatory Guide 1.200 to address the baseline PRA and application-specificanalyses. For PRA Standard "supporting requirements" important to the NFPA805 risk assessments, the NRC position is that Capability CategoryI/Iis generallyacceptable. Licensees should justify use of Capability Category I for specificsupporting requirements in their NFPA 805 risk assessments, if they contend thatit is adequate for the application. Licensees should also evaluate whetherportions of the PRA need to meet Capability Category/Ill, as described in thePRA Standard."The PMVROG performed a full scope internal events PRA peer review of CCNPP todetermine compliance with ASME PRA Standard,oand RG 1.2001P(Reference 6.32) in June 2010. This review documented find Igs for all supportingI 21,requirements (SRs) which failed to meet at least Category II. Th findings for that peer 2review are documented below in Table U-i. This table also in ,ludes the disposition,status, and impact on the FPRA.The peer review found that 97% of the SR's evaluated Met Ca ability Category II orbetter. There were 3 SR's that were noted as "not met" and 8 there were noted asCategory 1. As noted in the peer review report the majority }f the findings weredocumentation related. Of the 11 SR's which did not meet Categ y 2 or better, 7 wererelated to conservatisms or documentation in LERF and 2 related to internalfloods. There were 39 findings. All findings which could be rel vant to FPRA wereupdated in the internal events model used to quantify the FPR .Thus, with theexception of minor documentation concerns, the internal events mo el meets CategoryII or causes conservative results for all SR's relevant to FPRA res Ilts. No significantchanges have been implemented in the internal events PRA. As there are no newmethods applied, no follow on or focused peer reviews were requirec1RA-Sa-2009(Reference 6.31)CCNIP PaIeIUICCNPPPage U-2 ConsteISuliofi bneIgy Nucber GroupSecurity.Related Information -Withhold Under 10 CFR 2.390Attachment S -IModiIicaions and Implameotatlon ItemsTabl S- Impleenaton tm~sWithhold under 10 CFR 2.390CCNPP Pug. S-IT Constellation Energy Nuclear GroupAttachment O -Orders and ExemptionCoselto Eng Nula rou Atcmn0-OdradEepTioExemptionsRescind the following exemptions granted against 10 CFR 50, Appendix R dated asfollows:* Auguo 16 105 AnY~ a I from thcy rgiromonte~ ofV Sootyion lll.G.2 ofthe v-e'effpti.. >, Deleted consistent with Attachment K* April 21, 1983 -An exemption from the requirements of Section lll.G.3 ofAppendix R for the control room complex and the intake structure related to theinstallation of fixed fire suppression systems.* March 15, 1984 -An exemption from the requirements of Section lII.G to allowalternatives to the 3-hour rated fire barriers for areas listed in the exemption. Anexemption was also granted for Section Ill.G for Fire Areas 10 and 11 related tothe installation of fixed fire suppression systems. Additionally, an exemption fromthe requirements of Section III.0 was granted regarding the capacity of the oilcollection systems for the reactor coolant pumps.* August 22, 1990 -An exemption from the requirements of Section lll.J to allowthe use of portable hand lights as an alternative to permanently installed 8-houremergency lights in the Unit 1 and 2 containment buildings.* April 7, 1999 -An exemption from the requirements for Section III.J to allow theuse of security lighting in exterior areas, the use of portable lights in highradiation areas and the use of helmet mounted lights inside of switchgearcabinets as alternatives to permanently installed 8-hour emergency lights.Specific details regarding these exemptions are contained in Attachment K.OrdersNo Orders need to be superseded or revised.CCNPP implemented the following process for making this determination:-A review was conducted of the CCNPP docketed correspondence. The reviewwas performed by reviewing the correspondence files and performing electronicsearches of internal CCNPP records and the NRC's ADAMS document system.A specific review was performed of the license amendment that incorporated themitigation strategies required by Section B.5.b of Commission Order EA-02-026 toensure that any changes being made to ensure compliance with 10 CFR 50.48(c) donot invalidate existing obligations applicable to the plant. The review of this orderdemonstrated that changes to the fire protection program will not affect measuresrequired by B.5.b.The Fukushima Orders are being independently evaluated. Any plant changes will beevaluated for impact on the fire protection program in accordance with the CCNPPdesign change process.CCNPP Page 0-2CCNPPPage 0-2 Security-Related Information -Withhold Under 10 CFR 2.390Attachment S -Modifications and Implementation ItemsConstellation Enemy Nuclear GIOuD.... ..i lllllllll IIII I III IIIIII II III II I III ITable S-3 Implementation ItemsWithhold under 10 CFR 2.390CCNPP Page S-15CCNPPPage 5-15}}

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+{{#Wiki_filter:~Exeton GenerationeGeorge GelirichSite Vice PresidentCalvert Cliffs Nuclear Power Plant1650 Calvert Cliffs ParkwayLusby. MD 20657410 495 5200 Office717 497 3463 Mobilewww~exeloncorp.comgeorge~gellrich@exeloncorp.com10 CFR 50.90February 9, 2015U. S. Nuclear Regulatory CommissionATTN: Document Control DeskWashington, DC 20555Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2Renewed Facility Operating License Nos. DPR-53 and DPR-69NRC Docket Nos. 50-317 and 50-318
+==Subject:==
+Request for Additional Information Regarding the National Fire ProtectionAssociation Standard 805 License Amendment Request
+==References:==
+.Letter from G. H. Gelirich (CCNPP) to Document Control Desk (NRC), datedSeptember 24, 2013, License Amendment Request re: Transition to 10 CFR50.48(c) -NFPA 805 Performance Based Standard for Fire Protection2. Letter from N. S. Morgan (NRR) to G. H. Gelirich (CCNPP), datedJanuary 12, 2015, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 -Request for Additional Information Regarding the National Fire ProtectionAssociation Standard 805 License Amendment Request (TAC Nos. MF2993and MF2994)in Reference 1, Calvert Cliffs Nuclear Power Plant, LLC submitted a license amendmentrequest to transition to 10 CFR 50.48(c) -NFPA 805 Performance Based Standard for FireProtection. In Reference 2 the NRC staff requested additional information regarding thisamendment request. Attachment (1) and the Enclosure provide the response to the request foradditional information. The schedule for providing responses to individual questions wasprovided in Reference 2. Enclosure 1 contains markups of the original license amendmentpackage pages and supersedes the previously provided pages.This additional information does not change the No Significant Hazards Determination providedin Reference 1. No regulatory commitments are contained in this letter.Should you have questions regarding this matter, please contact Mr. Douglas E. Lauver at(410) 495-5219.
+Document Control DeskFebruary 9, 2015Page 2I declare under penalty of perjury that the foregoing is true and correct.February 9, 2015.Executed onRespectfully,e6' 2/.,George l-L'GellrichSite Vice PresidentGHG/PSF/bjm
+==Attachment:==
+(1) Request for Additional Information Regarding the National Fire ProtectionAssociation Standard 805
+==Enclosure:==
+: 1. Updated pagescc: NRC Project Manager, Calvert CliffsNRC Regional Administrator, Region INRC Resident Inspector, Calvert CliffsS. Gray, MD-DNR ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THENATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805Calvert Cliffs Nuclear Power PlantFebruary 9, 2015 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805By letter dated September 24, 2013 Calvert Cliffs Nuclear Power Plant, LLC (CCNPP),submitted a license amendment request (LAR) for Calvert Cliffs Nuclear Power Plant, Units 1and 2 (Calvert Cliffs) to transition its fire protection licensing basis from Title 10 of the Code ofFederal Regulations (10 CFR) Section 50.48(b) to 10 CFR 50.48(c), National Fire ProtectionAssociation Standard (NFPA) 805, "Performance-Based Standard for Fire Protection for LightWater Reactor Electric Generating Plants," 2001 Edition. The Nuclear Regulatory Commission(NRC) staff is reviewing the application and has determined that the following additionalinformation is needed to complete the review of the LAR:Fire Protection Engineerin,. (FPE) Request for Additional In formation (RAI) 01:Section 3.3.4 of NFPA 805, 2001 Edition, requires that thermal insulation materials, radiationshielding materials, ventilation duct materials, and soundproofing materials be noncombustibleor limited combustible. in Attachment A, "NEI [Nuclear Energy Institute) 04-02 Table B-i -Transition of Fundamental Fire Protection Program & Design Elements," of the LAR, thelicensee stated that the plant "Complies with Clarification" on the basis that the referencedprocedures, specifications, and the Combustible Loading Analysis Database control andaccount for the use of thermal insulation materials, radiation shielding materials, ventilation ductmaterials, and soundproofing materials. The licensee does not state whether these materialsare specified in the documents to be noncombustible or limited combustible. Provide thefollowing information:a. Clarify that the procedure(s), specifications, and database specify that thermal insulationmaterials, radiation shielding materials, ventilation duct materials, and soundp roofingmaterials shall be noncombustible or limited combustible.b. Clarify in the compliance bases whether thermal insulation materials, radiation shieldingmaterials, ventilation duct materials, and soundp roofing materials that are eitherpermanently or temporarily installed in the plant are noncombustible or limitedcombustible.c. If installed materials are not noncombustible or limited combustible, describe how thesematerials are accounted for and managed in the fire protection program.CCNPP RESPONSE FPE RAI 01:Response to be provided by 3/11/15.FPE RAI102:Section 3.4.1(c) of NFPA 805 requires that the fire brigade leader and at least two brigademembers have sufficient training and knowledge of nuclear safety systems to understand theeffects of fire and fire suppressants on nuclear safety performance criteria (NSPC). InSection 1.6.4.1, "Qualifications," of NRC Regulatory Guide (RG) 1.189, "Fire Protection forNuclear Power Plants," Revision 2, September 2009 (ADAMS Accession No. ML 092580550),the NRC staff has acknowledged the following example for the fire brigade leader as sufficient:The brigade leader should be competent to assess the potential safety consequences of afire and advise control room personnel. Such competence by the brigade leader may beevidenced by possession of an operator's license or equivalent knowledge of plantsystems.1 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805In Attachment A, the licensee stated that it complies and references Procedure SA-1-105, FireBrigade Training, Section 4.4.A. 1, which includes the NFPA 805, Section 3.4. 1(c) requirementas a responsibility for the shift manager to assure the fire brigade members have the requisitetraining and knowledge. Provide additional detail regarding the training that is provided to thefire brigade leader and members that addresses their ability to assess the effects of fire and firesuppressants on NSPC.CCNPP RESPONSE FPE RA! 02:The compliance basis of Section 3.4.1(c) has been changed to "Complies with Clarification."Calvert Cliffs Nuclear Power Plant (CCN PP) is utilizing the exception to 3.4.1 (c), which states:"Exception: Sufficient training and knowledge shall be permitted to be provided by anoperations advisor dedicated to industrial fire brigade support."CCNPP administrative procedures and the UFSAR ensure that an operations technical advisor,a licensed operator position, is dedicated to respond with the industrial fire brigade.FPE RA1 03:In the compliance bases in Attachment A for NFPA 805, Sections 3.10.1(2) and 3.10.3, thelcensee refers to a required action in Attachment S, Table S-2, item 18 of the LAR.Attachment S, Table S-2 does not include an Item 18; however, Attachment 5, Table S-2,Item 17 appears to address these elements. Confirm that Attachment 5, Table S-2, Item 17 isthe correct reference for the implementation item or provide the correct implementation item forthe Halon system actions identified in the LAR.CCNPP RESPONSE FPE RAI 03:Attachment 5, Table S-2, Item 17 is the correct reference for the implementation item identifiedin Attachment A for NFPA 805 Sections 3.10.1(2) and 3.10.3. The compliance basis forNFPA 805 Sections 3.10.1(2) and 3.10.3, have been revised to reference Attachment 5,Table S-2, Item 17.FPE RAI 04:Section 3.11.3(2), ,'Fire Barrier Penetrations," of NFPA 805 requires that fire dampers complywith NFPA 90A, "Standard for the Installation of Air-Conditioning and Ventilating Systems." InAttachment A, the licensee requested NRC approval for the use of a performance-basedmethodology described in Electric Power Research Institute (EPRI) TR- 1006756, "FireProtection Surveillance Optimization and Maintenance Guide for Fire Protection Systems andFeatures," to change the surveillance frequencies for fire dampers. Attachment L of the LAB,Approval Request 1, which is related to the use of performance-based methodology describedin EPRI TR-1006756, only includes NFPA 805, Section 3.2.3(1), as the NFPA 805 requirementthat is applicable.Clarify if Attachment L, Approval Request 1, is also applicable to NFPA 805, Section 3. 11.3(2),and revise Approval Request 1 as necessary to accommodate the additional section.2 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805CCNPP RESPONSE FPE RAI 04:The intent of Attachment L, Approval Request 1, is to request NRC approval for the option toutilize the performance-based methodology described in EPRI TR-1006756 for all fire protectioninspection, testing, and maintenance at CCNPP. Therefore, the approval request is applicableto NFPA 805 Section 3.2.3(1).Attachment A, Section 3.11.3(2) has been revised to remove the "Submit for NRC Approval"compliance statement. The additional existing "Complies, with Required Action" compliancestatement remains in Attachment A as damper inspection frequencies may potentially beupdated based on the guidance in EPRI TR-1006756 during or after implementation.FPE RA! 05:In Attachment L, Approval Request 2, the licensee proposed a performance-based approach toevaluate the acceptability of unprotected cables located above the suspended ceilings forcompliance with the requirements of NFPA 805, Section 3.3.5.1. Provide the followinginformation:a. Provide further details that describe the extent of use of extension cords that are locatedabove the suspended ceilings, such as number, length, size, use (e.g., types of electricalloads), and if the extension cords are for permanent or temporary use.b. Describe the administrative controls that are (or will be) in place to maintain the technicalbases for the request (e.g., prevent/limit future placement of ignition sources andcombustible materials, periodic surveillance above the ceiling, etc.).c. Clarify the following:i.If the Nuclear Safety Capability Assessment (NSCA) credited cables that are routed inmetal conduit above the suspended ceiling need to be free from fire damage in order tosupport a nuclear safety function or fire risk evaluation (FRE) for a fire in the fire areasdescribed in this request.ii. The NSPC discussion implies fire damage will not occur because, in part, the cablesare protected in metal conduit or in metal covered trays. Metal conduit and metal traysare not generally sufficient to protect cables from exposure fire damage. Provideadditional discussion and/or details that provide assurance that NSCA credited cablesare not susceptible to damage from extension cords or other potential fire hazards inthe area above the ceiling.d. The licensee appears to conclude that because defense-in-depth (DID) Echelon 1 issatisfied, that Echelons 2 and 3 are also satisfied. The NRC staff notes that DID is basedon a balance of the three echelons. Provide additional details related to how Echelons 2(fire detection and suppression) and 3 (safe shutdown) of the DID concept are maintained.CCNPP RESPONSE FPE RAI 05:05a -Response to be provided 4/13/15.05b -Response to be provided 3/11/15.3 ATTACHMENT (1)REQUEST FOR ADDITIONAL iNFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 80505c -i.The NSCA credited cables that are routed above suspended ceilings were evaluated on afire area basis to determine if their failure would result in a VFDR. Cables routed in metalconduit were not screened out of the analysis or considered to be free from the effects offire in the area; those cables above the suspended ceiling were evaluated as failed in theNSCA. There are some conduits routed above suspended ceilings that contain NSCAcredited cables. A few conduits contained NSCA cables that resulted in VFORs. ThoseVFDRs were evaluated in accordance with NFPA 805, Section 4.2.4.2, performance-based approach -fire risk evaluation with simplifying deterministic assumptions. The riskassessment concluded for each of these VFDRs that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFFA 805 Section 4.2.4, therefore, no furtheraction is required.ii. As stated in the response to part (i) of this response, the NSCA did not credit the metallicconduit as a means to prevent fire induced failure of NSCA-credited cables routed abovesuspended ceilings. The NSCA credited cables that are routed above suspended ceilingswere evaluated on a fire area basis to determine if their failure would result in a VFDR.Cables routed in metal conduit were not screened out of the analysis or considered to befree from the effects of fire in the area; those cables above the suspended ceiling wereevaluated as failed in the NSCA. NSCA cable failures that resulted in a VFDR wereevaluated in accordance with NFPA 805, Section 4.2.4.2, performance-based approach -fire risk evaluation with simplifying deterministic assumptions. NSCA credited cables aresusceptible to damage from extension cords and other potential fire hazards in the areaabove a ceiling and no assurance is given that metal conduit will protect those cables.05d -Response to be provided 4/13/15.FPE RAt 06:In Attachment L, Approval Request 3, the licensee requested the use of procedural guidancethat will allow performance of welding, cutting and other hot work in sprinklered fire areas whilethe suppression system is impaired, as an acceptable performance-based approach to complywith NFP'A 805, Section 3.3.1.3.1. Provide the following information:a. In the bases for the request: the licensee stated that this request is applicable to any firearea containing a sprinkler system, as identified in Attachment C, Table C-2. Discuss thebases for limiting this hot work procedure request to only fire areas that contain requiredfire sprinkler systems identified in Attachment C, Table C-2.b. Describe the hot work administrative controls for the fire areas that contain a suppressionsystem that is no._t identified as a required suppression system in Attachment C, Table C-2,and whether the administrative controls are different than those for fire areas with requiredfire suppression systems.c. In the bases for the request, the licensee stated that permanent combustibles locatedwithin 35 feet of the work area that cannot be removed must be covered with theappropriate style of blanket. Clarify if the "appropriate style of blanket" is a listed orapproved welding curtain, welding blanket, welding pad, or equivalent, as required byNFPA 51B, "Standard for Fire Prevention During Welding, Cutting, and Other Hot Work."4 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805d. Describe any additional actions/controls to be used when hot work is performed in fireareas/zones where one or more sprinkler systems are impaired above and beyond thosetaken for any other hot work activity conducted when sprinklers are in service.CCNPP RESPONSE FPE RAI 06:a. Attachment L, Approval Request 3, is applicable to any fire area containing a sprinklersystem and is not limited to 0nly fire areas that contain required fire sprinkler systems.Attachment C, Table 0-2 lists all sprinkler systems in plant fire areas, regardless ofwhether the system is required. The text of Attachment L, Approval Request 3, has beenrevised to clarify that the request is applicable to all fire areas containing fire sprinklersystems.b. As discussed in (a), above, Approval Request 3 is applicable to all fire areas containingfire sprinkler systems.c. Permanent combustibles located within 35 feet of the work area that cannot be removedmust be covered with an NFPA 51 B compliant blanket. The text of Attachment L,Approval Request 3, has been revised to clarify that an "appropriate style of blanket" is anNEPA 51 B compliant blanket.d. The Technical Requirements Manual (TRM) ensures that appropriate contingencymeasures are in place when TRM sprinkler systems are not in service. Thesecontingency measures, which are above and beyond those taken for any other hot workactivity conducted when sprinklers are in service, may include, but are not limited to,ensuring backup suppression is available (i.e., laying hose from an operable hose stationin an adjacent fire area). This is addressed by one of the bases for Attachment L,Approval Request 3, which states "Back-up fire suppression equipment is available inareas where the fire suppression system is inactive."Note: Editorial changes were made to the text of Attachment L, Approval Request.FPE RAi 07:NRC endorsed guidance NE) 04-02, "Guidance for Implementing a Risk-Informed,Performance-Based Program Under 10 CFR 50.48(c)," states that, where used in Chapter 3,"power block" and "plant" refer to structures that have equipment required for nuclear plantoperations, such as containment, auxiliary building, service building, control building, fuelbuilding, radiological waste, water treatment, turbine building, and intake structure, or structuresthat are identified in the facility's pre-transition licensing basis.Section 4.1.3 and Attachment i, Table I-1, "Definition of Power Block," of the LAR state thatbuildings that are required for nuclear plant operations (i.e., required to meet the nuclear safetyor radioactive release (BAD) performance criteria identified in Sections 1.5.1 and 1.5.2 of NFPA805) are considered within the power block. The licensee reviewed the plant for compliancewith the BAD performance criteria, and the results are documented in Attachment E, whichincludes the following compartments as screened in for BAD review, but are not described aspart of the power block in Attachment I, Table 1-1:* Interim Resin Storage Facility (Lake Davies)*Material Processing Facility5 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE______PROTECTION ASSOCIATION STANDARD 805 ~-* Office and Training Facility* Original Steam Generator Storage Facility,Pre -Assembly Facility (Upper Laydown Area)* Sewage Treatment Plant* Unit 1 Butler Building* Unit 2 Butler Building* Warehouse No. 3o West Road CageDescribe the basis for excluding these structures from the power block, based on the criteriastated in Section 4.1.3, " ... those that contain equipment required to meet the nuclear safetyand HAD criteria ... ,"an consequently, exclusion from the NFPA 805, Chapter 3 elements thatapply to the power block.CCNPP RESPONSE FPE RAI 07:Response to be provided 3/11/15.FPE RAI 08:Section 3. 11.1 of NFPA 805 requires that each major building within the power block beseparated from the others by barriers having a designated fire resistance rating of 3 hours or byopen space of at least 50 feet or space that meets the requirements of NFPA 80A,"Recommended Practice for Protection of Buildings from Exterior Fire Exposures." InAttachment A, the licensee stated that it "Complies with Clarification" and described that theNorth Service Building and Turbine Building are analyzed as one fire area in the NFPA 805NSCA, and are, therefore, treated as one building from a building separation perspective. Thelicensee also stated that it "Complies with Use of EEEE's" with respect to excluding the 45"-0"elevation of the North Service Building from the power block. The licensee did not discuss thebasis for excluding this specific elevation from the power block in Attachment I.Provide the basis for excluding the 45'-0" elevation of the North Service Building from the powerblock.CCNPP RESPONSE FPE RAI 08:Attachment I has been revised to discuss the basis for excluding the 45' elevation of the NorthService Building from the power block.The 45' elevation is excluded from Fire Area TB/NSB/ACA and the power block as justified byan engineering equivalency evaluation. The bases for acceptability are summarized as follows:* There are no cables or equipment required to achieve nuclear safety performance criteria(NSPC) in the 45' elevation of the North Service Building.