Text

Nine Mile Point, Unit 1, License Amendment Request Pursuant to 10 CFR 50.90: Adoption of NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition) - Response to NRC Rai. Part
	ML13127A397
Person / Time
Site:	Nine Mile Point
Issue date:	04/30/2013
From:	Costanzo C R; Constellation Energy Group, EDF Group, Nine Mile Point
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	TAC ME8899
	Download: ML13127A397 (238)
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ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATIONREGARDING THE PROPOSED ADOPTION OF NFPA 805(2001 EDITION)Nine Mile Point Nuclear Station, LLCApril 30, 2013 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)By letter dated June 11, 2012, Nine Mile Point Nuclear Station, LLC (NMPNS) requested an amendmentto the Nine Mile Point Unit 1 (NMP1) Renewed Facility Operating License DPR-63. The proposedamendment would adopt a new risk-informed performance-based (RI-PB) fire protection licensing basiswhich complies with the requirements in 10 CFR 50.48(a) and 10 CFR 50.48(c); the guidance inRegulatory Guide (RG) 1.205, "Risk-Informed Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," Revision 1; and National Fire Protection Association (NFPA) 805,"Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants,"2001 Edition.This enclosure provides supplemental information in response to the following three requests foradditional information (RAIs) documented in the NRC's letter dated January 3, 2013: Safe Shutdown /Circuit Analysis RAI 01, RAI 03, and RAI 07. NMPNS agreed to provide responses to these three RAIsby April 30, 2013. Each individual NRC RAI is repeated (in italics), followed by the NMPNS response.Safe Shutdown / Circuit Analysis RAI 01The description in LAR Section 4.2.1.2 of safe and stable is defined as, "the ability to maintain Keff < 0. 99with a reactor coolant temperature at or below the requirement for hot shutdown and then subsequentlycool down and maintain NMP1 in a cold shutdown condition. " The nuclear safety capability assessment(NSCA) methodology review (LAR Attachment B) includes discussion of cold shutdown (CSD)methodology as appropriate and the methods for meeting performance goals in the fire area assessments(LAR Attachment C) include CSD components and systems.Additional information is needed regarding the timing, systems, actions, and any repairs, necessary toachieve and maintain CSD. There is no discussion of the risk associated with actions to achieve andmaintain CSD.VFDRs are identified in LAR Attachment Cforperfornance criteria related to CSD. In some cases, theseVFDRs are dispositioned on the basis that the risk, defense-in-depth (DID), and safety margins meet theacceptance criteria of NFPA 805 with a recovery action (RA) credited. The VFDR disposition furtherstates the RA has been evaluated for feasibility and reliability within the FPRA using HRA methods (e.g.,Attachment C, pg. 64, VFDR-05-025).Additional information is needed to address the following specific issues:a. Provide the timing assumed for sustaining hot shutdown (once achieved) and then transitioningfrom hot shutdown to, and achieving CSD.b. Describe how cold shutdown was modeled in the FPRA, including the risk of RAs credited fordisposition of VFDRs associated with CSD NSCA equipment.c. System or component capacity limitations are not specifically described for each applicableperformance goal. Provide a description of capacity limitations, need to replenish systems, andtime-critical actions for other systems needed to maintain safe and stable conditions (e.g.,nitrogen supply for valve operations, water supplies, boron supply, DC battery power, fuel, etc.).1 of 47 ENCLOSURE1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)d. Describe in more detail the resource (staffing) requirements, timing, and feasibility of operatoractions to recover NSCA equipment to achieve and sustain safe and stable conditions.e. Attachment G describes actions involving repairs to valve and pump wiring for shutdowncooling. Describe in more detail the resource (staffing) requirements, timing, and feasibility ofactions to repair NSCA equipment to achieve and maintain CSD safe and stable conditions.f Provide a more detailed description of the risk of failure of operator actions and equipmentnecessary to sustain safe and stable conditions.g. Describe the actions that are planned for MSOs for shutdown cooling or any time the need torestore decay heat removal is short based on time to boil.Response to Safe Shutdown / Circuit Analysis RAI 01GeneralNine Mile Point Nuclear Station, LLC (NMPNS) has elected to modify its NFPA 805 transition analysisfor NMP1 to revise the approach for demonstrating the ability to reach and maintain safe and stableconditions, as specified by NFPA 805. The original Nuclear Safety Capability Assessment (NSCA)established as its basis for demonstrating safe and stable conditions the requirement to maintain K~f <0.99 with a reactor coolant temperature at or below the requirements for hot shutdown and thensubsequently cool down and maintain the plant in a cold shutdown condition. Consistent with NFPA 805and supplemental guidance, NMPNS is revising its basis for the NMP1 NSCA to include only therequirement to establish hot shutdown conditions, including long-term hot shutdown capability. Theresponse to this RAI, including Parts a through g, is provided within the context of the aforementionedchange.Demonstration of the nuclear safety performance criteria for safe and stable conditions is performed intwo analyses based on the plant operating modes, as defined in the NMP1 Technical Specifications (TS).These analyses are defined as follows:* At-Power analysis for potential fires while in either: (i) the Power Operating Condition (Reactormode switch is in "Startup" or "Run" position and the reactor is critical or criticality is possible due tocontrol rod withdrawal), or (ii) the Shutdown Condition -Hot operating condition (Reactor modeswitch is in "Shutdown" position and reactor coolant temperature is greater than 212'F), with theShutdown Cooling (SDC) system not aligned for decay heat removal.* Non-Power analysis for potential fires while in Shutdown Condition -Hot operating condition andlower operating conditions.A copy of TS Section 1.1 containing the definitions of the NMP I reactor operating conditions is providedin Figure SSD/CA RAI 01-1 below to facilitate a clear understanding of the analytical coverage providedby the two analyses described above.2 of 47 ENCLOSURE1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)1.0 DEFINITION4S1.1 Reactor Operatina ConditionsThe various reactor operating conditions are defined below. Individual technical specifications amplify these definitions whenappropriate.a- Shutdown Condition -Cold(1) The reactor mode switch is in the shutdown position or refuel position.