* There are no cables or equipment required to achieve NSPC in the Yard within 50 feet ofthe 45' elevation of the North Service Building.A fire originating in the 45' elevation of the North Service Building will not impact cables orequipment required to achieve NSPO in adjacent fire area TB/NSB/ACA.6 A1-TACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805FPE RAI 09:Section 3.4.1l(a) of NFPA 805 requires that a fully staffed, trained, and equipped fire-fightingforce be available at all times to control and extinguish all fires on site. In Attachment A, thelicensee stated that in Section 5.5.B of Procedure SA-1-1O1, "If there are less than 5 brigademembers notify the Control Room."Current NRC guidance, Frequently Asked Question (FAQ) 12-0063, "Fire Brigade Make-Up"(ADAMS Accession No. ML 121980572), discusses conditions where fire brigade complementmay be less than the minimum for a period of time, in order to accommodate unexpectedabsence of on-duty shift members. Further, licensees may claim prior approval if their currenttechnical specifications or fire protection safety evaluation addresses the issue. If prior approvalhas not been granted, then the licensee should seek NRC approval in the NFPA 805 LMR.Pro vide additional detail on the compliance bases related to conditions when there are less than5 fire brigade members onsite.CCNPP RESPONSE FPE RAI 09:Response to be provided 4/13/15.FPE RA! 10:Section 3.11.5 of NEPA 805 requires that Electrical Raceway Fire Barrier Systems (ERFBS)that are required by NFPA 805, Chapter 4, be capable of resisting the fire effects of the hazardsin the area. In Attachment A, the licensee stated that there are no ERFBS credited forcompliance with Chapter 4, and, therefore, there is no compliance applicable to NEPA 805,Section 3.11.5. However, in Attachment B (Attributes 3.4.1.3, 3.4.1.5, 3.4.2.2, and 3.4.2.3) thelicensee described that one of the means of addressing cable impacts of fire damage is toprotect the cables by an ERFBS.a. Clarify if there were any cable resolutions in the NSCA that credit an EREBS to protect theaffected cables to meet NEPA 805, Chapter 4. If yes, then clarify if the ERFBS are incompliance' with NFPA 805, Section 3.11.5.CCNPP RESPONSE FPE RAI 10:LAR Attachment B documents the Nuclear Safety Capability Assessment Methodology Review.Sections 3.4.1.3, 3.4.1.5, 3.4.2.2, and 3.4.2.3 of LAR Attachment B identify that ERFBS may beutilized as an acceptable method to protect cables from fire damage and be credited within theanalysis. The NSCA has not credited any EREBS.Safe Shutdown Analysis (SSA) RA1 01:Attribute 3.2.1.2 of NE/ 00-0 1, Revision 2, includes the assumption that exposure fire damage tomanual valves and piping does not adversely impact their ability to perform their pressureboundary or safe shutdown function, and that any post-fire operation of a rising stem valvelocated in the fire area of concern should be well justified using an engineering evaluation. InAttachment B, the alignment bases for NEI 00-0 1, Attribute 3.2.1.2, states that manual valvesthat are repositioned for credited NFPA 805 recovery actions (RAs) are included in the NFPA805 NSPC equipment list and are subject to assessment of feasibility.7 ATT'ACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FiRE_______ PROTECTION ASSOCIATION STANDARD 805Provide the following information:a. Clarify if any rising stem valves involved in an RA are subjected to fire damage.b. If any of the valves in the fire area of concern being repositioned by an RA are rising stemnvalves, then clarify if an engineering evaluation was performed to evaluate the exposurefire damage to manual valves and piping to determine if the exposure to fire wouldadversely impact their ability to perform their pressure boundary or safe shutdownfunction. If used, describe the method and results obtained from the engineeringevaluation.CCNPP RESPONSE SSA RAI 01:a. All recovery actions as documented in LAR Attachment G were reviewed. There are norecovery actions that credit the manipulation of rising stem valves that have been exposedto the effects of fire.b. An engineering evaluation is not required as there are no recovery actions that require themanipulation of a rising stem valve that has been exposed to fire.SSA RAI102:Attribute 3.3.1.1.4 of NEI 00-01, Revision 2, includes criteria and assumptions for evaluatingpower cables for breaker coordination concerns and includes safe shutdown cables and thosenon-safe shutdown cables that can impact safe shutdown. In Attachment B of the LAR, thealignment bases for NEI 00-0 1, Attribute 3.3.1.1.4, states that the NSCA circuit identification andanalysis should utilize a "building block" approach and include only, as applicable, the powercable from the NSCA component to the upstream electrical power source.Provide the following information:a. Clarify if cables that supply loads not required to meet the NSPC off of the nuclear safetybuses are classified as "required" cables. if non-nuclear safety cables are not included,then provide the justification for not considering the failure of non-nuclear safety cables inmeeting the breaker coordination criteria for protection.b. The alignment basis states that plant modifications have been identified to achieveselective coordination of breakers/fuses and identified as implementation items inAttachment S, Table S-2. Identify the specific modifications that are required to achievethe selective coordination of breakers/fuses.CCNPP RESPONSE SSA RAI 02:Response to be provided 3/11/15.SSA RA1 03:Attribute 3.5.1.3 of NE/ 00-0 1, Revision 2, includes an assumption that circuit contacts areinitially positioned (i.e., open or closed) consistent with the normal mode/position of the"required for hot shutdown" equipment, and that the analyst must consider the position of the"~required for hot shutdown" equipment for each specific shutdown scenario when determiningthe impact that fire damage to a particular circuit may have on the operation of the equipment.In LAB Attachment B, the alignment basis for Attribute 3.5.1.3 states that the circuit analysismay discount spurious operation based on a fire affected cable being routed in a dedicated8 A1FrACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805conduit and the cable being protected from* external sources of voltage (also taking intoconsideration the potential impact from ground equivalent hot shorts).For multi-conductor cables routed in dedicated conduit, provide a description if intra-cable hotshorts (wire-to-wire shorts) are considered as a potential impact of fire damage on requiredposition of the NSCA equipment (i.e., the function of the initial position of circuit contacts are notaffected by intra-cable hot shorts).CCNPP RESPONSE SSA RAt 03:Response to be provided 3/11/15.SSA RAI 04:The nuclear safety goal described in NFPA 805, Section 1.3.1, is to provide reasonableassurance that a fire during any operational mode and plant configuration will not prevent theplant from achieving and maintaining the fuel in a safe and stable condition. In Section 4.2.1.2,the licensee stated that the NSCA will demonstrate that the plant can achieve and maintain safeand stable conditions for at least 12 hours with the minimum shift operating staff. After12 hours, the Emergency Response Organization (ERO) will be available to support "safe andstable" actions to extend hot standby conditions.a. in Section 4.2.1.2, subsection "Methods to Maintain 'Safe and Stable' and Extend HotStandbY Conditions," of the LAR, local manual actions are described to align varioussystems and functions. In Item No. 8, the licensee stated that should alternating current(AC) charging sources be lost, local manual operator action may be required, and thatstation batteries are capable of providing a minimum of 4 hours of 125 V direct currentpower to their respective loads during a station blackout without AC charging sources.The licensee further stated that this time allowance credits securing 11NVIT1 1 in thecable spreading room (CSR) within 45 minutes. Clarify if this local manual action iscredited as an RA in any fire area.b. In Section 4.2.1.2, subsection "Assessment of Risk," the licensee stated that the EROprovides sufficient resources for assessment of fire damage and completion of repairs toequipment necessary to maintain hot standby for an extended period, transition to coldshutdown, or return to power operations as dictated by the plant fire event. Describe ifany repair activities are necessary to maintain hot standby for an extended period (safeand stable conditions), including a detailed description of the specific repairs that would beneeded, the success path(s) being restored, and the time frame required to complete therepair.CCNPP RESPONSE SSA RA! 04:Response to be provided 3/11/15.SSA RA! 05:Section 2.4 of RG 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," Revision 1, dated December 2009 (ADAMS Accession No.ML092 730314) describes the treatment of RAs supplemented by guidance provided inNEI 04-02 and FAQ 0 7-0030, "Establishing Recovery Actions" (ADAMS Accession No.ML 110070485). In RG 1.205, the NRC staff clarifies that operation of alternative or dedicatedshutdown controls while the main control room (MCR) remains the command and control9 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805location would primarily be considered an RA because, for such scenarios, the dedicated oralternative controls are not considered primary. Attachment G of the LAR describes the primarycontrol stations (PCS) and identifies RAs performed at the PCS in Fire Area 16 (Unit 1 CSR and1C cable chase) and Fire Area 17 (Unit 2 CSR and 2C cable chase). Provide the followinginformation:a. Clarify if the control room remains the command and control location for a fire in FireAreas 16 and 17, and if so, discuss how the RAs at the PCS are evaluated for compliancewith NFPA 805, Section 4.2.4.b. In Fire Areas 16 and 17, there are RAs at the PCS that are not associated with a variancefrom deterministic requirement (VFDR).o For Fire Area 16, the RAs are:61CHECKRXSD1; 161CONSERVE1; 161SECHTR1 1_13; 161ADV1C43;1611C43CONTROL; and 161RCSTEMP1.* For Fire Area 17, the RAs are:1 71CHECKRXSD2; 1 71CONSERVE2; 1 71SECHTR2 1_23; 1 71AD V2C43;1712C43CONTROL; and 17IRCSTEMP2.Clarify the purpose of performing these RAs, and whether the actions are required to meetthe NSPC required by NFPA 805, Section 1.5.1.c. In Attachment G, Table G-1 of the LAR, disposition of VFDR 16-19-1 credits RAs at thePCS to energize pressurizer backup heater banks 11 and 13; however, another non-VFDR related BA (161SECHTR1 1_ 13) is credited to secure the pressurizer backup heaterbanks 11 and 13. Discuss how the contradicting RAs are evaluated in the NSCA.d. In LAR Attachment G, Table G-l, RAs are credited to disposition VFDRs 16-27-1 and1 7-25-2 to control atmospheric dump valve (ADV) hand valves to support control of theADVs at the PCS locations 1C43 and 2C43, respectively. However, the RAs(161ADVIC43 and 171ADV2C43) to control the ADVs at the PCS location do not have aVFDR associated with them. Discuss the method for crediting RAs to support the VFDRdisposition without crediting the BA at the PCS.CCNPP RESPONSE SSA RAt 05:Response to be provided 3/11/15.SSA RA1 05:Attachment W Tables W-6 and W-7 of the MAR appear to conflict with information described inAttachment C, Table C-i, and Attachment G, Table G-2. Clarify the following discrepancies:a. In Attachment C, Table C-i, Fire Area 34 is identified as transitioning deterministically inUnit 2 (Section 4.2.3.2 of NEPA 805) with no VFDRs identified. However, inAttachment W, Table W-7, Fire Area 34 is identified as transitioning using performance-based methods (Section 4.2.4.2 of NFPA 805), VFDRs are identified, RAs are credited,and the risk of the RA was calculated. Clarify the correct nuclear safety compliancestrategy for Fire Area 34.b. In Attachment C, Table C-1, the following fire areas are identified as transitioningdeterministically with no VFDRs identified. However, Attachment W, Table W-6 (for10 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805Unit 1) and Attachment W, W-7 (for Unit 2) identify that these fire areas have VFDRsidentified. Further, an FRE was performed that calculated a delta core damage frequency(CDF) and delta large early release frequency (LERF) value as follows:Unit 1: 2, 8, 13, 18, i8A, 22, 23, 25, 26, 27, 28, 31, 38, 40, and 2CNMTUnit 2: 3, 4, 6, 14, 15, 19, 19A, 21, 30, 33, 39, and 1CNMTFor each of these fire areas, clarify the correct nuclear safety cornpliance strategy, andjustify the bases for performing an FRE that is not discussed in the NSCA in LARAttachment C, Table C-i, and the bases for crediting RAs that are not included in LARAttachment G, Table G-1.c. in Attachment C, Table C- 1, the following fire areas are identified as transitioning usingperformance-based methods (FRE) to meet the NSPC, and no RAs were credited (eitherfor risk or DID). However, Attachment W, Table W-6 (for Unit 1) and Attachment W, W-7(for Unit 2) identify these fire areas as crediting RAs and the risk of the RA was calculated:Unit 1: 12, 14, 15, 19A, 21, 30, 32, 33, 35, 36, 39, 1 CNMT, and ISUnit 2: 12, 13, 18A, 20, 26, 27, 28, 32, 34, 35, 36, 40, 2CNMT and ISFor each of these fire areas, clarify the correct nuclear safety compliance strategy forthese fire areas and the bases for crediting RAs that are not included in Attachment G,Table G- 1.d. In Attachment C, Table C-i, the following fire areas are identified as transitioning usingdeterministic methods to meet the NSPC, and no RAs were credited (either for risk orDID). However, Attachment W, Table W-6 (for Unit 1) and Attachment W, W-7 (for Unit 2)identifies these fire areas as crediting RAs and the risk of the RA was calculated:Unit 1: 13, 18, 18,A, 22, 23, 25, 26, 27, 28, and 2CNMTUnit 2: 14, 15, 19, 19,4,21, 30, 33, 39 and 1CNMTFor each of these fire areas, clarify the correct nuclear safety compliance strategy andjustify the bases for not including these RAs in Attachment G, Table G-1, if these RAs areactually credited in the NSCA.CCNPP RESPONSE SSA RAI 06:Response to be provided 4/13/15.SSA RAI107:Modifications were identified in Attachment S, Table S-2, that appear to resolve certain VFDRissues. However, the disposition of the certain VFDRs as summarized in Attachment C,Table C-i1, do not describe whether the modification was credited or not. Provide *clarification on*how the modifications described below were addressed in the disposition of the VFDRs listed:a. Attachment S, Table S-2, Item 7, involves modifying control circuits for the PressurizerPower Operated Relief Valves (PORVs), 1(2)ERV402 and 1 (2) ER V404, to prevent thePORVs from spuriously opening. However, VFDRs 1 6-46-1, 24-26-1, 16-47-1, 24 -2 7-1,17-41-2, 24-63-2, 17-42-2, and 24-64-2 involve fire damage to cables which could result inspurious opening of the Pressurizer PORV, and the VFDR dispositions credits an RA forDID.11 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805b. Attachment S, Table S-2, Item 8, involves modifying the control circuits for the auxiliaryfeed water (AEW) steam admission valves 1(2)CV4070 and 1(2)CV4071 to ensureadequate separation such that one set of valves will be available during a fire in either theCSR or switchgear rooms. However, VFDRs (16-22-1, 17-16-2, 16-26-1, and 1 7-26-2)involve fire damage to cables that could cause the loss of control and/or spuriousoperation of i(2)CV4070 and 1 (2)4071, and the VEOR dispositions credit an BA either toreduce risk (VFDRs 16-22-1 and 17-16-2) or for DID (VFDRs 16-26-1 and 1 7-26-2).c. Attachment S, Table S-2, Itemn 11, involves modifying control circuits for the Main SteamIsolation Valves (MS/Vs), 1(2)CV40430P and 1(2)CV404#OP, to ensure at least onesolenoid dump valve can be energized to close the MS/ Vs. However, VFDRs 1 6-31-1,16-32-1, 1 7-23-2, and 1 7-24-2 involve fire damage to cables that could cause a loss ofcontrol and/or spurious operation of the associated MS/V, and the VFDR dispositionscredit an BA for DID (VFDRs 16-31-1, 1 6-32-1, 1 7-23-2, and 1 7-24-2).CCNPP RESPONSE SSA RAI 0}7:Response to be provided 4/113115.SSA RAI 08.:In Attachment K, Licensing Action 5, the licensee requested that a previously approvedexemption, related to dedicated water curtains as being adequate to maintain the 3-hour firerating of barriers, be transitioned to the NFPA 805 program. The licensee described thesprinkler systems located in Room 216A and Room 106 as supplying the sprinkler heads for thededicated water curtains. In the summary of the exemption approved by the NRC in a letterdated March 15, 1984 (ADAMS Accession No. ML0 10430325), the licensee stated that on theCorridor No. 110 side of the hatch, a dedicated sprinkler head will be supplied from the RoomNo. 116 sprinkler system. However, in the Baltimore Gas & Electric Company submittal datedNovember 21, 1983 (ADAMS Accession No. 8311290159), the licensee stated that on thecorridor No. 110 side of the hatch, a dedicated sprinkler head will be supplied from the RoomNo. 106 sprinkler system. The NRC staff also noted that Attachment C, Table C-i, refers toroom numbers in the "Required Fire Protection System and Features," adAttachment C, TableC-2, refers to fire zones. The NRC staff also noted in the discussion for Licensing Action 1 thatroom numbers at the plant may have changed over time. Provide the following information:a. Describe if the fire zone numbers listed in Attachment C, Table C-2, are the same as theroom numbers listed in the fire area summary in Attachment C, Table C-1. Describe if theroom numbers in Attachment C correspond with the room numbers cited in the previouslicensing actions in Attachment K.b. Provide a description of the water curtain arrangement, including the sprinkler systemsthat supply the required sprinkler heads using the current terminology for rooms, fireareas, and/or fire zones such that the staff can fully understand the installation and howthe installation is represented in the various tables in the submittal and the previouslicensing actions.CCNPP RESPONSE SSA RAl 08:a. The "fire zone" numbers listed in Attachment C, Table 0-2 are the same as the "roomnumbers" listed in the fire area summary in Attachment C, Table C-1. The "roomnumbers" in Attachment C also correspond with the "room numbers" cited in the previous12 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805licensing actions in Attachment K. There is a typographical error in Section II of theEnclosure to NRC letter dated March 15, 1984 (ADAMS Accession No. ML01 0430325),which summarizes the exemption request. This is discussed in part (b.), below.b. Attachment C, Table C-i and Table C-2, identify water curtain sprinkler in room 110 asprotecting the hatch to room 216A and the water curtain sprinkler in room 216A asprotecting the hatch to room 110.The room 216A water curtain is supplied by the room 216A sprinkler system as describedin the BG&E exemption request dated November 21, 1983 (ADAMS Accession No.8311290159) and in the summary of the exemption approved by the NRC in the letterdated March 15, 1984 (ADAMS Accession No. ML01 0430325).The room 110 water curtain is supplied by the room 106 sprinkler system as described inthe BG&E exemption request dated November 21, 1983 (ADAMS Accession No.8311290159). There is a typographical error in Section II of the Enclosure to NRC letterdated March 15, 1984 (ADAMS Accession No. ML01 0430325) which inaccuratelydescribes the system as being supplied by the sprinkler system in room 116. The systemhas been confirmed to be supplied by the sprinkler system in room 106.SSA RA1 09:/In Attachment C, Table C-i, under the heading "Fire Suppression Effects on Nuclear SafetyPerformance Criteria," the majority of the fire areas contain the concluding statement, "Firesuppression in this fire area will not impact the ability to achieve the NSPC in accordance withNFPA 805, Sections 4.2.1 and 4.2.4.1.5." NFPA 805, Section 4.2.4.1.5, is associated with thefire modeling performance-based approach, which the licensee stated it did not use inSection 4.5.2.1 of the LAR. In addition, the suppression effects sections for several other fireareas (e.g., 18A, 20, 21, 22, 23, 35, and 36) contain the statement, "There is no suppressioneffect concern for this fire area as the fire area does not contain NSCA equipment," yet the firearea contains VFDRs. Address the following:a. Clarify the basis for discussing the fire suppression effects for a fire modelingperformance-based approach when the fire areas used a risk evaluation performance-based approach.b. Provide additional discussion for those fire areas where VFDRs are identified, but thesuppression effects discussion states there is no NSCA equipment in the fire area.CCNPP RESPONSE SSA RAI 09:Response to be provided 3/11/15..SSA RAI 10:In Section 4.5.2.2, the licensee stated that there are no VFDF~s that involved performance-based evaluations related to wrapped or embedded cables. However, in Attachment C,Table C-i, Fire Areas 18, 19, 35, 36, and TB/NSB/ACA are performance-based fire areas andcredit EEEE, "ECP- 13-000359 -"Generic Letter (GL) 86-10 Evaluation of Embedded Conduit inthe Turbine Building and Barrier Thickness of the Floor/Ceiling Barrier between AB-4/AB-5 and517/518," which justifies the acceptability of conduits embedded in the Turbine Building floorslab (elevation 279), the floor/ceiling slab between stairwells AB-4 and AB-5, and the horizontal13 ATrACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805cable chases (Rooms 517 and 518). Clarify if the disposition of the VFDRs in Fire Areas 18, 19,35, 36, and TB/NSB/ACA credit the embedment as evaluated in the EEEE.CCNPP RESPONSE SSA RAi 10:Response to be provided 3/11/15.:SSA RAI 11:in Attachment C, Table C-1, the licensee identified Marinite boards as fire protection featuresthat are credited for "S" (required for Chapter 4 separation criteria) and "R" (required for risksignificance) to protect cables for a fire in Fire Area 1CNMT (Unit 1 Containment) and 2CNMT(Unit 2 Containment). Provide the following information:a. Describe the extent that Marinite boards are credited for Chapter 4 separation ("S") andfor risk significance ("R'f) in the Unit 1 and Unit 2 Containments. In addition, describe thedesign and plant configuration of the Marinite boards and the nuclear safety functions thatthe passive fire protection features are protecting.b. Provide previous NRR staff approval (if any) for the use of Marinite boards in containmentto demonstrate meeting the requirements of Appendix B, Section III.G.2, which can becredited to meet the requirements of NEPA 805, Section 4.2.3.4, or evaluate acceptabilityusing a performance-based analysis approach in accordance with NFPA 805,Section 4.2.4.CCNPP RESPONSE SSA RA! 11:Response to be provided 4/13/15.SSA RAI 12:In Attachment C, Table C-2, the licensee makes reference to "Unit 1 Containment (App-BPurposes Only)" and "Unit 2 Containment (App-B Purposes Only)," for fire protection systemsand features. The fire protection systems and features are identified as required for "S"(Re quired for Chapter 4 Separation Criteria), "R" (Required for Risk Significance), and/or "D"(maintain an adequate balance of DID in a change evaluation or FRE).Clarify the meaning of 'Appendix-B Purposes Only" and if these fire protection systems andfeatures are credited with respect to compliance with NFPA 805, Chapter 4.CCNPP RESPONSE SSA RAI 12:The containment areas were divided into rooms under the Appendix R program. The currentdescriptions for these rooms, as identified in plant documents, contain "(App-R Purposes Only)"within their name. The phrase "(App R Purposes Only)" will be removed from plantdocumentation upon implementation of NFPA 805. The phrase can be found in the followingsections of the LAR:* In attachment C, Table C-1 -NEI 04-02 Table B-3 Fire Area Transition* In attachment C, Table 0-2 -Summary of NFPA 805 Compliance Basis and RequiredFire Protection Systems and Features* In attachment E, NEI 04-02 Radioactive Release Transition14 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE____________PROTECTION ASSOCIATION STANDARD 805-The fire protection systems and features are credited with respect to compliance with NFPA 805Chapter 4. The containment fire areas will remain subdivided as described in plant documents.SSA RAI 13:Provide the following pertaining to non-power operations (NPO) discussions provided inSection 4.3 and Attachment D:a. Section 4.3.2 and Attachment D state that incorporation of the recommendations from the"KSF [key safety function] pinch point" evaluations into appropriate plant procedures priorto implementation will be done to ensure the requirements of NFPA 805 are met. Identifyand describe the changes to outage management procedures, risk management tools,and any other document resulting from incorporation of KSFs identified as part ofNFPA 805 transition. Include changes to any administrative procedures such as "Controlof Combustibles."b. For those components that had not previously been analyzed in support of the at-poweranalysis or whose functional requirements may have been different for the non-poweranalysis, cable selection was performed in accordance with approved project procedures.Provide a list of the additional components and a fist of those at-power components thathave a different functional requirement for NPO. Describe the difference between the at-power safe shutdown function and the NPO function. Include with this list a generaldescription by system indicating why components would be selected for NPO and not beincluded in the at-power analysis.c. Section 4.3.1 and Attachments D and H- state that the NPO analysis was performed inaccordance with FAQ 0 7-00410, "Non-Power Operations Clarifications (ADAMS AccessionNo. ML 082200528). However, the LAR did not provide the results of the KSF pinch pointanalysis. Provide a list of KSF pinch points by fire area that were identified in the NPO firearea reviews using FAQ 0 7-0040, including a summary level identification of unavailablepaths in each fire area. Describe how these locations will be identified to the plant staff forimplementation.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 of anyactions being credited to minimize the impact of fire-induced spurious actuations on poweroperated valves (e.g., air-operated valves and motor-operated valves) during NPO(e.g., pre-fire rack-out, actuation of or pinning of valves, and isolation of air supplies).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 KSFs areachieved via RAs or for which instrumentation not already included in the at-poweranalysis is needed to support RAs required to maintain safe and stable conditions.Identify those RAs and instrumentation relied upon in NPO and describe how RAfeasibility is evaluated. Include in the description whether these variables have been orwill be factored into operator procedures supporting these actions.15 ATTACHMENT (1)REOUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE__________ PROTECTION ASSOCIATION STANDARD 805CCNPP RESPONSE SSA RAI 13:13a -Plant procedure updates are to be completed during NFPA 805 implementation. Theseupdates will include procedures such as NO-1-103, Conduct of Lower Mode Operations, toincorporate Key Safety Function (KSF) Pinch Point Analysis as identified in LARAttachment S, Table S-3 IMP-4. Changes will include the following:* Limiting/prohibiting hot work in select Fire Zones.* Detection/Suppression Systems should be verified to be functional, (not tagged out etc.).SLimiting/Prohibiting the hazard of combustible materials.a Using alternate equipment and/or the equipment's position whenever removing power.*Appropriate compensatory measures required during periods of increased vulnerability.=Activities that may impact KSFs should be limited and strictly controlled to mitigate losses.* Consider the hazards from the introductions of combustible materials and sources of fireprecursors.* Limiting work during periods of High Risk Evolution (HRE) conditions.&deg; Ensure HRE are identified in a manner consistent with NUMARC 91-06 and FAQ07-0040.13b -Report NFPA 805-00008, Section 4 provides an overview of the Non-Power Operations(NPO) model development methodology, including cable selection. Section 7 of this reportdocuments systems included and excluded. Attachment A to this report provides detaileddescription of the NPO model. Attachment C to this report contains the NPO Equipment Listincluding required/credited function. NPO only components appear in this attachment but notin NFPA 805-00006, Nuclear Safety Capability Assessment (NSCA) Attachment 7-5, theNSCA Equipment List. Functional differences can be identified by comparing the requiredNSCA and NPO positions which are identified in these reports and also in a plant database.Cable selection packages for components credited in the NPO evaluation that have adifferent function from the function required by the NSCA, were reviewed, evaluated andupdated as required, for all credited functions. Differences in equipment and functions aretypically attributable to the difference in plant operating state.Some example systems where a change in state or different equipment selection may occurinclude:&deg; Process Monitoring -Different instruments are required due to differences in plantoperating state and differences in credited systems (e.g. Shutdown Cooling).* Shutdown Cooling (Low Pressure Safety Injection) -Credited for decay heat removal(DHR) KSF in NPO, not credited in NSCA.* Shutdown Cooling Isolation Valves -Required closed High Low Pressure interface forthe NSCA, required open for DHR KSF in NPO.* High Pressure Safety Injection -Credited for Inventory KSF in NPO, not credited inNSCA.* Auxiliary Feedwater System -Required operable for DHR in NSCA, not credited inNPO.13c -Report NFPA 805-00008, Attachment B documents the results of the pinch point analysis.This attachment identifies the key safety functions (KSFs) that are evaluated and the status16 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805 __of each path that could be used to accomplish the KSF for each fire area. As identified inLAR Attachment D, there were 51 fire areas for which the evaluation identified one or morepinch points for Unit 1 and 46 fire areas for which the evaluation identified one or more pinchpoints for Unit 2. Thirty-five fire areas were identified to have at least one pinch point for bothunits. Report NFPA 805-00008, Attachment B will be used as a reference document insupport of site procedure updates, as discussed in response to item (a) of this RAI, and willbe available during outage planning and HRE fire risk mitigation reviews.13d -Response to be provided 3/11/15.13e -Response to be provided 3/11/15.13f -Recovery actions have not been credited as the sole means of mitigating KSF pinchpoints. However, recovery actions have not been excluded as a method of mitigating fireimpact to KSFs. Recovery actions have been evaluated for several failure modes includingloss of HVAC systems, loss of Instrument Air and loss of control room indicators (where localor backup indication is available). These recovery actions were evaluated using existingplant procedural guidance which will be reviewed and updated as necessary during NFPA805 implementation. Any recovery actions that will be implemented during a HRE will beevaluated for feasibility in a manner consistent with NSCA credited recovery actions.SSA RAI 14:Describe if any RAs require the cross-tie of Unit 1 and Unit 2 systems to achieve the NSPC.Provide the following information:a. Describe whether these cross-connecting RAs require staff from both units. If so,describe how the feasibility analysis reflects the Unit 1 and Unit 2 staffing, communication,and operational interface.b. Describe the operational impacts (by fire), if any, on the unaffected unit created by cross-tying these systems. Describe whether Technical Specification 3.0.3 is entered once thecross-tie with the opposite unit has been completed for fire safe shutdown.CCNPP RESPONSE SSA RAI 14:The only recovery actions that credit a cross-tie between Unit 1 and Unit 2 systems to achievethe NSPC are cross-connecting air systems. Recovery actions to cross-tie the Unit 1Instrument Air system to the Unit 2 Plant Air system are credited for fires in Fire Areas 19, 20and 34.. Recovery actions to cross-tie the Unit 2 Instrument Air system to the Unit 1 Plant Airsystem are credited for fires in Fire Areas 18, 22 and 25. In each area a recovery action wascredited for the VFDR to reduce the risk due to fire in that area.These recovery actions are updates to Table G-1 which currently credits N2 recharge.Additional changes to Table G-1 are anticipated as a result of other RAIs being resolved atdifferent RAI milestone response requirements. LAR Attachment G, Table G-1, RecoveryActions and Activities Occurring at the Primary Control Station(s) will be updated and a markupprovided with the 120 day submittal (4/13/15) that is required to support responses for additionalRAIs.17 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FiREPROTECTION ASSOCIATION STANDARD 805a. Recovery action to cross-tie plant air systems requires staff from both units. The fireimpacted unit staff will direct/request the supporting unit to perform the requiredalignment. The supporting unit staff will then operate the necessary equipment undertheir cognizance and report back to the fire impacted unit operators. The feasibilityanalysis identifies recovery actions, including sub-steps, by unit and operator to ensureadequate staffing. Communications between the units is specifically directed by plantprocedures (fire AOPs) and this process will be maintained by updated procedures.b. There are no recovery actions which credit unit cross-tie which require entry intoTechnical Specification LCO 3.0.3.SSA RAI 15:In Attachment B of the LAR, the alignment basis discussion for Attribute 3.2.1.2 provides thefollowing statement on possible fire damage to instrument air tubing that includes copper tubingwith soldered joints that are susceptible to separation during a fire and could cause the loss ofinstrument air to components:These affects were evaluated on an area basis to determine if the instrument air systempressure could be maintained. Calculation CA07971 demonstrates that the instrument airsystem can maintain system pressure with a 1 inch line pipe rupture.Calculation CA07971 states, "Evaluation Of Maximum Air Line Break Size in Which NominalInstrument Air Pressure Can Be Maintained at 50 psig." The NRC staff noted apparentdiscrepancies in the use and recovery of instrument air as described in Attachment C. Providethe following:a. Provide justification that 50 psig of instrument air pressure will not prevent instrument airoperated valves from changing position.b. Provide justification for limiting the size of the line to 1" soldered joints being susceptibleto separation during a fire. Describe the soldered joints used in the plant instrument airsystem. For any soldered joints larger than 1 ", describe how they were treated in theNSCA and Fire Probabilistic Risk Assessment (PRA).C. For several fire areas in Attachment C (such as Fire Areas 18, 19, 20, 21, and 22), thelicensee stated in the method of accomplishment for the vital auxiliaries performance goalthat instrument air may be recoverable from the opposite unit plant air system. However,the VFDRs associated with the fire areas (such as VFDRs 18-16-2, 19-01-1, 20-02-1,2 1-02-1, and 22-05-2) state that plant air from the opposite unit cannot be used becauseof failure of 1CV2061 or 2CV2061, and the VFDR? disposition credits an RA that involvesaligning backup nitrogen to the affected unit control valves. Clarify the discrepancybetween the method described in the subject fire areas for achieving the performancegoal, the VFDRs that state this method is not available, and the RAs cited in LABAttachment G for resolution of the VFDRs.d. In Attachment C, the discussion of fire suppression effects on the NSPC for Fire Areas 39and 40 addresses the impact of suppression damage to redundant instrument aircompressors and the saltwater air system, and states that the AFW air accumulators canbe charged from the nitrogen system with an RA. However, the disposition of VFDRs39-01-1 and 40-01-2, which address fire damage to the respective unit's instrument airsystem, stated that the VFDR has been evaluated with no further action required. Inaddition, the RA to align the nitrogen system to the AFW air accumulators is not discussed18 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE__________ PROTECTIONASSOCIATION STANDARD 805 _in LAR Attachment G for these fire areas. Clarify the apparent discrepancy between theeffects of fire damage and suppression damage on the instrument air system and saltwater air compressors (SWAG) with regard to the need for an RA. If an RA is necessaryto mitigate the suppression effects on the instrument air compressors and SWAG, thendescribe the feasibility and additional risk of the RA.CCNPP RESPONSE SSA RAI 15:15a -Response to be provided 4/1 3/1 5.15b -Soldered joints are only used at endpoint/load connections within the instrument airsystem. There are no soldered joints equal to or larger than 1" in the system. Any joints inthe system of this size are pipe fittings. The parts of the system that could fail are the "soft"'components connecting the instrument air distribution system to their loads. Each load thatcould be fire affected for a given fire area was reviewed. The cumulative impact of thesefailures was evaluated to determine if the system would blowdown due to the rupture beinggreater than the makeup capacity of the operable air compressors. This analysis was doneon a deterministic basis and in each case, where system blowdown could occur, it wasidentified with a VFDR.15c -Response to be provided 4/13/1 5.i15d -Response to be provided 4/1 3/1 5.SSA RAI 16:In Attachment C, the licensee stated in the summary of vital auxiliaries for Fire Area 178 thatthe control room and CSR heating, ventilating, and air conditioning (HVAC) is not availablewithout an RA and referenced VFDR 178-01-0. However, the disposition discussion for VFDR178-01-0 states that no further actions are required based on the performance-based analysisfor the VFDR, and no RAs required for risk or DID were identified in Attachment G. Clarify thebases for the discrepancy between the description of the vital auxiliaries' discussion and theVFDR disposition.CCNPP RESPONSE SSA RAI 16:The deterministic NSCA evaluation identified that the Control Room and Cable Spreading RoomHVAC System could be affected for a fire in Fire Area 1 7B. This failure was documented withVFDR 17B-01-0. The VFDR was evaluated in accordance with NEPA 805, Section 4.2.4.2,performance-based approach -fire risk evaluation with simplifying deterministic assumptions.The results of this fire risk evaluation determined that the risk, safety margin, and defense-in-depth, meet the acceptance criteria of NFPA 805 Section 4.2.4, with no further action required.A recovery action is not required for this VFDR. The Vital Auxiliary section of Fire Area 17B,Attachment C will be updated to state, 'Control Room and Cable Spreading Room HVAC maynot be available."SSA RAI 17:In .Attachment G, there are numerous RAs to provide portable fans for temporary cooling ofswitchgear rooms for Unit 1 Fire Areas 11, 16, 17, 18, and 20, and for Unit 2 Fire Areas 22, 25,19 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 80534, and yard. Plant procedures indicate the use of portable generators to power the fans ifnormal power is not available. Provide the following additional in formation:a. Describe the location of the portable generators and the location of NSCA structures,systems, and components (SSCs), if any, in the vicinity of these location(s), In yourdescription, include a summary of the procedure guidance for the use of portable gasgenerators and how the BA aligns with each of the feasibility criteria of FAQ 07-0030(i.e., training, procedures, drills, etc.).b. Describe the type of fuel and quantity associated with the portable generators and theavailability and the location(s) of sufficient fuel sources to support maintaining safe andstable conditions for the time period required.c. Provide justification that refueling the generators does not present a fire exposure hazardto NSCA SSCs.d. Describe the installation of temporary power cables, connections to distribution panels,and any disruptions to fire area boundaries.e. Describe the method (e.g., the analyzed ventilation path configuration) of providingtemporary cooling when portable fans are used for these RAs.CCNPP RESPONSE SSA RA1 17:Response to be provided 4/13/15.RAD RAI 01:The radioactive material (RAM) described in the CENG [Constellation Energy Nuclear Group]Calculation No. CA07953 provides a quantification of the maximum amount of RAM that may bestored in various areas. Provide information, if'any, on site procedures that are (or wiil be)established to limit the amount of RAM in storage containers to the levels identified in theanalyses (e.g., West Road Cage area, Warehouse #3, Pre-Assemb/y Facility, and UpperLaydown Area).CCNPP RESPONSE RAD RAi 01:Response to be provided 3/11/15.RAD RAi 02:Provide information, if any, on site procedures that establish operational controls to restrict theopening of storage containers in open, uncontained areas (e.g., West Road Cage area,Warehouse #3, Pre-Assemb/y Facility, and Upper Laydown Area).CCNPP RESPONSE RAD RAI 02:Response to be provided 3/11/15.RAD RAI 03:In the Upper Laydown Area, there are "sealed" Sealand containers, casks, and othercontainers. Describe what is meant by "~sealed" (e.g., are the containers locked and access isnot allowed, and do site procedures prevent the opening of these containers?). Also, describe20 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE________PROTECTION ASSOCIATION STANDARD 805how potential effluent will be contained based on the "sealing" of containers and concluding thatthere will be negligible RAD.CCNPP RESPONSE RAD RAI 03:Response to be provided 3/11/15.RAD RA1 04:Describe any compensatory actions that may be taken during fire suppression activities tominimize RAD (e.g., diking of liquid effluent, use of storm drain covers, radioactive monitoring,or use of other gaseous effluent controls (e.g., use of eductors, effluent filtration)).CCNPP RESPONSE RAD RAI 04:Response to be provided 3/11/15.Fire Modeling (FM) RA1 01:Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall beacceptable to the NRC. The NRC staff noted that fire modeling comprised the following:*The algebraic equations implemented in Fire Dynamics Tools (FDTs) were used tocharacterize flame radiation (heat flux), flame height, plume temperature, ceiling jettemperature, and hot gas layer (HGL) temperature, and the latter in the multi-compartment analysis (MCA).*Fire Dynamics Simulator (EDS) was used to assess MCR habitability, to calculatetemperatures and heat fluxes for damage assessment to critical targets in selectedcompartments, calculate the flame height and how that affected certain targets, andcalculate temperature rise for the purposes of estimating smoke detector activation.*The Thermally-Induced Electrical Failure model, as part of EDS, was used as a secondarycheck on the temperature and heat flux calculations using EDS for zone of influence (ZOI)purposes.Section 4.5.1.2, "Fire PRA" of the LAR states that fire modeling was performed as part of theFire PRA (FPRA) development (NFPA 805, Section 4.2.4.2). Reference is made toAttachment J, "Fire Modeling V& V,"3 for a discussion of the acceptability of the fire models thatwere used.Regarding the acceptability of the PRA approach, methods, and data:a. Identify whether any fire modeling tools and methods have been used in the developmentof the LAR that are not discussed in Attachment J. In addition, identify any fire modelingtools and methods that are discussed in Attachment J that were not used in the firemodeling analyses performed at the plant.b. It is discussed in the detailed fire modeling analysis that, "the FDTs are not setup forsecondary ignition or for the effects of suppression systems on a fire scenario." Thisimplies that secondary combustibles were not considered for any fire modeling analysis atthe plant, except those using FDS.21 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805c. Provide justification for ignoring the effects of flame spread and fire propagation insecondary combustibles (for example, cable trays) and the corresponding heat releaserate (HRR) on the calculated ZOl and HGL temperature.d. Provide in formation on how non-cable intervening combustibles were identified andaccounted for in the fire modeling analyses.e. Typically, during maintenance or measurement activities in the plant, electrical cabinetdoors are opened for a certain period of time. Explain what administrative controls are inplace to minimize the likelihood of fires involving such a cabinet, and describe howcabinets with temporary open doors were treated in the fire modeling analyses.f. Describe the criteria that were used to decide whether a cable tray in the vicinity of anelectrical cabinet will ignite following a high energy arcing fault (HEAF) event in thecabinet. Explain how the ignited area was determined and subsequent fire propagationwas calculated. If applicable, describe the effect of tray covers and fire-resistant wraps onHEA F-induced cable tray ignition and subsequent fire propagation.g. Provide justification for the assumed fire areas and elevations that were used in thetransient ZOI calculations. Explain how the model assumptions in terms of location andHRR of transient combustibles in a fire area or zone will not be violated during and post-transition.h. Explain how wall and corner effects were accounted for in the fire modeling calculations,or provide justification if these effects were not considered.i.Specifically regarding the use of FDS in the MCR abandonment calculations:I. It appears that the ceiling height of the MCR used in the calculations is rather high(- 17ft.). Explain how the MCR dimensions specified in the FDS in put files wereestablished, and confirm that they are consistent with the actual dimensions of thecontrol room. In addition, if a false ceiling is present to separate the interstitial spaceabove the operator and back panel areas, provide justification for ignoring it in thecontrol room abandonment calculations.ii. Explain if the doors of the MCR were assumed to be closed or open at all times, orwere assumed to be open at a specified time. Discuss the impact of this assumptionon the calculated abandonment times. Describe the additional leakage paths thatwere specified in the FDS input files, and provide the technical basis for the assumednatural vent areas.iii. The abandonment calculations consider two mechanical ventilation modes: HVACinoperative and HVAC in smoke purge mode. Explain why the normal HVAC modewas not considered in the analysis, and why the two modes that were considered arebounding.iv. The MCR abandonment calculations for a specified ignition source appear to includeFDS runs for 10 HRR bins. Appendix-E of NUREG/CR-6850, "EPRi/NRC-RES FirePRA Methodology for Nuclear Power Facilities: Volume 1: Summary and Overview,"September 2005 (ADAMS Accession No. ML0525800 75) uses a 15-bin discretization.Explain why only 10 bins were used, and describe how the 10-bin discretization wasestablished.v. Describe the technical basis for choosing the location of the ignition source in theelectrical cabinet and transient fire scenarios that were modeled in FDS, and confirm22 A1TrACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805that locations in the opera tar area and the back panel area were considered for bothtypes of ignition sources. Provide technical justification for not considering firescenarios with the ignition source against a wall or in a corner.vi. Explain how the area and elevation of electrical cabinet and transient fires weredetermined, and demonstrate that the assumed areas and elevations are consistentwith plant conditions or lead to conservative estimates of the abandonment times.vii. Provide justification for not considering scenarios that involve secondarycombustibles in the MCR abandonment calculations.viii. Explain how the HRRs for electrical cabinets were determined and whether thevalues are consistent with the type(s) of cabinets present in the MCR at the plant.ix. Provide technical justification for not considering electrical cabinet fires thatpropagate to adjacent cabinets.x. Provide the technical basis for the material properties that were specified in FDS forthe cables inside the cabinets in the MCR. Provide confirmation that the assumedsoot yield and heat of combustion values (the latter either explicitly or implicitlythrough the specified fuel composition) lead to conservative estimates of the sootgeneration rate.xi. Describe the transient fire growth rate(s) used in the control room abandonmentcalculations and provide the technical basis for the assumed time(s) to peak HRR.xii. Provide the technical basis for the material properties that were specified in FDS forthe transient combustibles in the MCR. Provide confirmation that the assumed sootyield and heat of combustion values (the latter either explicitly or implicitly through thespecified fuel composition) lead to conservative estimates of the soot generation rate.xiii. Describe the habitability conditions that were used to determine the time to MCRabandonment. FDS "devices" (temperature and optical density) were placed at aheight of 6 feet and at four different locations in the MCR. Describe the basis forchoosing these locations and demonstrate that these locations are eitherrepresentative of where operators are expected to be, or lead to conservativeabandonment time estimates. Confirm that heat flux sensors were not specified and,if so, provide technical justification for using temperature sensors as a surrogate forheat flux sensors.xiv. Variations in the input parameters such as ambient temperature, soot yield of the fuel,fire base height, etc., affect the output of FDS calculations. The abandonmentanalyses for the MCR were performed using a single set of input parameters for eachscenario. Demonstrate that the FDS calculations obtained using this set of inputparameters provide conservative or bounding results. Alternatively, demonstrate thatthe abandonment times for a given scenario are not sensitive to variations within theuncertainty of the input parameters.xv. Explain how the results of the MCR? abandonment time calculations were used in theFPRA.j. Specifically regarding the MCA:i. Describe the criteria that were used to screen multi-compartment scenarios based onthe size of the exposing and exposed compartments.23 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STAN DARD 805ii. Explain how the methods described in Chapter 2 of NUREG-1805, "Fire DynamicsTools (FDTs): Quantitative Fire Hazard Analysis Methods for the U.S. NuclearRegulatory Commission Fire Protection Inspection Program," December 2004(ADAMS Accession No. ML043290075) were used in the calculations to screen anignition source based on insufficient HRR to generate a HGL condition in theexposing compartment. In addition, clarify which FDTs were used for the HGLcalculations.iii. In the MCA scenario analysis, explain the technical basis of modeling the ZOI as avertical cylinder with the radius equal to 0.2 times the ceiling height in scenarioswhere the fire occurs near the opening between the two compartments and damagesitems on both sides within its ZOI.iv. Some of the EDT calculations make the following assumption: "It is assumed that theforced ventilation of air flow rate is distributed among the interconnectedcompartments, especially corridors, based on the volume of the compartments."Provide technical justification for this assumption.v. The screening process based on the ZOl specifies that if there are cable trays,conduits, or targets on the exposed side of the barrier within the ZOI, which may notbe the same as those inside the exposing compartment, the scenario should beanalyzed further. Provide details about this additional analysis.k. Specifically regarding the use of FDS in the CSR (physical analysis units (PA Us) 306 and302) calculations:i.It is stated that engineering judgment is used to assess that the delay in smokedetector activation, which is associated with cross-train logic that is not possible toincorporate in FDS, would be in the range of 2 to 10 seconds. Provide technicaljustification for this estimate.ii. The FDS "devices" (temperature and heat flux) were placed at different locationsaround the switchgear rooms. Describe the basis for choosing these locations.iii. The analysis highlights the location of possible electrical cabinet fires that wereconsidered. Provide technical justification for selecting these specific fire locations ordemonstrate that these locations lead to bounding or conservative estimates.iv. A number of transient fires were postulated in the CSRs, but the documentationindicates that the walkdown identified no transient combustibles and there were nostorage areas for more permanent combustibles in the fire areas, Provide justificationfor selection of the transient fire areas and indicate if this selection is dependent onany administrative controls of transient combustibles in the CSRs.v. The HRR used for the cabinet fires indicates that the cabinet doors were assumed tobe closed. Provide justification for this assumption (e.g., on the basis of the actualplant configuration or operational condition).vi. As stated in FM RAI 1.b, it is expected that secondary combustibles (ignition, flamespread, and cable tray fire propagation) would be part of the FDS analysis for theCSRs. Clarify how secondary combustibles were considered in the FDS analysis ofthe CSRs, and if they were not considered, provide justification for their omission.I. During the walkdown of the MOR, several observations were made, which requireadditional information:24 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805I.The main horseshoe and back panel cabinet configurations consist of open cabinetswith a steel mesh open top with the open sides facing each other across a narrowaisle. The EDS analysis utilizes an HRR case from Appendix G of NUREG-CR 6850,which assumes closed cabinets. Provide justification for not using an HRR caseapplicable to open cabinets or update the analysis with the appropriate HRR.ii. During the discussion about the open cabinets, it was also discussed that the currentanalysis does not consider the potential for fire spread across the aisle (i.e., within thehorseshoe) from the front to back or vice versa. Provide justification for notconsidering this potential fire spread or update the analysis to include this scenario.iii. During the walkdown of the MCR, several combustible items, which could beconsidered transient fire sources, were observed that could potentially have an/-HRRof greater than 317 kW. Examples include the kitchen area, the upholstered furniturein the shift manager's office and space below the shift manager's office, andphotocopiers. Provide additional information that can justify that the transient firesource selected in the FDS analysis is conservative and bounding.CCNPP RESPONSE FM RAI 01:Response to be provided 4/13/15.011.i -Response to be provided 3/11/1 5.FM RA! 02:The ASME/ANS Standard RA-Sa-20 09, "Standard for Level 1/Large Early Release FrequencyProbabilistic Risk Assessment for Nuclear Power Plant Applications," Part 4, requires damagethresholds be established to support the FPRA. Thermal impact(s) must be considered indetermining the potential for thermal damage of SSCs and appropriate temperature and criticalheat flux criteria must be used in the analysis.a. Describe how the installed cabling in the power block was characterized, specifically withregard to the critical damage threshold temperatures and critical heat fluxes for the rmosetand thermoplastic cables as described in NUREG/CR-6850. If thermoplastic cables arepresent, explain how raceways with a mixture of thermoset and thermoplastic cables weretreated in terms of damage thresholds.b. Explain how the damage thresholds for non-cable components (i.e., pumps, valves,electrical cabinets, etc.) were determined. Identify any non-cable components that wereassigned damage thresholds different from those for thermoset and thermoplastic cables,and provide a technical justification for these damage thresholds.c. Explain how exposed temperature-sensitive equipment was treated, and provide atechnical justification for the damage criteria that were used.CCNPP RESPONSE FM RA! 02:02a -Response to be provided 4/13/15.02b -Response to be provided 3/11/15.02c -Response to be provided 4/13/15.25 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805FM RAI 03:Section 2. 