(2) No core alterations leading to an addition of reactivity are being performed.(3) Reactor coolant temperature is less than or equal to 2121F.b. Shutdown Condition -Hot(1) The reactor mode switch is in the shutdown position. **(2) No core alterations leading to an addition of reactivity are being performed.(3) Reactor coolant temperature is greater than 212'F.c. Refueling Condition(1) The reactor mode switch is in the refuel position.(2) The reactor coolant temperature is less than 212°F.(3) Fuel may be loaded or unloaded.(4) No more than one operable control rod may be withdrawn.d. Prwer Operating Condition(1) Reactor mode switch is in startup or run position.(2) Reactor is critical or criticality is possible due to control rod withdrawal.e. Major Maintenance Condition(1) No fuel is in the reactor.The reactor mode switch may be placed in the startup position to perform the shutdown margin demonstration. See Special TestException 3.7.1.The reactor mode switch may be placed in the refuel position to perform reactor coolant system pressure testing, control rod scramtime testing and scram recovery operations.Figure SSD/CA RAI 01-1: NMP1 Technical Specification Section 1.1, Definitions for ReactorOperating ConditionsThe practical manifestation of the redefined basis for safe and stable is that the At-Power analysis nowincludes only equipment necessary to achieve and maintain hot shutdown conditions, including some newequipment required to demonstrate long-term hot shutdown capability. The NSCA no longer requires theability to achieve and maintain cold shutdown. On this basis, equipment associated with the SDC systemand any associated variances from deterministic requirements (VFDRs) of NFPA 805 Section 4.2.3 havebeen removed from the NSCA. Cold shutdown issues are now addressed only within the context of theNon-Power Operations (NPO) analysis, and only to the extent that they apply (refer to the response forSafe Shutdown / Circuit Analysis RAI 03). A formal screening process based on the criteria shown inFigure S SD/CA RAI 01-2 was used to screen and identify VFDRs associated only with the SDC system;i.e., cold shutdown only VFDRs.3 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)1 If the credited HSD path is "C" or"D," as defined in EIR 51-9133191 (NMP1 NSCA), all VFDRsrelevant to the fire area are required for long term "Safe and Stable" operation. Shutdown paths "C"and "D" utilize decay heat removal capabilities independent of the shutdown cooling system, andrequire the use of Core Spray (CS), Emergency Relief Valves (ERVs), and Containment Spray(CTS)/Containment Spray Raw Water (CTSRW) systems to assure appropriate HSD conditions.2. The following criteria apply to the screening of VFDRs where the pertinent fire area HSD path is "A"or "B," as defined in EIR 51-9133191 (NMP1 NSCA). Shutdown paths "A" and "B" require the use ofEmergency Condenser Cooling (supports HSD) and Shutdown Cooling (supports CSD). Thus,VFDRs may be removed if they are associated with only CSD operation." VFDRs associated with systems or components required for initial plant inventory or pressurecontrol are required for the "At-Power analysis." These systems include Emergency CondenserCooling (EC), Main Feedwater Isolation, spurious ERV actuation, and Reactor Water Cleanup(RWCU) isolation. Note that the Control Rod Drive (CRD) System is required to ensure adequateReactor Pressure Vessel (RPV) makeup is available to account for nominal inventory losses overtime, thus VFDRS associated with availability of the system are required for the "At-poweranalysis." VFDRs associated with systems or components necessary to support vital plantdiagnostic indication are required for the "At-Power analysis." These systems include ReactorCoolant System (RCS) pressure and level indication and torus level indication. Torus levelindication is required because increasing torus level may reduce the available margin to transitionto CSD and require operators to depressurize earlier in the event, thereby potentially jeopardizingthe ability to maintain "stable" conditions while in HSD. Thus, scenarios that could impact theavailability of torus level are required for the "At-Power analysis," such as spurious CTSRWInjection)." VFDRs associated with systems or components associated with Reactor Building Closed LoopCooling, Emergency SW, or Normal SW availability are required for the "At-Power analysis."These systems are necessary to provide control room cooling, and thus must remain availableprior to the transition to CSD." VFDRs associated with electrical power availability are generally required for the "At-Poweranalysis" (e.g. Loss of Train A/B battery charging capability). Loss of individual power suppliesare considered on a case-by-case basis, and are binned in accordance with the function of theequipment supported by the power supply." VFDRs associated with the availability of HVAC units are generally required for the "At-Poweranalysis" unless the affected equipment/component supports a purely CSD function." VFDRs associated with availability of the Shutdown Cooling System are required for CSD/NPO.Similarly VFDRs associated with Torus cooling are required for CSD/NPO. Torus cooling is notrequired for the "At-Power analysis" because EC cooling is the primary method for decay heatremoval, and the system is not adversely impacted by a loss of torus cooling.Figure SSD/CA RAI 01-2: Criteria for Screening and Identifying Cold Shutdown VFDRs4 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Changes to the NSCA and License Amendment Request (LAR) documentation necessary to implementthe newly established safe and stable analysis basis include:* Elimination of SDC system components from the NSCA equipment list. This change resulted inelimination of 73 VFDRs associated with cold shutdown (see Table SSD/CA RAI 01-1 for a list ofremoved cold shutdown VFDRs).* Include in the NSCA the fire water system valves associated with refilling the Emergency Condensermakeup tanks to support long-term decay heat removal capability for hot shutdown operation. Thischange resulted in the addition of 22 new VFDRs (see Table SSD/CA RAI 01-2 for a list of newVFDRs).* Addition of two new recovery actions associated with manual valve alignment and operation of theDFP to refill the Emergency Condenser makeup tanks to support long-term operation of theEmergency Cooling (EC) system, to satisfy decay heat removal requirements for hot shutdown." Re-quantification of the Fire PRA model and re-calculation of ACDF and ALERF. In both cases,slight improvements in risk were realized as a result of the changes (refer to LAR Attachments C andW for specific ACDF and ALERF values). Although delta risk decreased overall, there were slightincreases in the contributions from recovery actions for ACDF and ALERF. This small increase isattributable to the reduction in reliance on SDC system components and increased reliance on theDFP and fire system valves for demonstrating that safe and stable conditions are achieved andmaintained. The net effect makes the post-transition plant more like the deterministically compliantplant in terms of risk. The new approach to demonstrating safe and stable conditions results in a slightincrease in reliance on recovery actions due to the addition of the two new recovery actions notedabove.* Updates to the following LAR Transition Report sections and attachments:-Update Section 4.2 to incorporate the new basis for safe and stable, including discussion on long-term maintenance of hot shutdown conditions (see Enclosures 3 and 4).-Update Table 4-3 to capture summary-level changes to the analysis (see Enclosures 3 and 4).-Update Attachment A (Table B-i), Section 3.5.16, to address the new time frame for alternate useof the DFP to refill the Emergency Condenser makeup tanks (see Enclosures 3 and 4).-Update Attachment B (Table B-2) to address the revised methodology for achieving andmaintaining safe and stable conditions (see Enclosures 3 and 4).-Update Attachment C (Table B-3) to remove cold shutdown VFDRs, add new VFDRs associatedwith the DFP, and update the fire risk summary results (see Enclosures 5 and 6).-Update Attachment G to add new recovery actions associated with manual alignment andoperation of the DFP to refill the Emergency Condenser makeup tanks (see Enclosures 3 and 4).5 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)-Update the Attachment W Fire PRA insights and results to reflect re-quantification of the FirePRA model and re-calculation of ACDF and ALERF values with the VFDR changes considered(see Enclosures 5 and 6).Specific responses to Parts a through g of this RAI are provided below and are based on the revised safeand stable analysis basis. Thus, in some cases, the questions pertaining only to cold shutdown are nolonger relevant.Part aSustaining hot shutdown conditions (once achieved) for an extended period of time is accomplishedby (1) ensuring a continual source of water to the Emergency Condensers in support of decay heatremoval using the EC system, (2) ensuring a long-term source of inventory for makeup to the reactor,and (3) ensuring continual operation of at least one emergency diesel generator to supply AC powerto the electrical distribution system.Upon achieving hot shutdown conditions, the plant is able to maintain safe and stable operation for anextended period of time using the EC system. After 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , the Emergency Condenser makeup tankscan be replenished as needed using the DFP, which draws water from Lake Ontario (effectively aninfinite source). In the event water from the condensate storage tanks (CST) can be transferred to theEmergency Condenser makeup tanks, operation of the DFP would not be required until some pointbeyond 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Periodic refueling of the DFP is accomplished in accordance with existing plantprocedures using the DFP fuel oil storage tank. The DFP day tank contains sufficient fuel for 17hours of operation. The DFP fuel oil storage tank contains fuel to support 6.1 days of operation.Reactor coolant makeup is required after 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , assuming a nominal TS leakage rate of 25 gpm.Makeup is provided via the Control Rod Drive (CRD) system using one of the CRD pumps drawingsuction from the CST. Alternating current (AC) power is required to operate a CRD pump. The DFPmay be aligned to provide reactor coolant makeup in the event that no CRD pump is available.In the event the EC system is not available, the plant can be maintained in hot shutdown by openingthree Electromatic Relief Valves (ERVs) in the automatic depressurization system (ADS) andblowing steam to the Torus to reduce pressure. When reactor pressure reaches approximately 365psig, the Core Spray (CS) system may be utilized to provide core cooling. Both AC and direct current(DC) electrical power are required for this method of decay heat removal. This alternate means ofdecay heat removal can be used to maintain safe and stable conditions until such time that the SDCsystem is placed in service. The CS system is a two loop system. Operation of one loop is adequate toensure core cooling. When utilizing the CS system, the ERVs pass steam and then, eventually, waterto the Torus to remove decay heat from the reactor, in essence placing the Reactor Coolant System(RCS) in recirculation through the Torus. During this process, decay heat is removed by operation ofthe Containment Spray (CTS) system in conjunction with the Containment Spray Raw Water(CTSRW) system. This method of decay heat removal negates the need for another system to provideinventory makeup. AC power is required to initiate and maintain this method of decay heat removal;thus, long-term maintenance of this operating mode is dependent on maintaining AC electrical power.For either of the hot shutdown methods used to achieve and maintain long-term safe and stableconditions, AC power availability from either the station Emergency Diesel Generators (EDGs) oroffsite power is necessary. Offsite power is not credited in the NSCA. The EDGs can be refueled in6 of 47 ENCLOSURE1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)accordance with existing plant procedures using an on-site fuel source (tanker truck), until such timethat offsite power is restored. Refueling of a continually operating EDG is estimated to be requiredafter four days (assuming one EDG operating at full load). Given the long timeframe before EDGand/or DFP refueling is necessary, additional resources from the emergency response organizationwill be available to support EDG and DFP refueling activities.Transition to cold shutdown is no longer a requirement in the NSCA for ensuring that safe and stableconditions are achieved and maintained.Part bSuccess in the Fire PRA is defined to be a controlled stable state with the reactor