7.3.2 of NFPA 805 states that each calculational model or numerical method usedshall be *verified and validated through comparison to test results or comparison to otheracceptable models.Section 4.5.1.2 of the LAR states that fire modeling was performed as part of the FPRAdevelopment (NFPA 805, Section 4.2.4.2). Reference is made to Attachment J for a discussionof the verification and validation (V& V) of the fire models that were used. Furthermore,Section 4.7.3 of the LAR states that "Calculation models and numerical methods used insupport of complance with 10 CFR 50.48(c) were verified and validated as required bySection 2.7.3.2 of NFPA 805."For any tool or method identified in the response to FM RAI 1.a above, provide the V& V basis ifnot already explicitly provided in the LAR (for example, in Attachment J). Provide technicaldetails to demonstrate that these models were applied within the validated range of inputparameters, or justify the app lication of the model outside the validated range in the V& V basisdocuments.CCNPP RESPONSE FM RAI 03:Response to be provided 3/11/15.FM RAI104:Section 2.7.3.3 of NFPA 805 states that acceptable engineering methods and numerical modelsshall only be used for applications to the extent these methods have been subject toverifications and validation. These engineering methods shall only be applied within the scope,limitations, and assumptions prescribed for that method.Section 4.7.3 of the LAR states that, "Engineering methods and numerical models used insupport of compliance with 10 CFR 50.48(c) were applied appropriately as required bySection 2.7.3.3 of NFPA 805."Regarding the limitations of use, the NRC staff notes that algebraic models cannot be usedoutside the range of conditions covered by the experiments on which the model is based.NUREG-1805 includes a section on assumptions and limitations that provides guidance to theuser in terms of proper and improper use for each FDT.Identify uses, if any, of FDS and the FDTs outside the limits of applicabilty of the model, and forthose cases, explain how the use of FDS and the FDTs was justified.CCNPP RESPONSE FM RAI 04:Response to be provided 4/13/15.FM RAI 05:Section 4.5. 1.2 of the MAR states that fire modeling was performed as part of the FPRAdevelopment (NFPA 805, Section 4.2.4.2). The NRC staff notes this requires that quailified firemodeling and PRA personnel work together. Furthermore, Section 4. 7.3 of the LAB states thefollowing:26 A1TACHM ENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FiREPROTECTION ASSOCIATION STANDARD 805Cognizant 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.For personnel performing fire modeling for FPRA development and evaluation, CCNPP[Calvert Cliffs Nuclear Power Plant] develops and maintains quailification requirements forindividuals assigned various tasks. Position specific guides were developed to identify anddocument required training and mentoring to ensure individuals are appropriately qualifiedper the requirements of NFPA 805, Section 2.7.3.4, to perform assigned work.Qualification cards provide evidence that Design Engineering and PRA personnel have theappropriate training and technical expertise to perform assigned work, including the use ofengineering analyses and numerical models.Qualification requirements are contained in procedure CNG-TR-1.01-1014 (Reference 6. 47).CCNPP will maintain qualification requirements for the performance of NFPA 805 relatedtasks. Position specific qualification cards identify and document required training andmentoring to ensure cognizant individuals are appropriately qualified to perform assignedwork per the requirements of NFPA 805, Section 2. 7.3.4.Regarding qualifications of users of engineering analyses and numerical models (i.e., firemodeling techniques):a. Describe the requirements to qualify personnel for performing fire modeling calculations inthe NFPA 805 transition.b. Describe the process for ensuring that fire modeling personnel have the appropriatequalifications not only before the transition, but also during and following the transition.c. When fire modeling is performed in support of the FPRA, describe how propercommunication between the fire modeling and FPRA personnel is ensured.CCNPP RESPONSE FM RAI 05:05a -Fire modeling calculations were performed by engineers who meet the qualificationrequirements of Section 2.7.3.4 of NEPA 805. The qualification process through December2014 followed the guidance of ACAD 98-004, "Guidelines for Training and Qualification ofEngineering Personnel," and the CENG procedure on "Conduct of Training." All thoseperforming Fire Modeling for the Fire PRA were qualified and their qualifications weredocumented in the CENG training database. This qualification includes basic fire modelingtechniques as well as Fire PRA techniques. The CENG PRA Engineering Supervisorreviewed experience and education for all fire modeling work. Those performing detailed firemodeling analysis using tools such as CFAST (Consolidated Model of Fire and SmokeTransport) or FDS (Fire Dynamics Simulator) were required to have the relevantqualifications and experience in fire modeling to perform the analysis.In the case of the initial fire modeling, the vendor provided the credentials of the firemodelers, which were reviewed and approved by Risk Management Supervision. Duringand following transition, the existing engineering staff will continue to be knowledgeable infire modeling techniques, including interpreting and maintaining the fire modeling database.If new fire modeling personnel are needed in the future, their credentials will also bereviewed and approved by Exelon supervision. Currently the Risk Management organizationhas transitioned to Exelon qualification processes which include the Fire PRA qualification.27 ATTACHM ENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805Engineering Supervisors have responsibility for "Verifying qualifications prior to assigningpersonnel to perform job performance requirement independently." This requires reviewingthe qualifications in the training server to verify the qualification is current.05b -Response to be provided 3/11/15.05c -Response to be provided 3/11/15.FM RAI 06:Section 4.7.3 of the LAR states that, "Uncertainty analyses were performed as required bySection 2.7.3.5 of NFPA 805 and the results were considered in the context of the application.This is of particular interest in fire modeling and FPRA development."Regarding the uncertainty analysis for fire modeling:a. Describe how the uncertainty associated with the fire mode! input parameters wasaccounted for in the fire modeling analyses.b. Describe how the "model" and "completeness" uncertainties were accounted for in the firemodeling analyses.CCNPP RESPONSE FM RAI 0]6:Response to be provided 4/13/15.PRA RAI 01 -Fire Event Facts and Observations:Section 2.4.3.3 of NFPA 805 states that the probabilistic safety assessment (also referred to asPRA) approach, methods, and data shall be acceptable to the authority having jurisdiction,which is the NRC. RG 1.2 05 identifies NUREG/CR-6850 as documenting a methodology forconducting an FPRA and endorses, with exceptions and clarifications, NE! 04-02, Revision 2, asproviding methods acceptable to the NRC staff for adopting a fire protection program consistentwith NEFA 805. RG 1.200 describes a peer review process utilizing an associated ASME/ANSstandard (currently ASME/ANS-RA-Sa-2009) as one acceptable approach for determining thetechnical adequacy of the PRA, once acceptable consensus approaches or models have beenestablished for evaluations that could influence the regulatory decision. The primary result of apeer review are the facts and observations (F& Os) recorded by the peer review and thesubsequent resolution of these F&Os.Clarify the following dispositions to fire F&Os and Supporting Requirement (SR) assessmentsidentified in Attachment V of the LAR that have the potential to impact the FPRA results and donot appear to be fully resolved:a) PRM-83-01: The disposition to F&O PRM-83-01 appears to indicate that the EPRA wasupdated to address events involving a fire induced loss of MCR HVAC, which the peerreview suggests has a conditional core damage probability (CCDP) of 1.0, by increasingthe likelhood of functional failures in lieu of assuming their occurrence. Justify thefunctional failures modeled by the FPRA to address this loss of MCR HVAC. In addition,explain how the FPRA evaluates the degradation of equipment due to elevatedtemperatures caused by loss of HVAC as an increase in equipment failure rates, andprovide a technical basis for doing so.28 A1TACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805b) FSS-A5-01: This F&O states that some PAUs are further divided into "sub-PA Us" andappears to indicate that there is no explicit process for evaluating the fire spread acrosssub-PA U boundaries, which, as the peer review noted, are not defined by physicalbarriers. The disposition to this F&O, however, does not discuss such a process, and byreferencing a sensitivity analysis limited to a number of "representative" PAUs, suggeststhat this apparent deviation from acceptable methods has not been fully addressed for allPAUs for which sub-PA Us have been defined. Explain how the fire effects across non-physical sub-PA U boundaries are identified and evaluated. Discuss how this approach isconsistent with or coniservatively bounds acceptable methods.c) FSS-G4-01: The disposition to this F&O indicates that the MCA did not postulate apropagation scenario if doing so would require failure of a penetration seal. Thelicensee's analysis (CO-FSS-08) suggests that a similar approach may have also beenfollowed for other barrier types (e.g., walls). As a result, identify each barrier type forwhich propagation scenarios were not postulated, and provide quantitative justification(e.g., an evaluation demonstrating that MCA scenarios involving barrier failure are lowrisk, even considering the risk associated with the multi-compartment fire) for notaddressing propagation. As an alternative, provide updated risk results as part of theintegrated analysis requested in PRA RAI 03, summing the generic barrier failureprobabilities for each type of barrier present between communicating compartments,consistent with NUREG/CR-6850.d) FSS-G5-01: The disposition to this F&O indicates that unreliability values were applied toall normally open, self-closing dampers and doors; however, the disposition neitherprovides a basis for the values applied nor mentions active elements discussed elsewhere(e.g., water curtains in F&O PP-B5-0 1). Summarize the types of active fire barrierelements credited in the EPRA, and provide quantitative justification for their unreliabilityand unavailability.e) HRA-B2-01: The disposition to this F&O indicates that "adverse" operator actions, whichinclude actions to de-energize electrical busses as a means to address spuriousoperations, are modeled in the FPRA by assuming all equipment disabled by the action isfailed (i.e., the action is successful). Although the licensee's analysis (Section 2.2 ofCO-HRA -001) indicates that this assumption is conservative, the basis for this conclusionis unclear if the action is taken to reduce risk. In light of this:i. Provide justification for the assumption that modeling "adverse" actions as successfulis conservative. Note that guidance in NUREG-1921 offers considerations forevaluating fault clearing strategies in the FPRA human reliability analysis (HRA).ii. Clarify how "adverse" actions are addressed by the FPRA HRA dependency analysis,given that these actions are modeled by failing associated equipment directly within thePRA logic model.iii. Explain the statement in Attachment G that "[nione of the recovery actions were foundto have an adverse impact on the EPRA." In doing so, clarify how "adverse" riskimpact was defined. Note that FAQ 07-0030 states that "[i]f activities (recovery actionsor other actions in the post-fire operational guidance) are determined to have anadverse risk impact, they should be resolved during NFPA 80.5 implementation via analternate strategy that eliminates the need for the action in the NSCA. "f) CS-B1-01: The licensee's analysis (Appendix F of ECP-13-O00321, "Common PowerSupply and Common Enclosure Study') identifies several MCC 208/120 Volts alternating29 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805current load breakers that were not coordinated with their respective feed breakers. Thedisposition to this F&O indicates that these 120V panel breaker coordination issues are tobe addressed by plant modification; however, Attachment S does not appear to containsuch a modification. Identify the Attachment S modification(s) being credited to resolvethe 120V panel breaker coordination issues identified in the disposition to this F&O.CCNPP RESPONSE PRA RAI 01:01a -Modeling of the loss of control room HVAC impact was developed for internal events priorto the NFPA 805 application. The Calvert Cliffs control room and cable spreading roomsshare the common control room HVAC system. A GOTHIC thermal hydraulic analysis of theheat-up following a loss of control room HVAC determined the maximum temperature in therooms is 1 12&deg;F. The design temperature limit is 104&deg;F and the normal control roomtemperature is considered to b~e 720F. Given the maximum temperatures are slightly abovethe design temperature, the failure rates of the equipment were increased rather thanassuming a complete failure.The basis for the failure rate increases is IEEE 500, "Guide to the Collection andPresentation of Electrical, Electronic, Sensing Component, and Mechanical EquipmentReliability Data for Nuclear-Power Generating Stations." IEEE 500 listed a range of potentialfailure rate increases for various equipment types. The maximum recommend increase wasselected from among all equipment types in the control room and cable spreading rooms.Since all of the control room and cable spreading room controls and instrumentation aresupplied from 125VDC and/or 120VAC buses, the failure likelihood of these power supplieswas increased by the maximum recommended increase over the mission time of theventilation loss.01ib -Response to be provided 4/1 3/1 5.01c -Response to be provided 4/13/15.O1d -Normally open fire dampers are not considered active fire barriers in the CCNPP FPRA.The CCNPP FPRA does not model any normally open doors with closures which initiate doorclosure due to fire.The following failure probabilities were used for fire barriers that were credited in the CCNPPFPRA:For fire dampers the probability of failure 2.70E-03 was used based on the suggestedvalues in NUREG/CR-6850 Table 11-3.SFor doors, including watertight doors, the probability of failure 7.40E-03 was used basedon the suggested values in NUREG/CR-6850 Table 11-3. It may be noted that thewatertight doors included in the scenarios are used on a regular basis to enter and exitthe compartments. Also, the probability values used is assumed to include thepossibility of finding a door being propped open given a fire in the exposingcompartment.* For credited, installed water curtains, the fire barrier failure probability was based on anon-suppression probability (NSP) derived per the guidance in NUREG/CR-685030 AI-FACHM ENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805Appendix P, as suggested in NUREG/CR-6850 Section 11.5.4.4. Installed watercurtains were credited in the CCNPP FPRA for the following potential multi-compartmentfires:o Fires that could potentially propagate from PAU 101 (Unit 2 ECCS Pump Room) toPAU 1 20E (Unit 2 Containment Recirculation Pipe Tunnel -East)o Fires that could potentially propagate from PAU 102 (Unit 2 ECCS Pump. Room) toPAU 120W (Unit 2 Containment Recirculation Pipe Tunnel -West)o Fires that could propagate from PAU 118 (Unit 1 #12 EGOS Pump Room) to PAU122E (Unit 1 Containment Recirculation Pipe Tunnel -East)o Fires that could propagate from PAU 119 (Unit 1 #11 ECCS Pump Room) to PAU122E (Unit 1 Containment Recirculation Pipe Tunnel -East)Plant specific failure data was not collected to verify these probability values; however,procedures are implemented that ensure integrity of the active fire barriers that are creditedin the CCNPP fire protection program, including the active fire barriers that have beencredited in the CCNPP FPRA. The active fire barriers that have been credited in the CCNPPEPRA are, therefore, inspected periodically per plant procedures. Also, the active firebarriers that have been credited in the CCNPP FPRA that are relied upon to maintain safety /separation have established compensatory measures that are put in place whenever anissue is discovered with the credited feature.01e -i.There are both positive and negative aspects to "adverse" operator actions. On thepositive side, a successful "adverse" operator action will preclude a spurious actuationthat could otherwise have negative consequences. On the negative side, an "adverse"operator action disables equipment that may be credited in the CCNPP FPRA. Whenthe "adverse" operator action is assumed to be successful, the negative impact mustalso be assumed (i.e., the associated mitigation function credited in the CCNPP FPRAmodel must be assumed to be failed). When an "adverse" operator action is credited, adetailed human reliability analysis (HRA) quantification is required. This was themethodology employed in the CCNPP FPRA HRA. As an example, when the CCNPPFPRA HRA credited a procedure to de-energize a valve such as a PORV to prevent aspurious opening which can lead to a loss of coolant accident (LOCA) scenario, suchcredit was only applied following a detailed HRA quantification, it also always assumedthe negative impact that the valve was not available to energize to support a feed andbleed type function (the negative impact is incorporated by, setting the associated basicevent, failure to open the pilot operated relief valve (PORV) in this case, to TRUE in thequantification, such that the risk of core damage or large early release frequency (LERF)increases).i.The CCNPP FPRA will model all equipment disabled by the "adverse' action as beingfailed (i.e., the action is successful and, as a consequence, equipment is disabled);therefore, this is not evaluated in the dependency evaluation. If an "adverse" action iscredited to prevent spurious actuation, then a detailed human reliability analysis (HRA) isdeveloped. All of the detailed HRAs are evaluated in the dependency evaluation.31 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805iii. A recovery action is considered to have an adverse impact on the CONPP FPRA if itcauses more than a negligible increase in core damage frequency (CDF) or large, earlyrelease frequency (LERF). Through implementation of Item 15 in LAR Table S-3,CCNPP will ensure that there is not an adverse impact on the CON PP FPRA.Implementation Item 15 in LAR Attachment S table S-3 includes updating the AbnormalOperating Procedures (AOPs) for severe fires for the recovery actions evaluated. Aspart of this effort, actions are either functionally based actions or direct procedure stepsthat are always required. Functional actions are only implemented as required tosupport a functional loss. For example, if 4kV Bus 21 cannot be re-powered, then action11 I0PEN4KVBKRS in Table G-1 would be implemented. In other cases, the actions areonly conditional on some fire-related condition such as the MCR being abandoned. Thenegative aspects of these actions will always be assumed to occur when the fire-relatedcondition occurs. Although some of the directly implemented actions may have bothnegative and positive impacts, overall recovery actions are only credited when they arejudged to be risk beneficial. In support of Implementation Item 15, the actions that areultimately credited in the AOPs for severe fires will be risk beneficial or will include thequantification of the adverse impact of the actions in the FPRA quantification.All risk adverse actions will be removed from the AOPs for severe fires except thoserequired due to operational concerns. Risk adverse actions that are required due tooperational concerns will be appropriately modeled in the CON PP FPRA.01f -While Appendix F of ECP-13-000321, "Common Power Supply and Common EnclosureStudy," does state that several MOO 208/120 VAC load circuit breakers were not coordinatedwith their respective feed circuit breakers; a later, more detailed, analysis documented inECP-1 3-000776 concluded that those same power supplies did not have any coordinationissues. A plant modification is not, therefore, required.PRA RAi 02 -Internal Event F&Os: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 an FPRA and endorses, with exceptions and clarifications, NEI 04-02,Re vision 2, as providing methods acceptable to the staff for adopting a fire protection programconsistent with NFPA 805. RG 1.200 describes a peer review process utilizing an associatedASME/ANS standard (currently ASME/ANS-RA-Sa-2009) as one acceptable approach fordetermining the technical adequacy of the PRA once acceptable consensus approaches ormodels have been established. The primary results of a peer revieW are the F&Os recorded bythe peer review and the subsequent resolution of these F&Os.Clarify the following dispositions to internal events F&Os and SR assessments identified inAttachment U of the LAR that have the potential to impact the EPRA results and do not appearto be fully resolved:a) 4-5: This F&O indicates that the alignment strategy assumed by the PRA for the OC dieselgenerator (DG) is not appropriately justified and may be non-consen'ative. While thedisposition to this F&O clarifies how alignment of the CC DG is modeled in the PRA, ajustification for this treatment is not provided. Provide a technical and/or procedural basisfor the alignment strategy assumed in the PRA for the CC DG, and indicate whether anyoperator interviews were conducted to support the analysis.32 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE_________PROTECTION ASSOCIATION STANDARD 805b) 6-23: This F&O indicates that some joint human error probabilities (HEPs) applied withinthe internal events PRA (IEPRA) may not accurately reflect the sequential timing ofassociated operator actions. While the disposition appears to address the specificexample referenced by the F&O, it is not clear that the broader issue has been fullyresolved in the fire PRA, particularly noting that the status of this F&O in Table U-i isidentified as "open."i. Explain how the HRA methods used by the EPRA for developing HEP and joint HEPvalues are consistent with or conservatively bound NRC-accepted guidance inNUREG/CR-6850 or NUREG- 1921. Alternatively, provide updated risk results as partof the aggregate change-in-risk analysis requested in PRAN RAI 03 applying HEPand joint HEP values developed using NRC-accepted guidance.ii. NUREG-1921 indicates and NUREG-1792 (Table 2-1) states that joint HEP valuesshould not be below 1. OE-05. Confirm that each joint HEP value used in the FPRAbelow 1.0E-05 includes its own justification that demonstrates the inapplicability of theNUREG- 1792 lower value guideline. Provide an estimate of the number of these jointHEPs below 1.0E-05 and at least two different types of justification.CCNPP RESPONSE PRA RAI 02:a) The alignment of the 0C Diesel Generator (DG) is not fixed. in the AOPs for severe fires, the00 DG is aligned to both 4kV Buses 11 and 24. Depending on the circumstance, Operationspersonnel will align the 00 DG to the location where the most equipment can be restored.4kV Buses 11 and 24 support a motor driven Auxiliary Feedwater (AFW) pump and MCRheating, ventilation and air conditioning (HVAC). 