subcritical, its waterinventory stable, and its heat being removed. The Fire PRA success criteria require that this stablecondition be maintained for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; i.e., cold shutdown is not required for success in the Fire PRA.Nevertheless, some equipment that may be used to establish cold shutdown conditions is modeled inthe Fire PRA; e.g., the SDC system is modeled in the Fire PRA to provide a backup to the loss ofother heat removal systems.The definition of safe and stable conditions for at-power analysis in LAR Section 4.2.1.2 is nowrevised to require achieving and maintaining hot shutdown conditions for an extended period of time,rather than to require achieving hot shutdown conditions and transitioning to cold shutdown. Some ofthe VFDRs originally presented in the LAR were identified as VFDRs only because they presented achallenge in meeting the nuclear safety performance criteria associated with achieving cold shutdown.Due to the elimination of the requirement for cold shutdown from the definition of safe and stableconditions, the cold-shutdown VFDRs have been eliminated (see listing in Table SSD/CA RAI 01-1).Thus, it is no longer necessary to evaluate the change in risk (ACDF and ALERF), including the riskof recovery actions, associated with cold-shutdown VFDRs.Part cThe NMP1 NSCA (EIR 51-9133191) was developed in accordance with the NFPA 805 requirementsand applicable Frequently Asked Questions (FAQs). Section 1.5.1 of NFPA 805 identifies thepertinent nuclear safety performance criteria that are to be satisfied in order to "provide reasonableassurance that, in the event of a fire, the plant is not placed in an unrecoverable condition." Thecriteria are:* Reactivity Control* Inventory and Pressure Control* Decay Heat Removal" Vital Auxiliaries* Process MonitoringLAR Attachment C (Table B-3) documents how each performance criteria is satisfied on a fire areabasis. When applicable, VFDRs are identified for each performance goal in Table B-3 and adisposition is provided. The revised NMP1 basis for safe and stable requires that hot shutdownconditions be achieved and maintained for an extended period of time, which introduces additional7 of 47 ENCLOSURE1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)system capacity limitations to ensure that the nuclear safety performance criteria are satisfied. Theadditional capacity limitations include:* A source of inventory makeup to the Emergency Condensers to ensure availability of the ECsystem for extended hot shutdown operation.* A source of inventory makeup to the reactor to compensate for primary system leakage over anextended time frame. Reactor inventory makeup was addressed originally as a cold shutdownconsideration, whereas the revised NMPI basis for safe and stable could require that inventorymakeup be established prior to the cold shutdown transition.* A source of fuel oil to assure long-term availability of the EDGs and DFP.The response to Part a of this RAI provides details for these three identified long-term considerations.Additional VFDRs (applicable to the pertinent fire areas) have been added to LAR Attachment C(Table B-3) to address deterministic concerns regarding availability of the Fire Water System.Specifically, local start of the DFP is addressed by plant procedure N1-OP-21A. The procedure alsoaddress actions necessary to replenish fuel oil to the DFP, thereby ensuring adequate fuel oil supply,and other actions necessary to ensure pump operation if instrument air is not available. Procedure N1-PM-V 19 provides guidance to replenish the EDG fuel oil storage tank to assure continued long-termoperation of the EDGs.Time-critical actions needed to support capacity limitations involve load shedding for Battery Boards11 and 12 to ensure DC power availability. These actions are included in Table G-1 of LARAttachment G.Part dRecovery actions credited in the NFPA 805 transition to bring the plant to and maintain it in a safeand stable condition (i.e., hot shutdown) fall into one of two categories, as follows:" Recovery actions modeled in the Fire PRA and analyzed as part of the human reliability analysis(HRA) of the Fire PRA. Appendix I of the Human Reliability Analysis (HRA) Fire PRAnotebook (Ni-HRA-FO001) provides feasibility evaluations for the recovery actions modeled inthe Fire PRA that are used to resolve VFDRs identified in the NSCA. The feasibility evaluationsare performed separately for each of the 17 recovery actions identified for analysis consideringthe 11 criteria from FAQ 07-0030 (demonstrations, systems and indications, communications,emergency lighting, tools, procedures, staffing, actions in the fire area, time, training, and drills).* Recovery actions not modeled in the Fire PRA and whose additional risk was found to beinsignificant based on a qualitative evaluation. The feasibilities of these recovery actions areevaluated in EIR 51-9156521.The recovery action feasibilities were evaluated using the 11 criteria given in FAQ 07-0030, whichinclude, among others, staffing and timing requirements. All recovery actions credited in the NFPA805 transition were found to meet the feasibility criteria of FAQ 07-0030. Operator impacts andstaffing considerations for the long-term safe and stable actions have been included in the updated8 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)analysis. Actions required to achieve and initially maintain hot shutdown conditions can be performedby the minimum shift complement of reactor operators, senior reactor operators, and non-licensedplant operators. As discussed in the response to Part a of the RAI, EDG and DFP refueling activitiesdo not occur for several days, are proceduralized, and can be implemented by emergency responseorganization personnel.Operator actions modeled in the Fire PRA (including those that are not recovery actions) wereevaluated for their feasibility and reliability as part of the development of their human errorprobability (HEP). This evaluation was documented in N1-HRA-F001 and relied on guidance fromNUREG/CR-6850/EPRI TR-1011989, NUREG-1792, NUREG-1852, and the ASME/ANS PRAstandard.The recovery actions that were previously credited in the fire risk evaluations to reduce the riskcontribution from cold shutdown VFDRs are removed. These recovery actions are no longer requiredto demonstrate that safe and stable conditions can be maintained. Recovery actions for the SDCsystem that are included in the Fire PRA as a means of reducing risk are retained.Part eThe recovery actions involving local repairs to valve and pump wiring for the SDC