4kV Buses 14 and 21 support a similar setof loads. As such, in many circumstances those buses would be the logical choice. There isnothing, however, to prevent a re-alignment if the equipment being powered from the 00 DGfails or is not otherwise satisfying operational needs. To prevent excessive modelcomplexity, a fixed alignment strategy is used which is conservative compared to the realityof the flexible alignment strategy.Additional operator interviews regarding alignment of the 00 DG will be conducted.Summaries of these interviews will be added to the CCNPP FPRA documentation. TheCCNPP FPRA documentation will also be revised to capture simulator observationsregarding alignment of the 00 DG. As revisions are drafted for the AOPs for severe fires, theimpacts on the 0C DG alignment modeling will be considered to ensure appropriate modelingfor the expected post transition configuration. These changes will also be discussed duringinterviews with operators to ensure realistic modeling is performed.The revised documentation containing the additional operator interviews described above willbe will be generated in conjunction with the update of CCNPP FPRA analysis documentationsupporting RAI PRA-03.b)i. As noted in Attachment U of the LAR, human failure event (HFE) timelines were reviewed.Some events in the CCNPP FPRA human reliability assessment (HRA) were split intomultiple HFEs, where appropriate, to account for the different scenarios and to ensure thatthe sequential timing of the associated operator actions is appropriate. The CCNPP FPRAHFEs were then assessed as part of the dependency analysis using the EPRI's HRA33 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805Calculator. This is considered an acceptable HRA method for developing HEP and joint HEPvalues.To confirm the appropriateness of the review mentioned in Attachment U of the LAR anadditional step was taken: A review all CCNPP FPRA HEPs with a Tdeiay of zero wasundertaken. Tdelay represents the period between the start time of an event and the time theoperator acknowledges the cue. All HEPs with a Tdelay of zero were reviewed to determine ifthese HEPs were discussed during interviews with CCNPP Operations personnel. Operatorinterviews provide valuable insights and understanding of plant-specific crew responses andfeasibility. It was determined that all HEPs with a Tdelay of zero were discussed duringoperator interviews conducted either during the Internal Events PRA's Peer Review or theFPRA's Peer Review, or as part of the original Internal Events PRA's HRA. it should benoted that, while all HEPs were discussed during these interviews, only a very small fractionof the HEPs were found to have a Tdelayof zero that might not be appropriate.Changes to the Tdalay could increase or decrease the risk. As the number of potentiallyaffected HEPs in the CCNPP FPRA is small as compared to the total number of HEPs in theCCNPP FPRA, any such changes to the Tdelay would likely have a negligible impact on theCCNPP FPRA results.b)ii. None of the individual HEPs included in the CCNPP FPRA have a value less than1 .OOE-05. The CCNPP FPRA evaluated 2700 joint human error probabilities (JHEPs); 2259of these 2700 JHEPs are below 1 .O0E-05.The documentation provided in the EPRI HRA Calculator justifies each unique HEP value,including the unique values used for all JHEPs. The evaluations of the individual HEPsinclude common cognitive actions where appropriate. JHEPs are only developed when thereis not a common cognitive failure mode.Justification examples of JHEPs CombinationU1_1770 and CombinationU1_1065 aredescribed below:CombinationU1_1770- 4.23E-090FW00HFMPZ9-FR -1 .9E-02 -Operations fails to control Main Feedwater flow post-tripto prevent Steam Generator overfill given fire-induced plant triplrn~act: SG overfill fails the running turbine driven Auxiliary Feedwater pumpLocation: Main Control Room*CSTOHF-DEPLETION-FR -6.00E-05 -Operations fails to detect Condensate StorageTank level dropping during fire (common cognitive)Impact: Failure of AFW pumpsLocation: MCR*CVCOHFOTA8HRS-FR -7.1E-03 -OTCC -All AFW / MFW failed after CST depletionduring fireImpact: Failure of once through core cooling and all AFW start actionsLocation: MOR34 ATrACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805All of the actions in combination U1_1770 are treated independently largely due to timeseparation. Operations fails to secure the MFW pumps prior to a SG overfill (FW00HFMPZ9-FR) this fails the running TO AFW pump. The minimum time to CST depletion is six hoursfollowing the trip. Although this is the time used in the HRA evaluation with a TO AFW pumpfailed, this would be longer. Due to this large time separation and different cues (SG waterlevel and CST water level), these actions are considered independent. If Operations fails torecognize that the condensate storage tank is depleting (CSTOHF-DEPLETION-), then all therunning AFW pumps fail. Operations could start the other unit's AFW pump or align forOTOC (CVCOHFOTA8HRS-FR). There is at least two hours available to align following aCST depletion AFW failure. A CST depletion failure in combination with an AFW failurewould provide even more time. Due to the large time separation and different cues (SGwater level and CST water level), this is also considered to be independent.CombinationUl_1065- 1.38E-07*AFW0HFCCSGDEC-FR -1.2E-03 -Operations fails to diagnose SG level decreasingduring fire (common cognitive)Impact: Failure of OTOC and AFW start actionsLocation: MCR*CA00HFN2C8-FRI -7.80E-04 -Operations fails to start both Salt Water AirCompressors (SWACs), in Emergency Operating Procedure (EOP) 8, no loss of airannunciators, dual unit trip during fireImpact: Loss of air to AFW flow control valves. AFW delivers full flow to the SGs.Location: MCR*AFWOHFHXB-FR -5.6E-3 -Operators fail to control AFW flow during fire, no CR flowsupport, EOP-8Impact: SG overfill fails the running TD AFW pumpLocation: AFW Pump RoomAction AFW0HFHXB-FR and AFW0HFCCSGDEC-FR are considered independent whileactions CA00HFN2C8-FRI and AFWOHFHXB-FR are considered a medium dependency (perTHERP, maximum credit for a medium action is 0.14). Operations fail to start both SWACsand provide long term air for AFW before the AFW accumulators deplete (CA00HFN2C8-FR I). Once air is lost, Operations has a limited time to prevent overfill (AFWOHFHXB-FR); assuch, this action has a medium dependency with CAO0HFN2C8-FRI. Following a late overfillevent caused by AFW accumulator depletion, Operations has over two hours to re-establishAFW flow. Due to the large time separation, these actions are considered independent.It should be noted that for these joint events to occur, not only does the whole Operationscrew need to fail, but the whole emergency response organization must fail as well. Foractions to be independent there needs to be at least 60 minutes of time separation betweenthe cues for the Operations actions as well as no common cognitive function. Any fire thatprogresses to core damage must affect multiple redundant groups of safety relatedequipment. The Calvert Emergency Action Level procedure requires us to declare an "Alert"when:FIRE or EXPLOSION resulting in EITHER:VISIBLE DAMAGE to ANY SAFETY-RELATED35 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805STRUCTURE, SYSTEM, OR COMPONENT within ANYTable H-i areaORControl Room indication of degraded performance of ANYSAFETY-RELATED STRUCTURE, SYSTEM, ORCOMPONENT within ANY Table H-i areaTable H-i areas include: MCR, Containment, Auxiliary Building, Diesel Generator Rooms,Intake Structure, 1A/OC Diesel Generator Buildings, Refueling Water Tank (RWT), RWTRooms, CST No. 12, Fuel Oil Storage Tank (FOST) No. 2 and AFW Pump Rooms.As a fire that progresses to core damage always affects safety related equipment in theseareas, a core damage or large early release fire will always result in the activation of theemergency response organization. The Technical Support Center is required to beoperational within 60 minutes of activation.PRA RAI 03 -Inteqrated Analysis: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.RG 1.174 provides quantitative guidelfines on ODE and LERF, identifies acceptable changes tothese frequencies that result from proposed changes to the plant's licensing basis, anddescribes a general framework to determine the acceptability of risk-informed changes. TheNRC staff's review of the information in the LAR has identified additional information that isrequired to fully characterize the risk estimates.The PRA methods currently under review in the LAB include:* PRA RAI 01.a regarding loss of MOR HVAC* PRA RAI 01.b regarding division of PAUs into "sub-PA Use'*PRA RA! 01.c regarding treatment of propagation in the MOA* PRA RA! 01. d regarding unreliability and unavailability of active barriers* PRA RAI 01.e regarding adverse operator actions* PRA RAI 01. f regarding 120V panel breaker coordination issues* PRA RAI 02.a regarding alignment of 00 diesel generator* PRA BAI 02.b regarding HRA methods, including sequential timing of operator actions* PRA RAI 04 regarding placement of transient fires* PRA RAI 05 regarding transient influence factors* PRA RA! 06 regarding reduced transient HRR* PRA BA! 07 regarding self-ignited cable fires and those caused by welding and cutting* PRA RAI 08 regarding treatment of junction boxes* PRA RAI 09 regarding treatment of sensitive electronicso PRA BA! 10 regarding circuit failure probabilities* PRA BA! 11 regarding counting and treatment of Bin 15 electrical cabinets* PRA BAI 12 regarding treatment of HEAF* PRA BAI 13 regarding MOB modeling* PRA BAI 14 regarding credit for MOR abandonment actions* PRA BA! 15 regarding MOB abandonment on loss of control36 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805_____* PRA RAI 16 regarding application of the state-of-knowledge correlation (SOKC)* PRA RA! 18 regarding ,4ICDF, ALERF and additional risk of RAs* PRA RA!123 regarding other deviations from acceptable methodsProvide the following:a) Results of an aggregate analysis that provide the integrated impact on the fire risk(i.e., the total transition CDF, LERF, ,4CDF, ALERF, and additional risk of RAs) ofreplacing specific methods identified above with alternative methods that are acceptableto the NRC. In this aggregate analysis, for those cases where the individual issues havea synergistic impact on the results, a simultaneous analysis must be performed. For*those cases where no synergy exists, a one-at-a-time analysis may be done. For thosecases that have a negligible impact, a qualitative evaluation may be done. It should benoted that this list may change depending on NRC's review of the responses to otherRAIs in this document.b) For each method (i.e., each bullet) above, explain how the issue will be addressed in1) the final aggregate analysis results provided in support of the LAR, and 2) the PRA thatwill be used at the beginning of the self-approval of post-transition changes. In addition,provide a process to ensure that all changes will be made, that a focused-scope peerreview will be performed changes that are PRA upgrades as defined in the PRAstandard, and that any findings will be resolved before self-approval of post-transitionchanges.c) in the response, explain how RG 1.205 risk acceptance guidelines are satisfied for theaggregate analysis. Additionally, discuss the likelihood that the risk increase in anyindividual fire area would exceed the acceptance guidelines, and if so, why exceeding theguidelines should be acceptable. If applicable, include a description of any newmodifications or operator actions being credited to reduce delta risk as well as adiscussion of the associated impacts to the fire protection program.d) If any unacceptable methods identified above will be retained in the PRA and will be usedto estimate the change in risk of post-transition changes to support self-approval, explainhow the quantification results for each future change will account for the use of thesemethods.CCNPP RESPONSE PRA RAI 03:Response to be provided 4/13/15.PRA RAl 04 -Transient Fire Placement at Pinch Points: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 an FPRA and endorses, with exceptions and clarifications, NE/ 04-02,Revision 2, as providing methods acceptable to the staff for adopting a fire protection programconsistent with NFPA 805. Methods that have not been determined to be acceptable by theNRC staff, or acceptable methods that appear to have been applied differently than described,require additional justification to allow the NRC staff to complete its review of the proposedmethod.The NRC staff could not identify in the LAR or licensee's analysis a description of how "pinchpoints" for transient fires were treated in the FPRA. Per NUREG/CR-6850, Section 11.5.1.6,37 A'1-ACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805transient fires should, at a minimum, be placed in locations within the plant PAUs where CCDPsare highest for that PAU (i.e., at "poinch points"). Pinch points include locations of redundanttrains or the vicinity of other potentially risk-relevant equipment. Cable congestion is typical forareas like the CSR, so placement of transient fire at pinch points in those locations is important.Hot work should be assumed to occur in locations where hot work is possible, even ifimprobable, keeping in mind the same philosophy.a) Clarify how "pinch points" were identified and modeled for general transient fires andtransient fires due to hot work.b) Describe how general transient fires and transient fires due to hot work are distributedwithin the PAUs at Calvert Cliffs. In particular, identify the criteria used to determinewhere such ignition sources are placed within the PA Us.CCNPP RESPONSE PRA RAI 04:Response to be provided 4/13/15.PRA RAI 05 -Transient influencinq Factors:Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall beacceptable to the NRC. The 1.205 identifies NUREG/CR-6850 as documenting a methodologyfor conducting an FPRA and endorses, with exceptions and clarifications, NEI 04-02,Revision 2, as providing methods acceptable to the NRC staff for adopting a fire protectionprogram consistent with NFPA 805. Methods that have not been determined to be acceptableby the NRC staff, or acceptable methods that appear to have been applied differently thandescribed, require additional justification to allowv the NRC staff to complete its review of theproposed method.Appendix H of the LAR does not indicate that FAQ 12-0064, "Hot Work/Transient FireFrequency influence Factors," dated January 17, 2013 (ADAMS Accession No. ML 12346A488),was used in preparation of the EPRA. According to this FAQ, transient influence factor may notbe assigned a ranking value of 0, unless associated activities and/or entrance during poweroperation are precluded by design and/or operation. The licensee's analysis (Table C-2 ofCO-/GN-O01) indicates, however, that a large number of PAUs are assigned ranking values of 0for one or more of the transient influence factors. As a result, clarify whether ranking valuesassigned to transient influencing factors were developed consistent with the guidance inNUREG/CR-6850 and FAQ 12-0064, in particular Section 6.5.7.2, and if not, providejustification. if justification cannot be provided, then provide treatment of transient influencefactors consistent with NRC guidance in the integrated analysis provided in response to PRARA/ 03.CCNPP RESPONSE PRA RAI 05:Response to be provided 4/13/15.PRA RAI 06 -Reduced Transient Heat Release Rates: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 an FPRA and endorses, with exceptions and clarifications, NE/ 04-02,Revision 2, as providing methods acceptable to the NRC staff for adopting a fire protectionprogram consistent with NFPA 805. Methods that have not been determined to be acceptable38 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805by the NRC staff, or acceptable methods that appear to have been applied differently thandescribed require additional justification, to allow the NRC staff to complete its review of theproposed method.It appears that reductions below the NUREG/CR-6850 98th percentile HRR of 317 kilowatt (kW)for transient fires may have been credited in the FPRA. In particular, the licensee's analysis(e.g., Section 6.5.4 of Addendum 1 to CO-FSS-O04) indicates that a 142 kW (75th percentile)HRR transient fire was postulated in the switchgear rooms. As a result, discuss the key factorsused to justify any reduced HRR below 317 kW, per the guidance endorsed by the June 21,2012, memo from Joseph Glitter to Biff Bradley, "Recent Fire PRA Methods Review PanelDecisions and EPRI 1022993, 'Evaluation of Peak Heat Release Rates in Electrical CabinetFires"' (ADAMS Accession No. ML 12171A583). In doing so:a) Identify all PA~ls for which a reduction in the HRR below 317 kW for transient fires iscredited.b) For each location where a reduced HRR is credited, describe the administrative controlsthat justify the reduced HRR, including how location-specific attributes and considerationsare addressed.c) Provide the results of a review of records related to violations of transient combustible andhot work controls, including how this review informs the development of administrativecontrols credited, in part, to justify an HRR lower than 317 kW.CCNPP RESPONSE PRA RAI 06:Response to be provided 4/113/15.PRA RA! 07- Self-Igqnited and Caused by Welding and Cuttinu,: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 an FPRA and endorses, with exceptions and clarifications, NEI 04-02,Revision 2, as providing methods acceptable to the NRC staff for adopting a fire protectionprogram consistent with NFPA 805. In letter dated July 12, 2006, to NEI (ADAMS AccessionNo. ML061660105), the NRC established the ongoing FAQ process where official agencypositions regarding acceptable methods can be documented until they can be included inrevisions to RG 1.205 or NEI 04-02.Appendix H of the LAR does not indicate that FAQ 13-0005, "Cable Fires Special Cases: Self-Ignited and Caused by Welding and Cutting," dated June 26, 2013 (ADAMS Accession No.ML 13322B260), was used in preparation of the FPRA. Explain whether the treatment of self-ignited fires and fires caused by welding and cutting is consistent with FAQ 13-0005, and if not,provide justification. If justification cannot be provided, then provide treatment of self-ignitedfires and fires caused by welding and cutting consistent with NRC guidance in the integratedanalysis provided in response to PRA RAI 03.CCNPP RESPONSE PRA RAi 07:Response to be provided 3/11/15.39 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805PRA RAI 08 -Junction Boxes: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 an EPRA and endorses, with exceptions and clarifications, NE! 04-02,Revision 2, as providing methods acceptable to the NRC staff for adopting a fire protectionprogram consistent with NFPA 805. In letter dated July 12, 2006, to NEi from SunilWeerakkody (ADAMS Accession No. ML061660 105), 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.Appendix H of the LAR does not indicate that FAQ 13-0006, "Modeling Junction Box Scenariosin an Fire PRA," dated May 6, 2013 (ADAMS Accession No. ML13149A527), was used inpreparation of the FPRA. Explain whether the treatment of junction box fires is consistent withFAQ 13-0006, and if not, provide justification. If justification cannot be provided, then providetreatment of junction box fires consistent with NRC guidance in the integrated analysis providedin response to PRA RA! 03.CCNPP RESPONSE PRA RAI 08:Response to be provided 3/11/15.PRA RAI 09 -Sensitive Electronics: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 an FPRA and endorses, with exceptions and clarifications, NEI 04-02,Revision 2, as providing methods acceptable to the NRC staff for adopting a fire protectionprogram consistent with NFPA 805. Methods that have not been determined to be acceptableby the NRC staff, or acceptable methods that appear to have been applied differently thandescribed, require additional justification to allow the NRC staff to complete its review of theproposed method.The NRC staff could not identify in the LAR or licensee's analysis a description of how potentialfire damage to sensitive electronics was modeled. Though the treatment of sensitive electronicsmay be consistent with recent guidance on the modeling of sensitive electronics, Appendix H ofthe LAR does not cite FAQ 13-0004, "Clarifications Regarding Treatment of SensitiveElectronics," dated December 3, 2013 (ADAMS Accession No. ML13322A085), as one of theFAQ guidance documents used to support the FPRA. Describe the treatment of sensitiveelectronics for the FPRA and explain whether it is consistent with the guidance in FAQ 13-0004,including the caveats about configurations that can invalidate the approach (i.e., sensitiveelectronic mounted on the surface of cabinets and the presence of louvers or vents). If theapproach is not consistent with FAQ 13-0004, justify the approach, or replace the currentapproach with an acceptable approach in the integrated analysis performed in response to PRARAI 03.CCNPP RESPONSE PRA RAI 09:Response to be provided 4/13/15.40 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE______ PROTECTION ASSOCIATION STANDARD 805 ______PRA RAI 10- Conditional Probabilities of Spurious 0perations: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.RO 1.174 provides quantitative guidelines on CDF, LERF, and identifies acceptable changes tothese frequencies that result from proposed changes to the plant's licensing basis anddescribes a general framework to determine the acceptability of risk-informed changes. TheNRC staff's review of the in formation in the LAB has identified additional information that isrequired to fully characterize the risk estimates.Attachment V of the LAB indicates that application of circuit failure probabilities was limited tocircuits without control power transformers and further clarifies that the probabilities appliedyield conservative risk and delta risk estimates relative to the July 1, 2013, interim guidance(ADAMS Accession No. ML13165A214). However, new guidance on using conditionalprobabilities of spurious operation for control circuits was recently issued by the NRC inSection 7 of NUREG/CR-7150, Volume 2. This guidance included a) replacement of theconditional hot short probability tables in NUREG/CR-6850 for Option #1 with new circuit failureprobabilities for single brepk and double break control circuits, b) Option #2 in NUREG/CR-6850is not an adequate methodi and should not be used, c) replacement of the probability of spuriousoperation duration figure in FAQ 08-005 1 for AC control circuits, d) aggregate values for circuitfailure probabilities should be used unless it is demonstrated that a cable is only susceptible to asingle failure mode, e) incorporation of the uncertainty values for the circuit failure probabilitiesand spurious operation duration in the SOKC for developing the mean CDF!LERF, andf) recommendations on the hot short probabilities to use for other cable configurations, includingpanel wiring, trunk cables, and instrument cables. Provide an assessment of the assumptionsused in the Calvert Cliffs FPRA relative to the updated guidance in NUREG/CR-7 150,Volume 2, specifically addressing each of the above items. If the FPRA assumptions are notbounded by the new guidance, provide a justification for each difference, or provide updated riskresults as part of the aggregate change-in-risk analysis requested in PRA RAI 03, utilizing theguidance in NUREG/CR-7150.CCNPP RESPONSE PRA RA1 10:Response to be provided 4/13/15.PRA RAi 11 -Counting and Treatment of Bin 15 Electrical Cabinets:Section 2.4.3.3 of NEPA 805 states that the PRA approach, methods, and data shall beacceptable to the NRC. RO 1.205 identifies NUREG/CR-6850 as documenting a methodologyfor conducting an FPRA and endorses, with exceptions and clarifications, NE! 04-02,Re vision 2, as providing methods acceptable to the NRC staff for adopting a fire protectionprogram consistent with NFPA 805. Methods that have not been determined to be acceptableby the NRC staff, or acceptable methods that appear to have been applied differently thandescribed, require additional justification to allow the NRC staff to complete its review of theproposed method.The licensee's analysis (Section 2.2.1 of CO-FSS-O02) appears to indicate that the EPRAevaluates the potential for propagation of electrical cabinet fires based solely on the text inAppendix 0 (Section G.3.3) to NUREG/CR-6850; however, portions of this text were either41 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805clarified or disregarded in Chapter 8 of Supplement 1 of NUREG/CR-6850. In fight of thisobservation, address the following:a) Per Section 6.5.6 of NUREG/CR-6850, fires originating from within "well-sealed electricalcabinets that have robustly secured doors (and/or access panels) and that house onlycircuits below 440V'" do not meet the definition of potentially challenging fires and,therefore, should be excluded from the counting process for Bin 15. By counting thesecabinets as ignition sources within Bin 15, the frequencies applied to other cabinets areinappropriately reduced. Clarify that this guidance is being applied. If not, then addressthe impact as part of the integrated analysis performed in response to PRA RAI 03.b) Clarify if the criteria used to evaluate whether electrical cabinets below 440V are "wellsealed" are consistent with guidance in Chapter 8 of Supplement 1 of NUREG/CR-6850.If not, then address the impact as part of the integrated analysis performed in response toPRA RAI 03.c) All cabinets having circuits of 440V or greater should be counted for purposes of Bin 15frequency apportionment based on the guidance in Section 6.5.6 of NUREG/CR-6850.Clarify that this guidance is being applied. If not, then address the impact as part of theintegrated analysis performed in response to PRA RAi 03.d) Por those cabinets that house circuits of 440V or greater, propagation of fire outside theignition source should be evaluated based on guidance in Chapter 6 of NUREG/CR-6850,which states that "an arcing fault could compromise panel integrity (an arcing fault couldburn through the panel sides, but this should not be confused with the high energy arcingfault type fires)." Describe how fire propagation outside of cabinets greater than 440V isevaluated (including those that are considered "well-sealed'). If propagation is notevaluated, then address the impact as part of the integrated analysis performed inresponse to PRA RA! 03.CCNPP RESPONSE PRA RAI 11:11 a -Response to be provided 4/13/15.1 lb -Response to be provided 3/11/15.1 lc -Response to be provided 3/11/15.1lid -Response to be provided 3/11/15.PRA RAI 12- High,,Energyv Arcingq Faults: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 an FPRA and endorses, with exceptions and clarifications, NEI 04-02,Revision 2, as providing methods acceptable to the NRC staff for adopting a fire protectionprogram consistent with NFPA 805. Methods that have not been determined to be acceptableby the NRC staff, or acceptable methods that appear to have been applied differently thandescribed, require additional justification to allow the NRC staff to complete its review of theproposed method.The NRC staff could not identify in the LAR or licensee's analysis a description of how HEAFwere modeled. The licensee's analysis (e.g., Appendix B to CO-FO-CO1) appears to indicate42 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805that HEAF ignition sources are combined with other ignition sources (e.g., transients) to formfire scenarios. Per Appendix P of NUREG/CR-6850, however, HEAF events and other types offires have different non-suppression probability curves. In addition, the NRC staff'sinterpretation of the NUREG/CR-6850 guidance is that the growth of a fire subsequent to aHEAF event, unlike other types of fires, instantaneously starts at a non-zero HRR because ofthe intensity of the initial heat release from the HEAF:. As a result, provide a detailedjustification of the FPRA's treatment of HEAF events and the ensuing fire that includes adiscussion of conservatisms and non-conservatisms relative to the accepted methods andassesses the associated impacts on the fire total and delta risk results. Alternatively, replacethe current approach with an acceptable approach in the integrated analysis performed inresponse to PRA RAI 03. Note that the response should address the treatment of all HEAFscenarios, including in the HGL analysis and MCA.CCNPP RESPONSE PRA RAI 12:Response to be provided 4/13/15.PRA RAI 13 -MCR Modelinq: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.RG 1.174 provides quantitative guidelines on CDF, LERE, and identifies acceptable changes tothese frequencies that result from proposed changes to the plant's licensing basis anddescribes a general framework to determine the acceptability of risk-informed changes. TheNRC staff's review of the information in the LAR has identified additional information that isrequired to fully characterize the risk estimates.The licensee's analysis (Section 11.1 of CO-FSS-007) appears to assume that all of the wiringinside MCR control panels is qualified, even though unqualified wiring is known to be present aswell. Describe how the presence of both qualified and unqualified wiring is incorporated into theNUREG/CR-6850 Appendix L evaluation. Alternatively, provide treatment of qualification that isconsistent with or bounds the actual MCR configuration in the integrated analysis provided inresponse to PRA RAI 03.CCNPP RESPONSE PRA RAI 13:Response to be provided 4/13/15.PRA RAI 14 -Credit for MCR Abandonment Actions: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 NEPA 805 basedprogram, and all future plant changes to the program, shall be acceptable to the NRC.RG 1.174 provides quantitative guidelines on CDF, LERF, and identifies acceptable changesto. these frequencies that result from proposed changes to the plant's licensing basis anddescribes a general framework to determine the acceptability of risk-informed changes. TheNRC staff's review of the information in the LAR has identified additional information that isrequired to fully characterize the risk estimates.43 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE________PROTECTION