system have beenremoved from LAR Attachment G. These recovery actions are associated with VFDRs pertaining tocold shutdown activities and are no longer required to demonstrate that safe and stable conditions(redefined as hot shutdown) can be maintained. The VFDRs associated with these repair actions havebeen removed from the NSCA.Part fThe risk of failure of operator actions and equipment necessary to sustain safe and stable conditions isevaluated in the models developed for the Fire PRA, since safe and stable conditions have beenredefined as hot shutdown and the Fire PRA covers hot shutdown conditions.Changes to the Fire PRA to implement the newly established safe and stable basis for the At-Poweranalysis involve credit for the DFP and two new operator recovery actions. The recovery actions areassociated with manual valve alignment and local operation of the DFP to refill the EmergencyCondenser makeup tanks in support of long-term operation of the EC system to satisfy decay heatremoval requirements for hot shutdown.The Fire PRA models are quantified to determine the fire-induced core damage frequency (CDF) andlarge early release frequency (LERF). These risk metrics are used in the fire risk evaluations (FREs),consistent with Regulatory Guides 1.205 and 1.174.The Fire PRA also supports the FREs by ensuring that the risk inherent to each fire area is properlycaptured and that the set of recovery actions credited for the NFPA 805 transition is appropriatelycharacterized, including the evaluation of their additional risk. LAR Attachment G is thereforeamended to address the removal of SDC recovery actions and the addition of the DFP and associatedvalve-related recovery actions.9 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Updates to the LAR Attachment W Fire PRA insights and results reflect the Fire PRA re-quantification values generated for the revised safe and stable basis.Part gMultiple Spurious Operations (MSOs) associated with the SDC system are not addressed in theNSCA per the revised NMP1 basis for safe and stable operation. As a result, VFDRs originallyassociated with the SDC system have been removed from the NSCA analysis.The NPO analysis (documented in EIR 51-9137629 and 51-9171174) identifies equipment that mustremain functional to satisfy a particular Key Safety Function (KSF) success path. These KSF successpaths were developed in accordance with the guidance in FAQ-07-0040, wherein higher riskevolutions (HRE) drive the selection of KSF success paths (including decay heat removal) based ontime to boil. The NPO analysis provides recommendations to best manage fire risk for "pinch points"(areas of the plant where complete loss of a KSF may occur due to fire).A number of MSO scenarios associated with the SDC system are identified in the NMP1 ExpertPanel MSO Report, "Technical Report on Identification & Classification of the NMP-l MSOScenarios using an Expert Panel -Review of New Generic Scenarios," dated May 2012. Thesescenarios are all addressed within the context of the SDC KSFs (1DHR-RX-SDC), as documented inthe NMP1 NPO KSF Equipment List (EIR 51-9137629). There is one KSF identified for each train ofthe SDC system. Accordingly, pinch points associated with the availability of these KSFs areidentified in the NMP1 NPO Component Pinch Point Analysis (EIR 51-9171174). Recommendationsto best manage fire risk for each scenario pinch point are also described in EIR 51-9171174(Appendix B) and are summarized in the response to Safe Shutdown / Circuit Analysis RAI 03 (seeTable SSD/CA RAI 03-2).The response to Safe Shutdown / Circuit Analysis RAI 03 addresses specific aspects of the NMP1NPO analysis, including the treatment of MSOs.10 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 01-1: Cold Shutdown VFDRs Eliminated from Table B-3Fire AreaFire Area VFDR ID VFDR Title Details CommentsDescription04 Foam Room, el. VFDR-04-005 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of261 instrument air. BV-70-53 may fail closed on loss of the Shutdown Cooling System. The Shutdowninstrument air. Valve is required open for cold shutdown to Cooling System is only required to supportsupport decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminatedrequired to open valve BV-70-53. (PC031) from the NSCA and will be addressed in theNPO analysis.04 Foam Room, el. VFDR-04-006 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of261 instrument air. FCV-38-09 may fail closed on loss of the Shutdown Cooling System. The Shutdowninstrument air. Valve is required open for cold shutdown to Cooling System is only required to supportsupport decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminatedrequired to open valve FCV-38-09. (PC031, OMC0O1) from the NSCA and will be addressed in theNPO analysis.04 Foam Room, el. VFDR-04-007 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of261 instrument air. FCV-38-10 may fail closed on loss of the Shutdown Cooling System. The Shutdowninstrument air. Valve is required open for cold shutdown to Cooling System is only required to supportsupport decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminatedrequired to open valve FCV-38-10. (PC031, OMC001) from the NSCA and will be addressed in theNPO analysis.04 Foam Room, el. VFDR-04-008 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of261 instrument air. FCV-38-11 may fail closed on loss of the Shutdown Cooling System. The Shutdowninstrument air. Valve is required open for cold shutdown to Cooling System is only required to supportsupport decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminatedrequired to open valve FCV-38-11. (PC031,OMC001) from the NSCA and will be addressed in theNPO analysis.05 Turbine VFDR-05-020 Loss of Instrument Air A deterministic assumption assumes potential loss of This VFDR is associated only with operation ofBuilding, el. 240 instrument air. RBCLC to SDC Flow control valve BV-70-53 the Shutdown Cooling System. The Shutdownto 369 may fail closed on loss of instrument air. Valve is required Cooling System is only required to supportopen for cold shutdown to support decay heat removal. A CSD. On this basis, the VFDR is eliminatedRecovery Action may be required to open valve BV-70-53 from the NSCA and will be addressed in the(PC031) NPO analysis.05 Turbine VFDR-05-021 Loss of Instrument Air A deterministic assumption assumes for potential loss of This VFDR is associated only with operation ofBuilding, el. 240 instrument air. SDC heat exchanger outlet flow control the Shutdown Cooling System. The Shutdownto 369 valve FCV-38-10 may fail closed on loss of instrument air. Cooling System is only required to supportValve is required open for cold shutdown (PC031) CSD. On this basis, the VFDR is eliminatedfrom the NSCA and will be addressed in theNPO analysis.05 Turbine VFDR-0S-025 Spurious Operation of A separation concern exists for postulated fire in this area This VFDR is associated only with operation ofBuilding, el. 240 Shutdown Cooling for the loss of PB-167 due to an uncoordinated associated the Shutdown Cooling System. The Shutdownto 369 System Valves IV-38-01 emergency lighting circuit. Non SSD Emergency Lighting 11 Cooling System is only required to supportand IV-38-13 is supplied by cable 167-101 from 600V Power Board 167, CSD. On this basis, the VFOR is eliminatedBreaker H01. The emergency lighting power cable 167-101 from the NSCA and will be addressed in thesupply breaker H01 does not coordinate with PB-167 supply NPO analysis.breaker. Fire damage to cable 167-101 could cause the lossof PB-167 preventing operation of Reactor ShutdownCooling Isolation Valves, IV-38-01 and IV-38-13 as directedin repair procedure N1-DRP-005. ( OP019)05 Turbine VFDR-05-044 Unavailability of A postulated fire in this area may damage cable 171-41 This VFDR is associated only with operation ofBuilding, el. 240 Shutdown Cooling adversely affecting credited SDC valve BV-38-04. SDC is the Shutdown Cooling System. The Shutdownto 369 Valve BV-38-04 required to support the decay heat removal function. Local- Cooling System is only required to supportManual operation of BV-38-04 per N1-SOP-21.1 may be CSD. On this basis, the VFDR is eliminatedrequired. (PC031) from the NSCA and will be addressed in theNPO analysis.05 Turbine VFDR-05-045 Unavailability of A postulated fire in this area may damage cable 12DV-10, This VFDR is associated only with operation ofBuilding, el. 240 Shutdown Cooling 12DV-11, 12DV-29, 12DV-9 or 167-11 adversely affecting the Shutdown Cooling System. The Shutdownto 369 Valve IV-38-02 credited SDC valve IV-38-02. SDC is required to support the Cooling System is only required to supportdecay heat removal function. Local-Manual operation of IV- CSD. On this basis, the VFDR is eliminated38-02 per N1-SOP-21.1 may be required. (PC031) from the NSCA and will be addressed in theNPO analysis.11 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 01-1: Cold Shutdown VFDRs Eliminated from Table B-3Fire AreaFire Area VFDR ID VFDR Title Details CommentsDescription06 Turbine Building VFDR-06-013 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofNorth, el. 250 instrument air. Flow control valve BV-70-53 may fail closed the Shutdown Cooling System. The Shutdownon loss of instrument air. Valve is required open for cold Cooling System is only required to supportshutdown to support decay heat removal. A Recovery CSD. On this basis, the VFDR is eliminatedAction may be required to open valve BV-70-53. (PC031) from the NSCA and will be addressed in theNPO analysis.06 Turbine Building VFDR-06-014 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofNorth, el. 250 instrument air. SDC heat exchanger outlet flow control the Shutdown Cooling System. The Shutdownvalve FCV-38-11 may fail closed on loss of instrument air. Cooling System is only required to supportValve is required open for cold shutdown to support decay CSD. On this basis, the VFDR is eliminatedheat removal. A Recovery Action may be required to open from the NSCA and will be addressed in thevalve FCV-38-11. (OMC001,PCO31) NPO analysis.06 Turbine Building VFDR-06-015 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofNorth, el. 250 instrument air. SDC heat exchanger outlet flow control the Shutdown Cooling System. The Shutdownvalve FCV-38-09 may fail closed on loss of instrument air. Cooling System is only required to supportValve is required open for cold shutdown to support decay CSD. On this basis, the VFDR is eliminatedheat removal. A Recovery Action may be required to open from the NSCA and will be addressed in thevalve FCV-38-09. (OMC001,PC031) NPO analysis.06 Turbine Building VFDR-06-018 Unavailability of A postulated fire in this area may result in the loss of Train This VFDR is associated only with operation ofNorth, el. 250 Shutdown Cooling 12 power adversely affecting credited SDC valve IV-38-02. the Shutdown Cooling System. The ShutdownValve IV-38-02 SDC is required to support the decay heat removal function. Cooling System is only required to supportLocal-Manual operation of IV-38-02 per N1-SOP-21.1 may CSD. On this basis, the VFDR is eliminatedbe required. (PC031) from the NSCA and will be addressed in theNPO analysis.07 Turbine Building VFDR-07-003 Unavailability of A separation concern exists for a postulated fire in the area This VFDR is associated only with operation ofSouth & West, Shutdown Cooling for shutdown cooling. Credited pump, PMP-38-152, is the Shutdown Cooling System. The Shutdownel. 250 Pump PMP-38-152 required to operate to support decay heat removal. Fire Cooling System is only required to supportdamage to cable 17-62 can prevent remote start of the CSD. On this basis, the VFDR is eliminatedcredited SDC pump due to loss of RCS Temperature Switch from the NSCA and will be addressed in thepermissive. Local breaker operation is required. (OPOIOA, NPO analysis.OMC001)07 Turbine Building VFDR-07-009 Loss of Instrument Air A deterministic assumption assumes a potential loss of This VFDR is associated only with operation ofSouth & West, instrument air for a fire in any area of the plant. BV-70-53 the Shutdown Cooling System. The Shutdownel. 250 may fail closed on loss of instrument air. Valve is required Cooling System is only required to supportopen for cold shutdown to support decay heat removal. A CSD. On this basis, the VFDR is eliminatedRecovery Action may be required to open valve BV-70-53 from the NSCA and will be addressed in thelocally to supply RBCLC water to the SDC heat exchanger. NPO analysis.(PCO31)07 Turbine Building VFDR-07-010 Loss of Instrument Air A deterministic assumption assumes a potential loss of This VFDR is associated only with operation ofSouth & West, instrument air for a fire in any area of the plant FCV-38-10 the Shutdown Cooling System. The Shutdownel. 