ASSOCIATION STANDARD 805Tables W-2 through W-5 of the LAR and the licensee's analysis (Section 9.0 of CO-FSS-007)appear to represent MCR abandonment on loss of habitability as a single scenario with unitspecific CCDP and conditional large early release probability (CLERP) values. However, theNRC staff could not identify in the LAR or the licensee's analysis the method(s) used to obtainthese values./In light of this:a) Describe how MCR abandonment was modeled for loss of habitability in both the post-transition and the compliant plant. Include identification of the actions required to executesafe alternate shutdown and how they are modeled in the FPRA, including actions thatmust be performed before leaving the MCR. Also, include an explanation of how theCCDPs and CLERPs are estimated for fires that lead to MCR abandonment.b) Explain how the CCDPs and CLERPa estimated for fires that lead to abandonment due toloss of habitability address various possible fire-induced failures. Specifically, provide adiscussion of how the following scenarios are addressed:i. Scenarios where fire fails only a few functions aside from forcing MOB abandonmentand successful alternate shutdown is straightforward;ii. Scenarios where fire could cause some recoverable functional failures or spuriousoperations that complicate the shutdown, but successful alternate shutdown is likely;and,iii. Scenarios where the fire-induced failures cause great difficulty for shutdown by failingmultiple functions and/or complex spurious operations that make successfulshutdown unlikely.c) Explanation of the timing considerations (i.e., total time available, time until cues arereached, manipulation time, and time for decision-making) made to characterize scenariosin Part (b). Include in the explanation the basis for any assumptions made about timing.d) Discussion of how the probability associated with failure to transfer control to the AuxiliaryShutdown Panel is taken into account in Part (b).CGNPP RESPONSE PRA RAI 14:Response to be provided 4/13/15.PRA RAt 15-= MCR Abandonment on Loss of Control:Section 2.4.3.3 of NEPA 805 states that the PRA approach, methods, and data shall beacceptable to the NRC. Section 2.4.4.1 of NEPA 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.RG 1.174 provides quantitative guidelines on CDF, LERF, and identifies acceptable changes tothese frequencies that result from proposed changes to the plant's licensing basis anddescribes a general framework to determine the acceptability of risk-informed changes. TheNRC staff's review of the information in the LAR has identified additional information that isrequired to fully characterize the risk estimates.LAR Table G-1 identifies several PCS actions for non-MCR fire areas (Fire Areas 16 and 17),which encompass, in part, the Unit 1 and Unit 2 CSRs. Additionally, the licensee's analysis(Table 6 of CO-HRA -001) appears to credit actions to transfer control from the MCR to theauxiliary shutdown panel for fires in the CSR. In light of this:44 A1-TACHM ENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805a) Clarify whether the above fire areas (or other non-MCR areas) contain fire scenarios forwhich primary command and control is not retained in the MCR (i.e., the MOR isabandoned), and if so, explain how this decision was reached.b) If primary command and control is retained in the MCR, then RG 1.205 states, "Operationof dedicated or alternative shutdown controls while the MCR remains the command andcontrol location would normally be considered a recovery action." If actions taken at thePCS are not considered RAs for scenarios in which primary command and control areretained in the MCR, assess the impact of treating such actions consistent with RG 1.205on both the delta risk and additional risk of RAs as part of the integrated analysisperformed in response to PRA RAI 03. Additionally, discuss the results of the feasibilityand reliability evaluation of any new RAs in accordance with FAQ 07-0030.c) For scenarios in which primary command and control is not retained in the MCR and isinstead transferred to the PCS, the actions taken at the PCS are not RAs, and the MOR isassumed to be abandoned on loss of control (or function). Describe these scenarios,discussing how actions taken prior to and after MCR abandonment are modeled in theFPRA and its HRA. Additionally, explain the cues that result in the decision to abandonand their timing, identify the instruments being relied upon to make the abandonmentdecision, discuss whether the identified instruments are protected, and discuss how failureto transfer control to the PCS is taken into account.CCNPP RESPONSE PRA RAI 15:Response to be provided 4/113/15.PRA RAi 16 -State-of-Knowledgqe Correlation: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 NEPA 805 basedprogram, and all future plant changes to the program, shall be acceptable to the NRC.RG 1.174 provides quantitative guidelines on CDF, LERF, and identifies acceptable changes tothese frequencies that result from proposed changes to the plant's licensing basis anddescribes a general framework to determine the acceptability of risk-informed changes. TheNRC staff's review of the information in the LAR has identified additional information that isrequired to fully characterize the risk estimates.Section 4.7.3 of the L4R explains that the sources of uncertainty in the FPRA were identified,and specific parameters were analyzed, for sensitivity in support of the NFPA 805 FRE process.It is further explained that during the FRE process, the uncertainty and sensitivity associatedwith specific FPRA parameters were considerations in the evaluation of the change in riskrelative to the applicable acceptance thresholds. Based on these explanations, it appears thatthe risk results presented in Attachment W of the LAR are point estimates and do not includeparameter uncertainty. Explain how the SOKC was taken into account in the FPRAquantification, including fire ignition frequencies, circuit failure likelihood and hot short duration,and non-suppression probabilities. If the SOKC for these parameters was not addressed in theFPRA quantification, then include the impact of the SOKC for these parameters in the integratedanalysis performed in response to PRA RA! 03.45 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805CCNPP RESPONSE PRA RAI 16:The risk information presented in Attachment W is based on point estimate calculations. This isconsidered a reasonable approximation to the equivalent mean calculation. The state-of-knowledge correlation (SOKO) only causes the mean to be larger when multiple correlatedparameters appear in a single cutset. The vast majority of the cutsets contain only uncorrelatedparameters (e.g. fire scenario frequency, a single HRA failure likelihood, and a non-suppressionprobability). There are cutsets that contain multiple operator actions, but these events arealready evaluated as a joint failure probability which addresses the correlation issue. There arealso cutsets which could contain multiple circuit failure likelihood values. The CCNPP FPRAuncertainty notebook (C0-UNC-001 Revision 1) indicates only a minor change in the risk whencircuit failure likelihoods are not credited (see sensitivity analysis summary below, for Unit 1(Unit 2 similar):Unit 1 Sensitivity Impact IU-.00USensitivity AnalysisFigure 15: Impact of Sensitivity In Terms of Change to Unit 1 CDFGiven the above, the use of point estimates in the delta risk calculations is considered to beappropriate.The uncertainty analysis documentation in C0-UNC-001 will be updated to include a comparisonof the mean to the point estimate, and a sensitivity study on the circuit failure likelihood.46 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805PRA RAt 17- Sensitivity Analysis on FAQ 08-0048 Fire Bin Frequencies: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 an FPRA and endorses, with exceptions and clarifications, NE! 04-02,Revision 2, as providing methods acceptable to the NBC staff for adopting a fire protectionprogram consistent with NEPA 805. Methods that have not been determined to be acceptableby the NRC staff, or acceptable methods that appear to have been applied differently thandescribed, require additional justification to allow the NBC staff to complete its review of theproposed method.The licensee's analysis appears to indicate that generic fire ignition frequencies were basedupon those provided in Supplement 1 to NUBEG/CR-6850. Chapter 10 of this supplement,however, states that a sensitivity analysis should be performed when using the fire ignitionfrequencies in the supplement instead of those provided in Table 6-1 of NUREG/CR-6850. Aspart of the response to PRA BAI 03, provide the results (i.e., CDF, LERF, AICDF and ALERF) ofa sensitivity analysis that evaluates the impact of using the supplement frequencies, consistentwith Chapter 10 of Supplement 1 to NUREG/CR-6850. If RG 1.17 4 risk acceptance guidelinesare exceeded, (1) discuss which ones are exceeded, (2) describe the fire protection or relatedmeasures that will be taken to provide additional DID, and (3) discuss conservatisms in theanalysis and the risk significance of these conservatisms.CCNPP RESPONSE PRA RAI 17:Response to be provided 4/13/15.PRA RAI 18 -Calculation of VFDRAzCDF andAzLERF: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.RG 1.174 provides quantitative guidelfines on CDF, LERF, and identifies acceptable changes tothese frequencies that result from proposed changes to the plant's licensing basis anddescribes a general framework to determine the acceptability of risk-informed changes. TheNBC staff's review of the information in the LAB has identified additional information that isrequired to fully characterize the risk estimates.Section W.2. 1 of the LAB provides some description of how the change-in-risk and theadditional risk of RAs associated with VFDRs is determined, but not enough detail to make theapproach completely understood. As a result, provide the following:a) A detailed definition of both the post-transition and compllant plant models used tocalculate the reported change-in-risk, including any special calculations for the MCR andother abandonment areas (if applicable). Include description of the model adjustmentsmade to remove VFDRs from the Compliant plant model, such as adding events or logic,or use of surrogate events. Also, provide an explanation of how VFDB- and non-VFDR-related modifications are addressed for both the post-transition and compllant plantmodels.47 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805b) Justification for the assumption in the licensee's analysis (Section 8.0 of EPM ReportB2215-008-024) that the risk associated with the post-transition p/ant model is consideredequivalent to that of the compliant plant model for scenarios requiring MCR abandonment.c) A description of how the reported additional risk of RAs was calculated, including anyspecial calculations performed for the MCR and other abandonment areas (if applicable).If non-VFDR-rela ted modifications are credited to reduce delta risk, equating theadditional risk of RAs (as discussed in W.2. 1) to the sum of the delta risks of the VFDRsthat are resolved by crediting an RA may be non-conservative. In this case, the additionalrisk of these RAs should be re-calculated consistent with FAQ 07-0030 as part of theintegrated analysis performed in response to PRA BAI 03.d) A summary of the types of VFDRs that were identified but not modeled in the EPRA.Include any qualitative rationale for excluding these from the change-in-risk calculations.e) A clarification of whether they DID RAs listed in Attachment G of the LAB are quantified inthe EPRA. Also, explain whether credit for such DID RAs is necessary for the change-in-risk to be acceptable.CCNPP RESPONSE PRA RAI 18:18a -Variances from deterministic requirements (VFDRs) were removed from the CCNPPFPRA compliant plant model by setting the VFDR related cables or basic events to false (norandom failure for basic events set to false). Basib events are set to false when equipment inthe room would be damaged by fire for those components whose loss causes the VFDR.The delta risk was obtained by quantifying this compliant case and comparing with the basemodel risk. All modifications, VFDR and non-VFDR related, were including in both the post-transition and compliant plant models.There were no special considerations for the MCR regarding the VFDR calculations.Further, in the submittal, fires in the MCR are the only fires that forced a complete MCRabandonment. The VEDR delta risk approach as described above was used in all areasMCR and non-MCR, abandonment and non-abandonment.1 8b -Response to be provided 4/13/15.18c -The compliant plant will be evaluated by setting the root cause failures for each VFDR(cables or basic events) to FALSE (i.e., by simulating a deterministically compliant version ofthe fire area). Risk will then be calculated and the difference between the base model andthe compliant case will be the delta risk. The delta risk of recovery action will be obtained bysubtracting the baseline risk from the calculated risk with the HRA successful (i.e., thedifference between the base model risk with the HRA set to zero, or equivalent compliantcontrol room HEP value, and the base model risk). This delta risk can be represented as:Delta Risk of Recovery Action = Rrecov base -Rrecov complWhere:Rrecov base = the baseline fire risk of associated scenario with credit for a recoveryModifications not associated with VFDRs are not credited to reduce delta risk.48 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805The results of the compliant plant evaluation described above will be will be submitted inconjunction with the response to RAI PRA-03.18d -Variances from deterministic requirements were identified for the loss of required powersupplies; loss of support systems, including HVAC; loss of pumps for credited flow paths;failures of valves for credited flow paths; and, spurious operations in both credited and non-credited paths which could impact the NSPC.VFDRs were excluded from the change in risk calculations if the component or componentscovered by the VFDR were screened out during the CCNPP FPRA's component selectionprocess. Examples include VFDRs covering non-FPRA credited instrumentation (e.g.,Reactor Coolant System temperature indication) and pressurizer heaters.18e -The instances in which recovery actions are specified as Risk!/ Defense In Depth (DID)(i.e., credited for both risk and DID), the actions will either be retained as Risk actions onlyand will be included in the CCNPP FPRA risk quantification or they will be re-classified asDID only and will not be credited in the FPRA risk quantification. This will ensure that all DIDactions are credited as additional actions for which credit is not included in the CCNPP FPRAquantification.PRA RA! 19 -Attachment W Inconsistencies:Several inconsistencies were noted within Attachment W as well as between its tables andthose in Attachments C and G for particular fire areas. In light of this:a) Provide clarification on the following inconsistencies, and discuss their significance to therisk results reported in Tables W-6 and W-7:i.In Table W-6, Unit 1 Fire Areas 2, 8, 13, 18, 18A, 22, 23, 25, 26, 27, 28, 31, 38, 40,and 2CNMT are indicated as Deterministically Compliant (4.2.3.2); however, they areindicated as having VFDRs (i.e., there is a "Yes"' under the "VFDR" column andsometimes under the "RAs" column) as well as very small risk values (i.e., Fire Area18) or epsilon for ACDF/ALERF. Similarly, in Table W-7, Unit 2 Fire Areas 3, 4, 6, 14,15, 19, 19A, 21, 30, 33, 39, and 1CNMT are noted as Deterministically Compliant(4.2.3.2); however, they are indicated as having VFDRs and very small risk values(i.e., Fire Areas 19 and 30) or epsilon for ACDF/A4LERF. Attachment C does notidentify any of the above deterministic fire areas as having VFDRs. Furthermore,while for most of these fire areas the AICDF/ALERF and additional risk of RAs isreported to be epsilon, actual (very small) numerical values are reported forA4CDF/ALERF for Unit 1 Fire Area 18 and for Unit 2 Fire Areas 19 and 30, and actual(very small) numerical values are reported for additional risk of RAs for Unit 1 FireArea 23.ii. In Table W-6, Unit 1 Fire Areas 12, 14, 15, 19A, 21, 30, 32, 33, 35, 36, 39, 1CNMT,and IS are indicated as Performance-Based (4.2.4.2) and as having an RA credited inthe FPRA (i.e.. there is a "Yes"' under the "RAs" column); however, no RAs aredescribed in the VFDR dispositions presented in Attachment C or listed in AttachmentG for these areas. Similarly, Unit 2 Fire Areas 12, 13, 18A, 20, 26, 27, 28, 32, 34, 35,36, 40, 2CNMT, and IS are indicated as Performance-Based (4.2.4.2) and identify a"Yes"' under RA; however, no RAs were described in the VFDR dispositionspresented in Attachment C or listed in Attachment G for these areas. Furthermore,49 ATT-ACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805while for most of these fire areas the additional risk of FAs is reported to be epsilon,actual (very small) numerical values are reported for Unit 1 Fire Areas 21 and 36 andfor Unit 2 Fire Area 13.iii. The LERF (9.49E-O8/year (yr) reported in Table W-4 for scenario PAU CC-1A-C(Complete Burn of Vertical Cable Chase 1A) is greater than the total LERF(4.04E-OS/yr) reported in Table W-6 for Fire Area 20 (Cable Chase 1A). For twoscenarios reported in Table W-4 (PAU 230E-C and PAU 230W-C,), which representfires in Unit 1 Containment, the summation of their LERF (1.99E-O7/yr) is greater thanthe total LERE (1.91 E-O7/yr) reported in Table W-6 for Fire Area 1CNMT (Unit 1Containment). These inconsistencies also exist between Tables W-5 and W-7 for thesame scenarios in Unit 2.iv. The Table W- 1 Unit 1 fire LERE of 3.2E-O6/(chemical reactor (rx)-yr) does not matchthe corresponding value reported in Table W-6. Similarly, the Table W-1 Unit 2 fireLERF of 4.4E-O6/(rx-yr) does not match the corresponding value reported in TableW-7.b) Describe what is meant by the use of "e," or epsilon, in columns for Fire Area CDF/LERF,ACDF/AlLERF, and additional risk of RAs. Address if epsilon is defined by a specific cut-off value(s). Also, clarify how an actual value for LERE can be reported while epsilon isreported for the corresponding CDF (i.e., Unit 1 Fire Area 24 for additional risk of RAs,*Unit 2 Fire Areas 8 and 10 for CDF/LERFandA CDF/LERF).c) Describe what is meant by the use of "N/A" in columns for Fire Area CDF/LERF,ACDF/A1LERF, and additional risk of RAs. In doing so, clarify the basis for not reportingFire Area CDF/LERF values (or epsilon) for Unit 1 and Unit 2 Fire Areas 44, A B-i, AB-3,ABFL, DGB1, DGB2, and TBFL.d) Tables W-6 and W-7 include a risk reduction credit for internal events that is described ina footnote to these tables as covering random failures and internal floods. This riskreduction credit is used to offset the increase in fire risk reported in these tables. Explainhow the risk reduction from internal events reported in these tables is calculated.CCNPP RESPONSE PRA RAI 19:1 9a -Response to be provided 4/13/15.1 9b -Epsilon is not defined by a specific cutoff. It is used to indicate that the risk contributionfrom that element is negligible. A review of the usage of epsilon will be done in conjunctionwith the final quantification and submittal of RAI 3 Attachment W tables to eliminateinconsistencies in the use of this term in the LAR tables.1 9c -Calvert Cliffs Nuclear Power Plant LAR Attachment W, Tables W-6 and W-7 utilized "N/A"in the columns for "Fire Area CDF/LERF, ACDF/ALERF" and "Additional Risk of RAs" for FireAreas where there were no VFDRs and/or no risk assessments. Attachment W will berevised to provide an explanation of this use of "N/A" in the final Tables W-6 and W-7 whichwill be submitted in conjunction with the response to RAI PRA-03.1 9d -Response to be provided 4/13/1550 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FlIREPROTECTION ASSOCIATION STANDARD 805-PRA RAI120- Implementation Item Impact on Risk Estimates:Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall beacceptable to the NRC. Section 2.4.4.1 of NFPA 805 further states that the change in publichealth risk arising from transition from the current fire protection program to an NFPA 805 basedprogram, and all future plant changes to the program, shall be acceptable to the NRC.AG 1.174 provides quantitative guidelines on CDF, LERF, and identifies acceptable changes tothese frequencies that result from proposed changes to the plant's licensing basis anddescribes a general/framework to determine the acceptability of risk-informed changes. TheNRC staff's review of the information in the LAR has identified additional information that isrequired to fully characterize the risk estimates.Table S-3, Implementation Item 12 of the EAR commits to updating the EPRA and verifying therisk results after "risk related" plant modifications have been incorporated. However, it isunclear to which modifications the implementation item refers. Update Implementation Item 12to reflect completion of both the Table S-2 modifications and Table S-3 implementation itemsbefore this verification.CCNPP RESPONSE PRA RAI 20:Any as-built changes that affect the CCNPP FPRA model will include an evaluation of whetherthe acceptance criteria and the delta risk criteria are still satisfied.We propose revising Implementation Item 12 to read as follows:Incorporate as built risk related modifications, implementation items and any other additionalrefinements that may be needed into the Fire PRA and Internal Events Model and verify the riskresults are not appreciably changed. The credited modifications will be tracked through thedesign input and the engineering configuration control process. The PRA model will be updatedas necessary to reflect the final change package, and impacts to the risk estimates will beverified. As the actual engineering implementation of each modification is developed in concertwith Fire PRA evaluations of the proposed change, if the Fire PRA indicates that the as-builtchange in risk would not meet the acceptance criteria as described in LAR Section 4.5.2.2, themodification under development or its representation in the Fire PRA will be refined to ensurethat the acceptance criteria are satisfied. In addition, CCNPP will verify the validity of thereported change-in-risk estimates of Attachment W following completion of both PRA-creditedmodifications and implementation items. If this verification determines that the risk metrics havechanged such that the RG 1.205 acceptance guidelines are not met, the NRC will be notifiedand additional analytical efforts, and/or procedure changes, and/or plant modifications will bemade to assure the RG 1.205 risk acceptance criteria are met.This implementation item is an on-going action initiated within the 180 day timeframe forcompletion of implementation items but only complete after completion of modificationimplementation per Table S-2.We also propose revising the last paragraph of LAR Section 4.8.2 to read as follows:The FPRA model represents the as-built, as-operated and maintained plant as it will beconfigured at the completion of the transition to NFPA 805. The FPRA model includes credit forthe planned implementation of modifications identified in Attachment S, Table S-2. Following51 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE____ PROTECTIONASSOCIATION STANDARD 805installation of modifications and the as-built installation details, additional refinementssurrounding the modifications may need to be incorporated into the FPRA model (the FPRA willverify the validity of the reported change-in-risk on as-built conditions after the modifications arecompleted). However, these changes are not expected to be significant. No other significantplant changes are outstanding with respect to their inclusion in the FPRA model (seeAttachment S, Table S-3, IMP-12). CCNPP will verify the validity of the reported change-in-riskestimates of Attachment W following completion of both PRA-credited modifications andimplementation items. If this verification determines that the risk metrics have changed suchthat the RG 1.205 acceptance guidelines are not met, the NRC will be notified and additionalanalytical efforts, and/or procedure changes, and/or plant modifications will be made to assurethe RG 1.205 risk acceptance criteria are met.We propose revising the first bullet of LAR Section 5.5 to read as follows:Implementation of new NFPA 805 fire protection program to include procedure changes,process updates, and training to affected plant personnel. This will occur 180O daysfollowing the issuance of an approved SER from the NRC unless that date falls within ascheduled refueling outage. Then, implementation will occur 60 days after startup fromthat scheduled refueling outage. See Attachment S, Table S-3. It should be noted thatimplementation item IMP-12 is associated with incorporation of the NPFA 805modifications and the completion of this implementation item is an on-going actioninitiated within the 180 day timeframe for completion of implementation items but onlycomplete after completion of modification implementation per Table S-2.these procedurechanges, process updates, and training represent implementation items that are on-goingactions initiated within the 180 day timeframe for completion, but will only be completeafter completion of modification implementation per Table S-2.We propose revising Item (3) on Page M-5 of LAR Attachment M to read as follows:(3) The licensee shall implement the items listed in Enclosure 1, Attachment S, Table S-3,"Implementation Items," from license amendment request dated ____within 1J80 days afterNRC approval unless that implementation date falls within a scheduled refueling outage. Then,implementation will occur 60 days after startup from that scheduled refueling outage. It shouldbe noted that implementation item IMP-i12 is associated with incorporation of the NPFA 805modifications and the completion of this implementation item is an on-going action initiatedwithin the 180 day timeframe for completion of implementation items but only complete aftercompletion of modification implementation per Table S-2.1t should be noted that theseimplementation items are on-going actions initiated within the 180 day timeframe for completionof implementation items but only complete after completion of modification implementation perEnclosure 1, Attachment 5, Table S-3.Markups of the submitted LAR pages for each of the above proposed revisions are enclosed.PRA RAI 21 -Internal Events Peer Review:Section 2.4.3.3 of NEPA 805 states that the PRA approach, methods, and data shall beacceptable to the NRC. The RG 1.205 identifies NUREG/CR-6850 as documenting amethodology for conducting an FPRA and endorses, with exceptions and clarifications,NE! 04-02, Revision 2, as providing methods acceptable to the NRC staff for adopting a fireprotection program consistent with NFPA 805. The RG 1.200 describes a peer review process52 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805utilizing an associated ASME/ANS standard (currently ASME/ANS-RA-Sa-2009) as oneacceptable