250 may fail closed on loss of instrument air. Valve is required Cooling System is only required to supportopen for cold shutdown to support decay heat removal. A CSD. On this basis, the VFDR is eliminatedRecovery Action may be required to open valve FCV-38-10 from the NSCA and will be addressed in thelocally to control SDC cooldown. (PC031) NPO analysis.09 Turbine Building VFDR-09-013 Loss of Instrument Air A deterministic assumption assumes a potential loss of This VFDR is associated only with operation ofEast, el. 250 instrument air for a fire in any fire area of the plant. Flow the Shutdown Cooling System. The Shutdowncontrol valve BV-70-53 may fail closed on loss of instrument Cooling System is only required to supportair. Valve is required open for cold shutdown to support CSD. On this basis, the VFDR is eliminateddecay heat removal. A recovery action may be required to from the NSCA and will be addressed in theopen valve BV-70-53 locally to supply RBCLC water to SDC NPO analysis.heat exchanger. (PC031)09 Turbine Building VFDR-09-014 Loss of Instrument Air A deterministic assumption assumes a potential loss of This VFDR is associated only with operation ofEast, el. 250 instrument air for a fire in any fire area of the plant. SDC the Shutdown Cooling System. The Shutdownheat exchanger outlet flow control valve FCV-38-11 may fail Cooling System is only required to supportclosed on loss of instrument air. Valve is required open for CSD. On this basis, the VFDR is eliminatedcold shutdown to support decay heat removal. A Recovery from the NSCA and will be addressed in theAction may be required to open valve FCV-38-11 (PC031, NPO analysis.OMCO01)12 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 01-1: Cold Shutdown VFDRs Eliminated from Table B-3Fire AreaFire Area VFDR ID VFDR Title Details CommentsDescription09 Turbine Building VFDR-09-015 Loss of Instrument Air A deterministic assumption assumes a potential loss of This VFDR is associated only with operation ofEast, el. 250 instrument air for a fire in any fire area of the plant. SDC the Shutdown Cooling System. The Shutdownheat exchanger outlet flow control valve FCV-38-09 may fail Cooling System is only required to supportclosed on loss of instrument air. Valve is required open for CSD. On this basis, the VFDR is eliminatedcold shutdown to support decay heat removal. A Recovery from the NSCA and will be addressed in theAction may be required to open valve FCV-38-09 (PC031, NPO analysis.OMC001)09 Turbine Building VFDR-09-021 Unavailability of A postulated fire in this area may result in the loss of Train This VFDR is associated only with operation ofEast, el. 250 Shutdown Cooling 12 power adversely affecting credited SDC valve IV-38-02. the Shutdown Cooling System. The ShutdownValve IV-38-02 SOC is required to support the decay heat removal function. Cooling System is only required to supportLocal-Manual operation of IV-38-02 per N1-SOP-21.1 may CSD. On this basis, the VFDR is eliminatedbe required. (PC031) from the NSCA and will be addressed in theNPO analysis.10 Cable Spreading VFDR-10-011 Failure of Shutdown A separation concern exists for a postulated fire in this area This VFDR is associated only with operation ofRoom, el. 250-0 Cooling System Valve for the Shutdown Cooling system. SDC valve IV-38-01 is the Shutdown Cooling System. The ShutdownIV-38-01 To Open required open for CSD to support decay heat removal. Fire Cooling System is only required to supportdamage (ground) to cable 12DV-29 prevents SDC IV-38-01 CSD. On this basis, the VFDR is eliminatedfrom opening. When power is restored to the valve and the from the NSCA and will be addressed in thecontrol switch operated, the control circuit fuse will blow NPO analysis.due to a dead short across the CPT. The EC's will attempt toinitiate automatically on either high reactor pressure orlow-low reactor level. However, both paths of EC's may beadversely impacted by the following: Potential inventorylosses via the Main Steam Lines and EC vent and drain linesdiscussed above(OP039, OP048)10 Cable Spreading VFDR-10-019 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofRoom, el. 250-0 instrument air. BV-70-53 may fail closed on loss of the Shutdown Cooling System. The Shutdowninstrument air. Valve is required open for cold shutdown to Cooling System is only required to supportsupport decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminatedrequired to open valve BV-70-53 locally to supply RBCLC from the NSCA and will be addressed in thewater to SDC heat exchangers. (PC031) NPO analysis.10 Cable Spreading VFDR-10-020 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofRoom, el. 2S0-0 instrument air. FCV-38-09 may fail closed on loss of the Shutdown Cooling System. The Shutdowninstrument air. Valve is required open for cold shutdown to Cooling System is only required to supportsupport decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminatedrequired to open valve FCV-38-09 locally to control SDC from the NSCA and will be addressed in thecooldown. (PC031) NPO analysis.10 Cable Spreading VFDR-10-026 Unavailability of A postulated fire in this area may damage cable 16-34 This VFDR is associated only with operation ofRoom, el. 250-0 Shutdown Cooling adversely impacting credited SDC pump PMP-38-149. PMP- the Shutdown Cooling System. The ShutdownPump PMP-38-149 38-149 is required to support decay heat removal. SDC Cooling System is only required to supportPMP-38-149 is repaired and operated locally at PB 16 per CSD. On this basis, the VFDR is eliminatedN1-DRP-GEN-003, Attachment 6. (PC037) from the NSCA and will be addressed in theNPO analysis.10 Cable Spreading VFDR-10-027 Unavailability of A postulated fire in this area may damage cable 12DV-11 or This VFDR is associated only with operation ofRoom, el. 250-0 Shutdown Cooling 12DV-29 adversely affecting credited SDC valve IV-38-02. the Shutdown Cooling System. The ShutdownValve IV-38-02 SDC is required to support the decay heat removal function. Cooling System is only required to supportLocal-Manual operation of IV-38-02 per N1-SOP-21.1 may CSD. On this basis, the VFDR is eliminatedbe required. (PC031) from the NSCA and will be addressed in theNPO analysis.13 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 01-1: Cold Shutdown VFDRs Eliminated from Table B-3Fire AreaFire Area VFDR ID VFDR Title Details CommentsDescription11 Control VFDR-11-012 Unavailability of A separation concern exists for a postulated fire in this area This VFDR is associated only with operation ofComplex, el. 261 Shutdown Cooling for shutdown cooling. Credited SDC pump, PMP-38-152, is the Shutdown Cooling System. The Shutdownand el. 277 Pump PMP-38-152 required to support decay heat removal for CSD. The Cooling System is only required to supportcredited SDC pump, PMP-38-152, may spuriously start and CSD. On this basis, the VFDR is eliminatedrun with no suction source. SDC pump PMP-38-152 may from the NSCA and will be addressed in thespuriously start due to