approach for determining the technical adequacy of the PRA once acceptableconsensus approaches or models have been established.Attachment U of the LAR indicates that the full-scope IEPRA peer review was performed againstASME/ANS PRA Standard, RA-S-2OO8a. In light of this observation, if RG 1.200, Revision 2,and ASME/ANS PRA Standard, RA-Sa-2009, were not used as the basis for the peer review ofthe IEPRA, then discuss whether any differences between SRs were evaluated and whetherthey had any impact on the application.CCNPP RESPONSE PRA RA! 21:The peer review of the CCNPP internal events PRA was completed against RegulatoryGuide 1.200 Revision 2 and American Society of Mechanical Engineers (ASME) / AmericanNuclear Society (ANS) PRA Standard RA-Sa-2009.PRA RA1 22- PRA Up~qrades:Section 2.4.3.3 of NEPA 805 states that the PRA approach, methods, and data shall beacceptable to the NRC. The RG 1.205 identifies NUREG/CR-6850 as documenting amethodology for conducting an FPRA and endorses, with exceptions and clarifications,NEI 04-02, Revision 2, as providing methods acceptable to the NRC staff for adopting a fireprotection program consistent with NEPA 805. The RG 1.200 describes a peer review processutilizing an associated ASME/ANS standard (currently ASME/ANS-RA-Sa-2009) as oneacceptable approach for determining the technical adequacy of the PRA, once acceptableconsensus approaches or models have been established.The LAR does not indicate whether any changes made to the IEPRA or FPRA since their mostrecent full-scope peer reviews are consistent with the definition of a "PRA upgrade" inASME/ANS-RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency for NuclearPower Plant Applications," as endorsed by RG 1.200, Revision 2. In light of this, identify anysuch changes. If a focused-scope peer review has not been performed for the identifiedchanges, describe what actions will be implemented to address this issue. If a focused-scopepeer review has been performed, confirm whether it was done consistent with the guidance inASME/ANS-RA-Sa-2 009, as endorsed by RG 1.200, and provide any findings and theirresolutions.CCNPP RESPONSE PRA RAl 22:No "PRA upgrades" (as defined in ASME/ANS-RA-Sa-2009, Regulatory Guide 1.200 Revision 2and closed out Frequently Asked Questions) have been performed, or are planned to beperformed, to the CCNPP Internal Events PRA model or the CCNPP FPRA model. As such, afocused scope peer review is not required. If in the course of developing the RAI responses achange to the FPRA is identified which constitutes a PRA upgrade, a focused scope peerreview will be initiated in accordance with PRA analysis upgrade procedure requirements inExelon Training and Reference Material (T&RM) guidance documents ER-AA-600-1015 (FPIEPRA Model Update) and ER-AA-600-1 061 (Fire PRA Model Update and Control).PRA RA1 23- Deviations from Acceptable Methods: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 public53 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805health risk arising from transition from the current fire protection program to an NFPA 805 basedprogram, and all future plant changes to the program, shall be acceptable to the NRC. TheRG 1.174 provides quantitative guidelines on CDF, LERF, and identifies acceptable changes tothese frequencies that result from proposed changes to the plant's licensing basis anddescribes a general framework to determine the acceptability of risk-informed changes. TheNRC staff's review of the in formation in the LAR has identified additional information that isrequired to fully characterize the risk estimates.Section 4.5.1.2 of the LAR states that the FPRA model uses "a methodology consistent with theguidance provided in NUREG/CR-6850 and subsequent clarifications documented in responsesto NFPA 805 FAQs" and that "[n]o unre viewed methods or deviations from NUREG/CR-6850were utilized in the EPRA model development." Indicate if any other methods were employedthat deviate from other NRC-accepted guidance (e.g., subsequent clarifications documented inFAQs, interim guidance documents, etc.). If so, describe and justify any proposed method thatdeviates from NRC guidance, or replace the proposed method with an accepted method. Also,include the proposed method as a method "currently under review" as part of the integratedanalysis in the response to PRA RAI 03.CCNPP RESPONSE PRA RA! 23:Only mutually agreed upon methods will be used in preparation of the final CCNNP FPRAsupporting the NEPA 805 LAR and the responses to the associated RAIs. The peer review ofthe CCNPP FPRA that was completed against Regulatory Guide 1.200 Revision 2 andAmerican Society of Mechanical Engineers (ASME) I American Nuclear Society (ANS) PRAStandard RA-Sa-2009 did not find any unapproved methods. It is our intent to continue to useonly mutually agreed upon methods for any future work done in support of the CCNPP FPRA.and the LAR. Agreement on acceptable approaches based on the interpretation of acceptablemethods will be resolved through the RAI process.PRA RAE 24 -Defense-in-Dep~th and Safety Margqin: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 NEPA 805 basedprogram, and all future plant changes to the program, shall be acceptable to the NRC. TheRG 1.174 provides quantitative guidelines on CDF, LERF, and identifies acceptable changes tothese frequencies that result from proposed changes to the plant's licensing basis anddescribes a general framework to determine the acceptability of risk-in formed changes. TheNRC staff's review of the information in the LAR has identified additional information that isrequired to fully characterize the risk estimates.LAR Section 4.5.2.2 provides a high-level description of how the impact of transition to NFPA805 impacts DID and safety margin was reviewed, including using the criteria from Section 5.3.5of NE/ 04-02 and from RO 1.205. However, no explanation is provided of how specifically thecriteria in these documents were utilized and/or applied in these assessments.a) Provide further explanation of the method(s) or criteria used to determine when asubstantial imbalance between DID echelons existed in the FREs, and identify the typesof plant improvements made in response to this assessment.b) Provide further discussion of the approach in applying the NEt 04-02, "Guidance forImplementing a Risk-Informed, Performance-Based Fire Protection Program Under 1054 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805CFR 50.48(c)," Revision 2 (ADAMS Accession No. ML081 130188) criteria for assessingsafety margin in the FREs.CCNPP RESPONSE PRA RAI 24:24aThe method used in the CON PP .FPRA Fire Risk Evaluations (FREs) to determine when asubstantial imbalance between DID echelons existed was based on the guidance in NEI 04-02,Revision 2. Specifically, a review was performed of each DID echelon on a fire area basisbased on the following considerations:Echelon 1 (Prevent fires from starting):Combustible and hot work controls are fundamental elements of DID and as such are always inplace. The issue considered during FRE was whether this element needs to be strengthened tooffset a weakness in another echelon, thereby providing a reasonable balance.Considerations included:* Creating a new Transient Combustible Free Area;* Creating a new Hot Work Restriction Area; and/or,* Modifying an existing Transient Combustible Free Area or Hot Work Restriction Area.The fire scenarios involved in the FRE quantitative calculation were reviewed to determine ifadditional controls should be added.The remaining elements of DID were reviewed to ensure an overreliance was not placed onprogrammatic activities for weaknesses in plant design.Echelon 2 (Rapidly detect, control, and extinguish promptly those fires that do occur therebylimiting fire damage):Automatic suppression and/or detection may or may not exist in the fire area in question. Theissue considered during the FRE was whether installed suppression and/or detection wasrequired for DID or whether suppression and/or detection needed to be strengthened to offset aweakness in another echelon, thereby providing a reasonable balance.Considerations included:Risk Insights:* If existing VFDRs were never affected in a potentially risk significant fire scenario, manualsuppression capability was generally considered adequate and no additional systemsrequired.Recovery Actions:* If the fire area required recovery actions, typically detection and manual suppressioncapability were considered required. Additionally, requiring automatic suppression wasconsidered.55 A'1-ACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FiREPROTECTION ASSOCIATION STANDARD 805* If the fire area contained neither suppression, nor detection; and a recovery action wasrequired, the addition of automatic detection and/or suppression systems wereconsidered.Firefighting Activities:* If firefighting activities in the fire area were expected to be challenging (either due to thenature of the fire scenario or accessibility to the fire location), then the addition of bothsuppression and detection were considered, if absent.Fire Scenarios:* If fire scenarios credited fire detection and/or fire suppression systems, then these werealready considered to form an integral part of DID.Echelon 3 (Provide adequate level of fire protection for systems and structures so that a fire willnot prevent essential safety functions from being performed):If fires occur and they are not rapidly detected and promptly extinguished, then the third echelonof DID would be relied upon. The issue considered during the FRE was whether existingseparation was adequate (or over relied on) and whether additional measures (e.g.,supplemental barriers, fire rated cable, or recovery actions) were required to offset a weaknessin another echelon, thereby providing a reasonable balance.Considerations included:Risk Insights:* If existing VFDRs were not affected in a "potentially risk significant" fire scenario, internalfire area separation was considered adequate and no additional reliance on recoveryactions was considered necessary.* If existing VFDRs were affected in a risk significant fire scenario, internal fire areaseparation may not be adequate and reliance on a recovery action, supplemental barrier,or other modification was considered.*If the consequence associated with existing VFDRs were considered high(e.g., CCDP>I E-01 or by qualitative Safe Shutdown (SSD) assessment), regardless of-whether it is in a risk significant fire scenario, a recovery action, supplemental barriers, orother modification was considered.*There are known modeling differences between a FPRA and NSCA due to differentsuccess criteria, end states, etc. Although a VFDR may be associated with a function thatis not considered a significant contribution to core damage frequency, in some cases, theVFDR may have been considered important enough to the NSCA to retain a recoveryaction as credited for DID, but not required for risk.The fire scenarios involved in the FRE quantitative calculation were reviewed to determine thefires evaluated and the consequence in the area to best determine options for this element ofDID.Each fire area was evaluated for the need to incorporate DID enhancements to provideassurance that plant performance goals can be achieved and maintained. Documentation of56 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTION ASSOCIATION STANDARD 805these DID enhancements was tied directly to applicable VFDR dispositions in the respective firearea FRE.DID enhancements that were specifically implemented consisted of implementing transientcombustible free and hot work restricted areas; credit for fire detection systems; creditsuppression systems; and credit for DID recovery actions. These DID enhancements wereimplemented based on the above considerations.24bIn accordance with NEI 04-02, Revision 2, the maintenance of adequate Safety Margin wasassessed by the consideration categories of analyses used by the CCNPP FPRA FREs. Safetymargins were considered to be maintained if:* Codes and standards or their alternatives accepted for use by the NRC are met; and,* Safety analysis acceptance criteria in the licensing basis (e.g., UFSAR, supportinganalyses, etc.) are met, or provide sufficient margin to account for analysis and datauncertainty.The requirements related to safety margins for the FREs were evaluated for each specificanalysis type. These analyses can be grouped into four categories. These categories are:1. Fire Modeling;2. Plant System Performance;3. PRA Logic Model; and,4. Miscellaneous.As described in the CCNPP FREs, Safety Margins were maintained as follows:1) Fire Modelinga) For all fire areas assessed, the "bounding risk assessment" or "combined analysisapproach" was used as follows:i) Where the "bounding risk assessment" was used per NE! 04-02, Section 5.3.4.2, theanalysis conservatively assumed that target set damage occurred for postulated fireevents which resulted in whole room burn up. Fire modeling was not performed insupport of the change evaluations for such areas and results were therefore basedon whole area burn up. As such, the results are considered bounding.ii) Where the "combined analysis approach" was used per NEi 04-02, Section 5.3.4.3,fire modeling was performed in support of the transition within the CCNPP FPRAusing codes and standards developed by industry and NRC staff which have beenverified and validated in authoritative publications.b) The Risk-Informed, Performance-Based (RIPB) processes used were based upon NFPA805, 2001 edition, as endorsed by the NRC in 10 CFR 50.48(c).2) Plant System Performancea) The FRE process was performed in accordance with NEI 04-02, Revision 2. RegulatoryGuide 1.205 endorses portions of NEI 04-02, Revision 2, where it has been found toprovide methods acceptable to the NRC for implementing NFPA 805 and complying with10 CFR 50.48(c).57 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE________ PROTECTION ASSOCIATION STANDARD 805__b) Fire protection systems and features determined to be required by NFPA 805 Chapter 4were confirmed to meet the requirements of NFPA 805 Chapter 3 and their associatedreferenced codes and listings, or provided with acceptable alternatives using processesaccepted for use by the NRC.3) FPRA Logic Modela) The CCNPP FPRA was developed in accordance with NUREG/CR-6850, which was developedjointly between the NRC and EPRI.b) The CCNPP FPRA has undergone an industry peer review, in order to ensure that itmeets the appropriate quality standards of ASME / ANS Joint Standard RA Sa 2009,"Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment forNuclear Power Plant Applications," dated February 2, 2009.c) In accordance with the requirements of 10 CFR 50.48(c)(iii), the CCNPP FPRA results,including the sequences for the scenarios of concern, have been reviewed and it wasverified that the results do not rely solely on feed and bleed as the fire-protected safeshutdown path for maintaining reactor coolant inventory, pressure control, and decayheat removal capability for the fire areas considered.4) Miscellaneousa) No analyses that were not addressed by the above categories were identified.Example of a Typical Safety Margin Review as Contained in a FRE for a Fire Area With One orMore VFDRsIn accordance with NEI 04-02, the maintenance of adequate safety margin is assessed by theconsideration categories of analyses utilized by this FRE.Safety margins are considered to be maintained if:&deg; Codes and Standards or their alternatives accepted for use by the NRC are met.AND&deg;Safety analyses acceptance criteria in the licensing basis (e.g., FSAR, supportinganalyses) are met, or provides sufficient margin to account for analysis and datauncertaintyThe following summarizes the bases for ensuring the maintenance of safety margins:* The risk-informed, performance based processes utilized are based on NFPA 805, 2001edition, endorsed by the NRC in 10 CFR 50.48(c).* The FRE process is in accordance with NEI 04-02, Revision 2, which is endorsed by theNRC in RG 1.205, Revision 1.* The FPRA is developed in accordance with NUREG/CR-6850, which was developedjointly between the NRC and EPRI.&deg;The FPRA has undergone an industry peer review, in order to ensure the FPRA meets theappropriate quality standards of ASME/ANS Joint Standard RA-Sa-2009.o The "combined analysis approach" is used during transition (NEi 04-02, Section 5.3.4.3);therefore, MEFS/LFS is not analyzed separately from the FPRA results, or58 ATTACHMENT (1)REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIREPROTECTIONASSOCIATIONSTANDARD 805 ______* The "bounding risk evaluation" is used during transition (NEI 04-02, Section 5.3.4.3);therefore, MEFS/LFS is not analyzed separately from the FPRA results.* The CCNPP internal events PRA model received a formal industry peer review conductedin accordance with applicable NEI guidelines. The full-scope peer review was performedin 2010. The Pressurized Water Reactor Owners Group (PWROG) peer review coveredall aspects of the CCNPP PRA model and administrative processes used to maintain themodel. All findings which could impact the FPRA results have been dispositioned, and theCCNPP PRA model has been revised as appropriate. Some findings not relevant toFPRA, such as those related to internal flooding, are pending. Refer to LAR Attachment U-Internal Events PRA Quality -for the final disposition of internal events F&Os as relatedto FPRA.* Fire protection systems and features determined to be required by NFPA 805 Chapter 4have been confirmed to meet the requirements of NFPA 805 Chapter 3 and theirassociated referenced codes and listings, or provided with acceptable alternatives usingprocesses accepted for use by the NRC (i.e., FAQ 06-0008, FAQ 06-0004, 07-0033).* Fire modeling performed in support of the transition has been performed within the FPRAutilizing codes and standards developed by industry and NRC staff which have beenverified and validated in authoritative publications, such as NUREG-1 824, "Verificationand Validation of Selected Fire Models for Nuclear Power Plant Applications." In general,the fire modeling performed in support of the FREs has been performed usingconservative methods and input parameters that are based upon NUREG/CR-6850 asdocumented in the detailed fire modeling notebook. While this is generally not ideal in thecontext of best estimate probabilistic risk analysis, it is a pragmatic approach given thecurrent state of knowledge regarding the uncertainties related to the application of the firemodeling tools and associated input parameters for specific plant configurations.* In accordance with the requirements of 10 CFR 50.48(c)(iii), the FPRA results, includingcutsets for the scenarios of concern, have been reviewed and it was verified that theresults do not rely solely on feed and bleed as the fire-protected safe shutdown path formaintaining reactor coolant inventory, pressure control, and decay heat removal capabilityfor the fire areas.59 ENCLOSURE1IUPDATED PAGESCalvert Cliffs Nuclear Power PlantFebruary 9, 2015 Constellation Energy Nuclear Group Attachment A -NEI 04-02 Table B-1 -Transition of Fundamental FP Program and Design ElementsNFPA 805 Ch. 3 Ref. RequirementelGuldance Compliance Statement Compliance Basis Reference Document3.4.1 (c)During every shift, the brigadeleader and at least two brigademembers shall have sufficienttraining and knowledge ofnuclear safety systems tounderstand the effects of fireand fire suppressants on nuclearsafety performance criteria.Exception: Sufficient training andknowledge shall be permitted tobe provided by an operationsadvisor dedicated to industrialfire brigade support.The industrial fire brigade shallbe notified immediately uponverification of a fire.Complies withClarificationAdo.3l ......... t. CCNPP complies byutilizing the exception to this section of NFPA805. CCNPP adminstrative procedures andthe UFSAR ensure that an OperationsTechnical Advisor (OTA), a licensed operatorposition, is dedicated to industrial fire brigadesupport.4A4Procedure NO-1 -200,Control of Shift Activities. Rev.055001/ Section 5.1 .AUpdated Final Safety AnaysisReport (UFSAR). Rev. 471ISection 9.9.5FPERAI023.4 1(d)CompliesNo Additional ClarificationProcedure ERPIP-3.0, ImmediateActions, Rev. 05101 /Attachment16, Section 2Procedure SA-1-101, Fire Fighting,Rev. 00600/1 Section 5.4.A and5.4.B3.4.1(e)Each industrial fire brigademember shall pass an annualphysical examination todetermine that he or she canperform the strenuous activityrequired during manualfirefighting operations. Thephysical examination shalldetermine the ability of eachmember to use respiratoryprotection equipment.CompliesNo Additional ClarificationProcedure CNG-MD-1 .01-3000,Physical Examination Process forEmployees and Contractors, Rev.00300 / Section 5.6.A.4Procedure SA-1-105, Fire BrigadeTraining, Rev. 00500 /Sections5.5.A and 5.5.A.4CCNPP Page A-42CCNPPPage A-42 Constellation Ene~rgy Nuclear Group Attachment A -NEI 04-02 Table B-I -Transition, of Fundamental FP Program and Design ElementsNFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document3.10.1If an automatic total floading andlocal application gaseous firesuppression system is requiredto meet the performance ordeterministic requirements ofChapter 4, then the system shallbe designed and installed inaccordance with the followingapplicable NFPA codes:NFPA 12, Standard on CarbonDioxide Extinguishing SystemsN/AN/A -General Statement; No TechnicalRequirementsNIAN/A3. 10. 1(1)N/ACarbon dioxide systems are not used at CCNPP.3.,10.,1(2)NFPA 12A, Standard on Haion Complies, with Required1301 Fire Extinguishing Systems ActionThe halon systems are in compliance with NFPA12A-1971 Edition as identified in report 51-9124415-~000. The halon systems are set for automaticoperation, and manual operation is accomplished bythe manual release station(s) located at the roomexit. Manual release stations are connected to thePyrotronics System 3 that is provided with batteryback-up in case of power outage per Document ES-2012-000156-001.Letter ES-2012-0001 56-001Evaluate Halon System ManualAction and Determine If CurrentSet Up and NRC CorrespondenceDeem It Acceptable, datedJanuary 31, 2013 /AllNFPA 12A, H-alon 1301Extinguishing Systems, 1971Edition / AllFAQ 08-0054, DemonstratingCompliance with Chapter 4 ofNFPA 805, Rev. 1 / AllReport 51-9124415-000, CCNPPCode Compliance Reviews, Rev.000 / Appendix C and Section 4.3See Attachment S, Table S-2, Item 44 17.IFPERA1 03ICIINII IIIIIPIIIgIeII II IIIICCNPPPage A-94 Constellation Energy Nuclear Group Attachment A -NEI 04-02 Table B-I -Transition of Fundamental FP Program and Design ElementsNFPA 805 Ch. 3 Ref. RequirementslGuldance Compliance Stlatment Compliance Basis Reference Document3.10.3Ventilation system design shalltake into account preventionfrom over-pressurization duringagent injection, adequatesealing to prevent loss of agent,and confinement of radioactivecontaminants.Complies, with RequiredActionDrawings 12263-0003, 12276-0006 and60714SH0004 show door control wiring/circuitsconnected to the halon system control panel.Procedures STP-M-699-1, STP-M-699-2, STP-M-698-1, STP-M-698-2, FTE-72, and FTE-73 outlinethe testing protocol for control functions related to ofthe integrity of the room enclosures. Confinement ofradioactive contaminants is not a concern as thereare no halon systems installed in radiological areas.Procedures STP-M-699-1, STP-M-699-2, STP-M-698-1 and STP-M-698-2 all state the followingpurpose in Section 1:"A. Verify the Halon systems for the Unit 2 CableSpreading Room (CSR) would actuate manually andautomatically upon actuation signal.B. Verify that upon a simulated actuation, theassociated ventilation dampers actuate.C. Verify that associated fire door monitor switchesare operable."Procedures STP-M-699-1, STP-M-699-2, STP-M-698-1, STP-M-698-2, FTE-72 and FTE-73 all statethe following purpose in Section 6 provides theprocedure for performance testing of the halonsystem(s) including verdfying door monitors.Item 48 of Appendix C of Report 9124415-000indicates: "The results of discharge testingsummarized in the referenced letters confirm that theenclosure strength is adequate such that pressurebuild-up is not a concern."See Attachment S, Table S-2, Item 418 17.Drawing 12263-0003, Halon 1301Fire Supression System, Rev. 11 /AllDrawing 12276-0006, Halon 1301Fire Suppression System, Rev. 6 IAllDrawing 60714SH0004, Halon1301 Fire Suppression System,Rev. 11 /AllProcedure FTE-72, Unit 1 DASRoom Halon System FunctionalTest, Rev. 00500 / Section 6.0Procedure FTE-73, Unit 2 DASRoom Halon System FunctionalTest, Rev. 00500 / Section 6.0Procedure STP M-698-1,Functional Test of Halon Systemfor the Unit 1 Cable SpreadingRoom, Rev. 0101 / Section 1.0 and6.0Procedure STP M-698-2,Functional Test of Halon Systemfor the Unit 2 Cable SpreadingRoom, Rev. 0201 / Section 1.0 and6.0Procedure STP M-699-t,Functional Test of Halon Systemfor the Unit 1 Switchgear Rooms,Rev. 00600 / Section 1.0 and 6.0Procedure STP M-699-2,Functional Test of Halon Systemfor the Unit 2 Switchgear Roams,Rev. 0501 /Section 1.0 and 6.0Report 51-9124415-000, CCNPPCode Compliance Reviews, Rev.000 / Appendix C, Item 48FPERAI 03CCNPP Page A-98CCNPPPage A-96 Constellation Energy Nuclear Group Attachment A -NEI 04.02 Table B-I -Transition of Fundamental FP Program and Design ElementsNFPA 805 Ch. 3 Ref. RequirementslGuidance Compliance Statement Compliance Basis Reference DocumentComplies, with Required Implementation items are identified below. NoneActionIMPLEMENTATION ITEMS ISee Attachment S. Table S-3IMP-2 Performance-based surveillance frequencies may be updated based on the guidance of Electric Power Research Institute (EPRI) TechnicalReport TR-1006756, "Fire Protection Surveillance Optimization and Maintenance Guide for Fire Protection Systems and Features." As aminimum, surveillance frequencies for fire dampers will be reviewed against the EPRI guidance and updated if necessary.Appreve&.....rt -'m-nte ar.c. freqm..... nPenetrations. Clarification d~~~~Potcors n fu.~ire amc pers are bou ndedb FA8 dto eto .... n .Penetrations. Clarification d~~oorsA 8 and fieFaPer ar1 onddb0FP10 Edto Sections8.2.3.2.1randrs2.NFPA 90A.CCNPP Page A-115FPERAI 04 ConstellaUon Energy Nuclear GroupAttachment L- NFPA 80S Chapter 3 Requirements for ApprovalConselltio Enegy uclar roup Attchmnt -NPA 05 haptr 3Req~reentsforAppovadetection above the ceilings in these areas. However, the NSCA-credited cables thatare routed through these above-ceiling areas are protected by metal conduit. Themetal conduits are not credited to prevent fire induced failure. ( FPE RAI 5C )This variance is technically acceptable based on the following:Based on walkdowns and above-ceiling surveys in these areas, no ignition sourceswere observed above the suspended ceilings except for extension cords which arepotentially susceptible to self-ignition. Exposed wiring above these ceilings wasobserved to be low-voltage communication and data type "network" cables which arenot prone to heat-generating overload faults. No other fixed ignition sources (i.e. fans,fan motors, etc.) were observed above the ceilings.Industry experience has shown that in the unlikely event of a self-ignited cable tray fire,the fire is not expected to spread beyond the cable tray of fire origin. The EPRI fireevents database shows that self-ignitable tray fires have only led to localized failures ina small number of cables within a single raceway. No event has led to sustained openflaming fires, or damage to cables beyond the initially impacted raceway.The extension cords above the ceilings in question are not bundled with cables or othercombustibles, nor are they routed in cable trays. There is even less likelihood that aself-ignited extension cord fire will lead to a sustained open flaming fire, due to a lack ofcombustible material in the vicinity of the extension cords. The only other significantcombustible material observed above the ceilings was ventilation duct wrap insulation.Documentation of this material identifies that the duct wrap insulation has a flamespread rating of less than 25. The duct wrap insulation will therefore not supportsustained combustion or fire growth. In the unlikely event of fire originating in theexposed non-plenum cable, fire will not spread to the duct wrap insulation.Attachment 3 of CNG-FES-007(Reference 6.38) states: uMinimize wiring abovesuspended ceilings. Where installed, electrical wiring shall be listed for plenum use,routed in armored cable, routed in metallic conduit, or routed in cable trays with solidmetal top and bottom covers." This procedure is in place to ensure that futurecompliance with this NFPA 805 requirement is achieved.Per Drawing 60739SH0001 (Reference L-2), ACA ventilation is served by one supplyunit (RTU-1) and two independent exhaust units (access control exhaust fans 11 and12).In the Fire Area 11 portion of the ACA, supply and exhaust registers in the ceiling areducted to and from these units, as shown in Drawing 60597 (Reference L-3). Theabove-ceiling space is therefore not used as an air plenum.On the Turbine Building side of the ACA, supply registers in the ceiling are ducted to thesupply unit, but some exhaust registers in the ceiling are not ducted, as shown inDrawing 60597 (Reference L-3). Exhaust air is pulled from the ceiling plenum into ductsthat lead to the Unit 2 Main Exhaust Plenum where it is exhausted by the main plantexhaust fan 21 or 22. Per Drawings 61085SH00057 (Referecne L-4); 61085SH0058(reference L-5); and 63085SH003B (Reference L-6); the ACA exhaust fans areinterlocked with the Main Plant Exhaust Fans as well as the ACA supply unit RTU-1.Per Drawings 60739SH0001 (Reference L-7) and 60722SH0001 (Reference L-8), theCCP Pag L-CCNPPPage L-9 Constellation Energy Nuclear GroupAttachment L- NFPA 805 Chapter 3 Requiroments for ApprovalApproval Request 3NFPA 805 Section 3.3.1.3.1 states:"A hot work safety procedure shall be developed, implemented, and periodicallyupdated as necessary in accordance with NFPA 51B, Standard for Fire PreventionDuring Welding, Cutting, and Other Hot Work, and NFPA 241, Standard forSafeguarding Construction, Alteration, and Demolition Operations."NFPA 51B, 1999 Edition, Section 3-2 states, in part:"Hot work shall not be allowed in the following areas:(b) In sprinklered buildings while such protection is impaired."CCNPP hot work procedures "ro doo....... in. Pr.ood.u.ro .. SA. 1 100 (Rofr-no L.13). S.A.-4.-00 allows hot work to be performed in plant areas that are protected byIautomatic sprinkler systems while such systems are impaired, contingent on firemarshal or Engineering Programs Unit approval. Administrative procedures 8A--1-4.00is are therefore not in compliance with the hot work requirements of section 3-2(b) ofI EditorialNFPA 51 B. CCNPP requests NRC approval for performance of hot work in sprinkleredbuildings while such systems are impaired as an acceptable variance from therequirements of NFPA 805 Chapter 3 requirements.Basis for Request:While expected to be a very uncommon occurrence, CCNPP anticipates that there maybe occasions where hot work is necessary in sprinklered plant areas while suchsystems are temporarily impaired. Any fire area containing a sprinkler systelm--as FPEidont.ifiod in. ^At..c...ont C, -rablo C- 2) of thie ....iio reo.... , is subject to the provisions RIOof this request. Attachment C, Table C-2 identifies all sprinkler systems in plant fire areas.The administrative procedures that are in place to limit combustibles and control hot workWith the exception of section 3-2(b) of NFPA 5IB, the procedure employed for hot workis a rigorous one and in compliance with the applicable requirements of the 1999Edition of NFPA 51B and the 2000 Edition of NFPA 241 (codes of record). A summaryof the key elements of the procedure is provided below.* A permit application for hot work is submitted to the fire marshal.* The fire marshal assigns a number to the permit, reviews the permit, and conductsand inspection of the area prior to commencing work.* A hot work permit hazard analysis checklist is successfully completed before startingwork.*The Operations group is notified prior to all hot work. This notification is requiredonce per shift.*A hot work fire watch, with the appropriate fire extinguisher for the type and size forthe hazard, is required to be present during the hot work activity and must remain inCCNPP Pge L-1CCNPPPage L-12 Constellation Energy Nuclear GroupAttachment L -NFPA 805 Chapter 3 Requlremente for ApprovalCoselaionrgy Nula ..u.Atcmn.. NP 0 hper3Rq~eetsfrApoathe immediate work area for a minimum of 30 minutes following completion of thehot work activity.* Back-up fire suppression equipment is available in areas where the fire suppressionsystem is inactive.*Combustibles located within 35 feet of the work area are removed prior to hot workoperations. For permanent combustibles that cannot be removed, they must becovre.wthth appro.iit ...... of .. an NFPA 5IB compliant blanket. FPE* Equipment is checked prior to the activity to ensure it is in good working condition.*If hot work is required in an area in which nuclear safety compensatory actions are inplace, completion of a form approved by the system manager, shiftmanager/operations maintenance coordinator, fire marshal, and responsiblemaintenance group supervisor is required per Prcdu... , A ... 102 (Ronfornco ... Editra46) administrative procedures.I*Hot work procedures are in compliance with all other applicable NFPA 51B andNFPA 241 requirements, including those related to management, permit-authorizingindividual, hot work operator training, fire watch (and training) alarm activation, hotwork areas, hot work permits, hot tapping, and fire prevention (precautions regardingcombustibles, inadvertent sprinkler discharge, etc.):T-hese-~Administrative procedures demonstrate the highest standard of care in fire Editodialprevention measures for hot work activities. The rigorous approval, documentation,training, hazard analysis, precautions, lack of combustibles, manual suppression,training, and vigilance ensure that the occurrence of a fire during hot work operations isvery unlikely. The risk of a fire growing uncontrolled beyond the incipient stage due tohot work is therefore not considered a credible scenario.Acceptance Criteria Evaluation:Nuclear Safety and RadIological Release Performance Criteria:Although the hot work requirements in administrative procedures SA---4-00 do not Editorialcomply with section 3-2(b) of NFPA 51iB, there are strict procedures in place to limit thecombustibles, control the hot work within the area, and provide a fire watch to promptlyextinguish any fires that do occur. Therefore, there is no impact on the nuclear safetyperformance criteria.The use of thev..... currant. proc.edurer, administrative procedures to perform hot work Editorialactivities has no impact on the radiological release performance criteria. Theradiological release performance criteria are satisfied based on the determination oflimiting radioactive release (Attachment E), ,.,hio..,h,..;. .,i"" no ..-c " ,.,c ^ b , S rA n 110.. EditorialSafety Margin and Defense-In-Depth:There are procedures in place to limit the combustibles and control the hot workwithin the area. Since fire prevention and manual suppression is maintained perCCNPPPage L-13 Constellation Energy Nuclear GroupAttachment L -NFPA 805 Chapter 3 Requirements for Approvaladministrative procedures ,$A-1.-4.OQ, the safety margin inherent in the analysis Editorialfor the fire event has been preserved.The three echelons of defense-in-depth are:(1) To prevent fires from starting (combustible/hot work controls)(2) Rapidly detect, control and extinguish fires that do occur, thereby limitingdamage (fire detection systems, automatic fire suppression, manual firesuppression, pre-fire plans)(3) Provide adequate level of fire protection for systems and structures so that a firewill not prevent essential safety functions from being performed (fire barriers, firerated cable, success path remains free of fire damage, recovery actions)Per NFPA 805 Section 1.2, defense-in-depth is achieved when an adequate balance ofeach of these elements is provided.Echelons 1, 2, and 3 are met through the limiting of combustibles, control of hot work,and availability of fire watch (i.e., manual suppression), through administrativeprocedures $A-4-4.1.0. The hot work procedures therefore do not compromise manual dorafire suppression functions or post-frie nuclear safety capability. Since a balance of theelements is provided, defense-in-depth is achieved.Conclusion:NRC approval is requested for the allowance of hot work activities in buildings withimpaired sprinkler systems in accordance with administrative procedures $A-4-400, Editorialcontrary to the requirements of Section 3.2(b) of NFPA 5IB, 1999 Edition. Based on theanalysis above, the level of risk encountered by maintaining this current practice isacceptable, and the approach is considered acceptable because it:(A) Satisfies the performance goals, performance objectives, and performancecriteria specified in NFPA 805 related to nuclear safety and radiologicalrelease;(B) Maintains safety margins; and(C) Maintains fire protection defense-in-depth (fire prevention, fire detection, firesuppression, mitigation, and post-frie nuclear safety capability).CCNPP Page L-14CCNPPPage L-14 Constellation Energy Nuclear GroupAttachment I -Definition of Power BlockCoselto nryNceaTru-tahetI-Deiiino oe lcThe term "power block" is clarified in Section K.2 of NEI 04-02 as "structures that haveequipment required for nuclear plant operations, such as containment, auxiliary building,service building, control building, fuel building, radiological waste, water treatment,turbine building, and intake structure, or structures that are identified in the facility'scurrent license basis." The determination of which buildings are required for nuclearplant operations (i.e., required to meet the nuclear safety or radioactive releaseperformance criteria identified in sections 1.5.1 and 1.5.2 of NFPA 805), and thusconsidered within the power block, is identified below.For the purposes of establishing the structures included in the CCNPP Fire Protectionprogram in accordance with 10 CFR 50.48(c) and NFPA 805, the buildings andstructures listed in the following table are considered to be part of the power blockbased on the application of the preceding evaluation criteria.Table I-1 -Power Block DefinitionPower Block Structures Fire Area(s)1, 2, 3, 4, 5, 6,7, 8,9, 10, 11, 12, 13, 14,15, 16, 16A, 16B, 16C, 17, 17A, 17B, 17C,18, 18A, 19, 19A, 20, 21, 22, 23, 24, 25,Auxiliary Building 26, 27, 28, 29, 30, 31, 32, 33, 34, 35, 36,37, 38, 39, 40, 41, 42, 43, 44, AB-1, AB-2,AB-3, AB-4, AB-5, ABFL, KVMAL1A Emergency Diesel Generator DB(EDG) Building ___________________0C Station Blackout (SBO) Diesel DBGenerator Building___________________Reactor Enclosure -Unit No. 1 1CNMTReactor Enclosure -Unit No. 2 2CNMTTurbine Building/ North ServiceBuilding (NSB 12' & 27' Elevations*) TB/NSB/ACA, TBFLIntake Structure IS13.8 kV Switchgear House Unit 113.8 kV Switchgear House Unit 2Condensate Storage Tank No. 12EnclosureYARDFire Protection Pump HouseNo. 2 Fuel Oil Storage Tank No. 21BuildingPretreated-Well Water House* Note: The 45' elevation of the North Service Building is excluded from Fire Area TB/NSB/ACA and the power block as justified by Engineering Equivalency EvaluationECP-1 3-000357. The bases for acceptability are summarized as follows:IFPEP Al 08FPERAI 08CCNPPPage I-2 Constellation Energy Nuclear GroupAttachment I -Definition of Power BlockCoselto nryNcea ru tahetI-Deii-no oe lc-There are no cables or equipment required to achieve nuclear safety performance FPEcriteria (NSPC) in the 45' elevation of the North Service Building. RAI 08-There are no cables or equipment required to achieve NSPC in the Yard within 50 feetof the 45' elevation of the North Service Building.-A fire originating in the 45' elevation of the North Service Building will not impact cablesor equipment required to achieve NSPC in adjacent fire area TBINSB/ACA.CCNPP ageI1-CCNPPPage 1.3 S u ly. eao IIIIIIIIIIIIIIIInfr ai -W th l UI deI 10 C II 2IIII III0Conutlahtlon Energy Nuclear GroupTable C-I -NEI 04-02 Table B-3 Fire Area TransitionWithhold under 10 CFR 2.390CC P Page C,-,,,,l 1, III II ,,,tPage C-SU Security-Related Information -Withhold Under 10 CFR 2.390Aftaahmsn C -NE! 114-2 Tabl B-3 Fire Aere TranuitlonConatelalonl Enemy Nude. GroupTable C-2 -Summary of NFPA 805 Conmpilance Basis and Required Firs Protecton Systems and FeaturesWithhold under 10 CFR 2.390CGNPP Page C.U6Page Security.Related Information -Withhold Under 10 CFR 2.390CitfstnEfuul Nuceamr Group Attachment C -NW1044G2 Table 5.3 Fir Are TrmdoTable C-2 -Suummary of NFPA 805 Compliance Basis and Required Fire Protection Systems and FeaturesWithhold under 10 CFR 2.390CCNPii i ii iii iiiiiiiiiiiiiaiiiiiiiiiiiI IIIII4nliiCCNlq3Pagan C41D7 Security-Related Information. Withhold Under 10 CFR 2.390AtahetC -NE104402 Table B-3 Fire Area TransitionConsisilslon Enugy Nuclear GroupTable C-2 -Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and FeaturesWithhold under 10 CFR 2.390CCNPP Pigs C-USCCNPPPage C.60B Security-Related Information -Withhold Under 10 CFR 2.390Atahmn C -NE1 04-02 Table B54 Fire Area Trminua~Conmtelmtli EnmYw Nueler GrouoI " I I II I II IIIII II IIIITable C-2 -Summary of NFPA 805 Compliance Basis and Required Fire Proeto Systems and FeaturesWithhold under 10 CFR 2.390CCNPP Page C.699CCNPPPage C-6H Constellation EmwW Nuclear Group Attachment E Radioactive Release TransitionNfl 04-02 Radioactive Release TransitionFind Room Pr-Fr PlnNo , n Englnwmlng Controls Training and Pro-FIrm PlanArea Room Descrpt~on and Title In Uquld Gaseous Rmilw Results Conclusonssuch as SA-1-I01 "FireFighting', SA-1 -1 05 "FireBrigade Training" (andassociated fire drills), theODCM, the Radiation SafetyManual, CNG-TR-1 .01 -1 025"Radiation Protection TrainingProgram', RP-2-100Radioactive MaterialsManagement', and RP-2-101"Radioactive Waste" { discusscontaining, monitoring, andreleasing of gaseous effluents.1 CNMT 230E Unit I SA-FFSM-A559, Yes See Room 230 See Room 230 See Room 230 See Room 230Corntainmenl Auxiliary BuildingjpR- Fire Fighting-weo. Strategy ManualICNMT 230N Unit I SA-FFSM-AB69. Yes See Room 230 See Room 230 See Room 230 See Room 230Containment Auxiliary Building..Fire Fighting..aeum- Strategy Manual1CNMT 230S Unit I SA-FFSM-AB69, Yes See Room 230 See Room 230 See Room 230 See Room 230Containment Auwdllary Building--p*- Fire Fighting..Ppuo. Strategy Manual1CNMT 230W Unit 1 SA-FFSM-AB69, Yes See Room 230 See Room 230 See Room 230 See Room 230Containment Auxiliary Building.-f Fire Fighting0.-upi.- Strategy Manual2CNMT 121 Unit 2 SA-FFSM-AB, Yes During Plant Operation: Floor The Containment Ventilation Training materials reinforce Based on the availability ofContainment Auxiliary Building drains are mouted to the System consists of several the use of FFSM's and the engineered controls for bothTendon Gallery Fire Fighting monitored Reactor Coolant subsystems to fulfill the overall Radiation Safety Manual for smoke and fire suppressionCCNIPP Pag E-371 Nudmr GroupA#admmnt E RIMime TranMtkmNE1 0402 Radioactive Release TransitionFire Room Pre.Flrm Plan No. Screeined Engineerng Controls Training and Pro.Fire PlanArs Room Deesrption and Tllo In Liquid Gaseous Rugl Results Conclusionsassociated fire drills), theODCM. the Radiation SafetyManuaml, CNG-TR-1 .01-1025"Radiation Protection TrainingProgram', RP-2-1 00"Radioactive Materialsand RP-2-101-Radioactive WasteManagement" disusscontaining, monitoring, andreleasing of gaseous effluents.2CNMT 229E Unit 2 SA-FFSM-AB69, Yes See Room 229 See Room 229 See Room 229 See Room 229Containment Auxiliary Building(p.- Fire FightingPweee Strategy Manual2CNMT 229N Unit 2 SA-FFSM-AB69, Yes See Room 229 See Room 229 See Room 229 See Room 229Containment Auxiliary Building-4pn." Fire Fighting-.pse. Strategy Manual2CNMT 229S Unit 2 SA-FFSM-AB69, Yes See Room 229 See Room 229 See Room 229 See Room 229Auxiliary Building"Fre FightingS Strategy Manual2CNMT 229W Unit 2 SA-FFSM-AB69, Yes See Room 229 See Room 229 See Room 229 See Room 229Containment Auxiliary Building-t.R.. Fire Fighting.-Pess- Strategy ManualDGBI DB001 Stair Na. 1 N/A No N/A N/A N/A N/A -Screened Out(EL-35)DGB1 DB002 Oil Separator NIA No N/A N/A N/A N/A -Screened OutRoom (EL-35)C G 11111ae111118E.M1 Security-Relatod Information -Withhold Under 10 CFR 2.390Conelellamt. Energy Nuclear Group Afalnn C -NW154-02 Table 843 Fire Area TruneonTable C-I -NEI 04.02 Table B.3 Fire Area TrnanitionWithhold under 10 CFR 2.390CCNPP Peg. V4CCNPPPage C-SS8 Security-Related Information -Withhold Under 10 CFR 2.390Attachment C -NEI 04-02 Table B-3 Fire Area TransitionConstellation Energy Nuclear GroupTable C-I -NE1 04-02 Table B-3 Fire Area TransitionWithhold under 10 CFR 2.390CCNPP Page C-245CCNPPPage C-245 Constellation Energy Nuclear GroupAttachment M -License Condition ChangesCos elltn EnryNula Grop ttchen U -cns odiinChn(3) The licensee shall implement the items listed in Enclosure 1, Attachment 8,Table S-3, "Implementation Items,TM from license amendment request dated ___within 180 days after NRC approval unless that implementation date falls within ascheduled refueling outage. Then, implementation will occur 60 days after startupfrom that scheduled refueling outage. I6.nsert(4) The licensee shall maintain appropriate e=lnsatory measures in place untilcompletion of the modifications listed above.jIt should be noted that implementation item IMP-I12 is associatedImplementation Date jwith incorporation of the NPFA 805 modifications and the RzApd 3, 218completion of'this implementation item is an on-going action RAIApi 0 08Iinitiated within the 180 day timeframe for completion of20implementation items but only complete after completion of 2modification implementation per Table S-2.CCNPP Page M-5CCNPPPage M45 Constellation Enemv Nuclear Grouo4.0 Compliance with NFPA 805 RequirementsC..... lla.i.n....... Nu ea ...... .. 4. o[ln ew t NFP 8 T R,,,,mntThe FPRA model represents the as-built, as-operated and maintained plant as it will beconfigured at the completion of the transition to NFPA 805. The FPRA model includescredit for the planned implementation of modifications identified in Attachment S, TableS-2. Following installation of modifications and the as-built installation details, additionalrefinements surrounding the modifications may need to be incorporated into the FPRAmodel (the FPRA will verity the validity of the reported change-in-risk on as-builtconditions after the modifications are completed). However, these changes are notexpected to be significant. oohr ..,... ..-.go .. r... ou,.....,n. ....r--ott tor ................. in' t ,b , FP.A. no.. Bo ts.mntB Tbo, IM......... ... ..... Insert4.8.3 Supplemental Information -Other LUcensee Specific IssuesNone. CCNPP will verify the validityof the reported change-in-riskestimates of Attachment Wfollowing completion of bothPRA-credited modificationsand implementation items. Ifthis verification determinesthat the risk metrics have P~changed such that the RG RATI1.205 acceptance guidelines 20* .1 20are not met, inc NiKt Will nenotified and additionalanalytical efforts, and/orprocedure changes, and/orplant modifications will bemade to assure the RG 1.205risk acceptance criteria aremet.CCNPP Page 59CCNPPPage 59 Constellation Energy Nuclear Group5.0 Regulatory EvaluationCoselaio TneyNula Gru ".0 R uaToyvautn51.22(c). That evaluation is discussed in Attachment R. The evaluation confirms thatthis LAR meets the criteria set forth in 10 CFR 51 .22(c)(9) for categorical exclusion fromthe need for an environmental impact assessment or statement.5.4 Revision to the UFSARAfter the approval of the LAR, in accordance with 10 CFR 50.71 (e), the CCNPP UFSARwill be revised. The format and content will be consistent with NEI 04-02 FAQ 12-0062.6.5 Transition Implementation ScheduleThe following schedule for transitioning CCNPP to the new fire protection licensingbasis requires NRC approval of the LAR in accordance with the following schedule:Implementation of new NFPA 805 fire protection program to include procedurechanges, process updates, and training to affected plant personnel. This willoccur 180 days following the issuance of an approved SER from the NRC unlessthat date falls within a scheduled refueling outage. Then, implementation willoccur 60 days after startup from that scheduled refueling outage. SeeAttachment S, Table S-3. Insert*Modifications will be complete April 30, 2018. This date assumes SERapproval within two years from uARtbmittal. Appropriate compensatorymeasures will be maintained until modifica'l&#xa2; are complete. See Attachment8, Table S-2. ________________________[it should be noted that implementation Item IMP-12 is associated jwith incorporation of the NPFA 805 modifications and thejcompletion of this implementation item is an on-going action RATIinitiated within the 180 day timeframe for completion of 2 0implementation items but only complete after completion ofmodification implementation per Table S-2.CCNPP Page 66CCNPPPage 66 Constellation Energy Nuclear GroupAttachment U -Internal Events PRA QualityConseaionii iiiiii Enrg Nula Gru tahetU-ItenrEet ultIn accordance with RG 1.205 position 4.3:"The licensee should submit the documentation described in Section 4.2 ofRegulatory Guide 1.200 to address the baseline PRA and application-specificanalyses. For PRA Standard "supporting requirements" important to the NFPA805 risk assessments, the NRC position is that Capability CategoryI/Iis generallyacceptable. Licensees should justify use of Capability Category I for specificsupporting requirements in their NFPA 805 risk assessments, if they contend thatit is adequate for the application. Licensees should also evaluate whetherportions of the PRA need to meet Capability Category/Ill, as described in thePRA Standard."The PMVROG performed a full scope internal events PRA peer review of CCNPP todetermine compliance with ASME PRA Standard, }}oand RG 1.2001P(Reference 6.32) in June 2010. This review documented find Igs for all supportingI 21,requirements (SRs) which failed to meet at least Category II. Th findings for that peer 2review are documented below in Table U-i. This table also in ,ludes the disposition,status, and impact on the FPRA.The peer review found that 97% of the SR's evaluated Met Ca ability Category II orbetter. There were 3 SR's that were noted as "not met" and 8 there were noted asCategory 1. As noted in the peer review report the majority }f the findings weredocumentation related. Of the 11 SR's which did not meet Categ y 2 or better, 7 wererelated to conservatisms or documentation in LERF and 2 related to internalfloods. There were 39 findings. All findings which could be rel vant to FPRA wereupdated in the internal events model used to quantify the FPR .Thus, with theexception of minor documentation concerns, the internal events mo el meets CategoryII or causes conservative results for all SR's relevant to FPRA res Ilts. No significantchanges have been implemented in the internal events PRA. As there are no newmethods applied, no follow on or focused peer reviews were requirec1RA-Sa-2009(Reference 6.31)CCNIP PaIeIUICCNPPPage U-2 ConsteISuliofi bneIgy Nucber GroupSecurity.Related Information -Withhold Under 10 CFR 2.390Attachment S -IModiIicaions and Implameotatlon ItemsTabl S- Impleenaton tm~sWithhold under 10 CFR 2.390CCNPP Pug. S-IT Constellation Energy Nuclear GroupAttachment O -Orders and ExemptionCoselto Eng Nula rou Atcmn0-OdradEepTioExemptionsRescind the following exemptions granted against 10 CFR 50, Appendix R dated asfollows:* Auguo 16 105 AnY~ a I from thcy rgiromonte~ ofV Sootyion lll.G.2 ofthe v-e'effpti.. >, Deleted consistent with Attachment K* April 21, 1983 -An exemption from the requirements of Section lll.G.3 ofAppendix R for the control room complex and the intake structure related to theinstallation of fixed fire suppression systems.* March 15, 1984 -An exemption from the requirements of Section lII.G to allowalternatives to the 3-hour rated fire barriers for areas listed in the exemption. Anexemption was also granted for Section Ill.G for Fire Areas 10 and 11 related tothe installation of fixed fire suppression systems. Additionally, an exemption fromthe requirements of Section III.0 was granted regarding the capacity of the oilcollection systems for the reactor coolant pumps.* August 22, 1990 -An exemption from the requirements of Section lll.J to allowthe use of portable hand lights as an alternative to permanently installed 8-houremergency lights in the Unit 1 and 2 containment buildings.* April 7, 1999 -An exemption from the requirements for Section III.J to allow theuse of security lighting in exterior areas, the use of portable lights in highradiation areas and the use of helmet mounted lights inside of switchgearcabinets as alternatives to permanently installed 8-hour emergency lights.Specific details regarding these exemptions are contained in Attachment K.OrdersNo Orders need to be superseded or revised.CCNPP implemented the following process for making this determination:-A review was conducted of the CCNPP docketed correspondence. The reviewwas performed by reviewing the correspondence files and performing electronicsearches of internal CCNPP records and the NRC's ADAMS document system.A specific review was performed of the license amendment that incorporated themitigation strategies required by Section B.5.b of Commission Order EA-02-026 toensure that any changes being made to ensure compliance with 10 CFR 50.48(c) donot invalidate existing obligations applicable to the plant. The review of this orderdemonstrated that changes to the fire protection program will not affect measuresrequired by B.5.b.The Fukushima Orders are being independently evaluated. Any plant changes will beevaluated for impact on the fire protection program in accordance with the CCNPPdesign change process.CCNPP Page 0-2CCNPPPage 0-2 Security-Related Information -Withhold Under 10 CFR 2.390Attachment S -Modifications and Implementation ItemsConstellation Enemy Nuclear GIOuD.... ..i lllllllll IIII I III IIIIII II III II I III ITable S-3 Implementation ItemsWithhold under 10 CFR 2.390CCNPP Page S-15CCNPPPage 5-15}}

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