an internal wire-to-wire short on NPO analysis.cable 17-23. SDC isolation valve IV-38-01 has powerremoved to prevent spurious opening for a fire in this firearea. SDC valves IV-38-02 and BV-38-04 are normally closed.Min-flow recirc valve FCV-38-131 may remain closed due toan internal wire-to-wire short on cable 1K-4. In the eventthe credited SDC pump is not available or other equipmentoperation causes a vessel overfill rendering EC's unavailable, various circuit failures in Train 11 and 12 CS valves couldrender the CS system unavailable for vessel injection andheat removal (OP044, OP026)11 Control VFDR-11-016 Loss of Instrument Air A deterministic assumption assumes a potential loss of This VFDR is associated only with operation ofComplex, el. 261 instrument air. Block Valve, BV-70-S3 may fail closed on loss the Shutdown Cooling System. The Shutdownand el. 277 of instrument air. This Valve is required to be open for cold Cooling System is only required to supportshutdown to support decay heat removal. A Recovery CSD. On this basis, the VFDR is eliminatedAction may be required to open valve BV-70- 53 (PC031) from the NSCA and will be addressed in theNPO analysis.11 Control VFDR-11-O24 Unavailability of Power A postulated fire in this area may cause the loss of PB 167 This VFDR is associated with a power supplyComplex, el. 261 Board PB 167 due to damage to the following Train 11 component cables: recovery action to assure availability of theand el. 277 SDC system. On this basis, the VFDR isBKR-(16B/013B)R1043/603: Cable 16-3 eliminated from the NSCA and will beBKR-(102/2-9)R1021/171: Cables 101-6, 102-2, 111-10 addressed in the NPO analysis.BKR-(102/2-1)R1022/571: Cables 101-6, 102-22, 102-28,102-29, 102-30, 102-31, 102-43, 102-44, 102-45, 102-46,102-49, 102-67, 11B-28, 1A-124, 1A-147, 1B-126, 1S-2386BKR-(101/2B-1)R1012/151: Cables 101-6, 11B-28, 1A-147,1A-60In support of shutdown from outside the control room,power is realigned from Train 12 to PB 167 per N1-DRP-GEN-004, Attachment 4. (PC044)11 Control VFDR-11-028 Unavailability of A postulated fire in this area may result in misoperation of This VFDR is associated only with operation ofComplex, el. 261 Shutdown Cooling credited SDC isolation valve IV-38-01 due to fire damage to the Shutdown Cooling System. The Shutdownand el. 277 Valve IV-38-01 cable 167-11 or 167-12. To support meeting the decay heat Cooling System is only required to supportremoval function, SDC valve IV-38-01 is repaired and CSO. On this basis, the VFDR is eliminatedoperated from PB 167 per N1-DRP-GEN-004, Attachment from the NSCA and will be addressed in the15. (PC049) NPO analysis.11 Control VFDR-11-029 Unavailability of A postulated fire in this area may result in misoperation of This VFDR is associated only with operation ofComplex, el. 261 Shutdown Cooling credited SOC isolation valve IV-38-13 due to fire damage to the Shutdown Cooling System. The Shutdownand el. 277 Valve IV-38-13 cable 167-15 or 167-16. To support meeting the decay heat Cooling System is only required to supportremoval function, SDC valve IV-38-13 is repaired and CSD. On this basis, the VFDR is eliminatedoperated from PB 167 per N1-ORP-GEN-004, Attachment from the NSCA and will be addressed in the16. (PCO5O) NPO analysis.11 Control VFDR-11-035 Manual Operation of In support of shutdown from outside the control room and This VFDR is associated only with operation ofComplex, el. 261 Shutdown Cooling the decay heat removal performance function, SDC valves the Shutdown Cooling System. The Shutdownand el. 277 Valves BV-38-04, FCV- BV-38-04, FCV-38-10 and IV-38-02 are operated locally per Cooling System is only required to support38-10, and IV-38-02 N1-DRP-GEN-004, Attachment 12, Attachment 13 or CSD. On this basis, the VFDR is eliminatedAttachment 14. (PCO48) from the NSCA and will be addressed in theNPO analysis.12 Administration VFDR-12-005 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofBuilding, el. instrument air. BV-70-53 may fail closed on loss of the Shutdown Cooling System. The Shutdown250.0 instrument air. Valve is required open for cold shutdown to Cooling System is only required to supportsupport decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminatedrequired to open valve BV-70-S3. (PC031) from the NSCA and will be addressed in theNPO analysis.14 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAT 01-1: Cold Shutdown VFDRs Eliminated from Table B-3Fire AreaFire Area VFDR ID VFDR Title Details CommentsDescription12 Administration VFDR-12-006 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofBuilding, el. instrument air. FCV-38-09 may fail closed on loss of the Shutdown Cooling System. The Shutdown250.0 instrument air. Valve is required open for cold shutdown to Cooling System is only required to supportsupport decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminatedrequired to open valve FCV-38-09. (PC031) from the NSCA and will be addressed in theNPO analysis.12 Administration VFDR-12-007 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofBuilding, el. instrument air. FCV-38-10 may fail closed on loss of the Shutdown Cooling System. The Shutdown250.0 instrument air. Valve is required open for cold shutdown to Cooling System is only required to supportsupport decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminatedrequired to open valve FCV-38-10. (PC031) from the NSCA and will be addressed in theNPO analysis.12 Administration VFDR-12-008 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofBuilding, el. instrument air. FCV-38-11 may fail closed on loss of the Shutdown Cooling System. The Shutdown250.0 instrument air. Valve is required open for cold shutdown to Cooling System is only required to supportsupport decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminatedrequired to open valve FCV-38-11. (OMC001) from the NSCA and will be addressed in theNPO analysis.13 Screenhouse VFDR-13-005 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofinstrument air for a fire in any fire area of the plant. FCV-38- the Shutdown Cooling System. The Shutdown09 may fail closed on loss of instrument air. Valve is Cooling System is only required to supportrequired open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminatedremoval. A Recovery Action may be required to open valve from the NSCA and will be addressed in theFCV-38-09 (PC031, OMC001) NPO analysis.13 Screenhouse VFDR-13-006 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofinstrument air for a fire in any fire area ofthe plant. FCV-38- the Shutdown Cooling System. The Shutdown10 may fail closed on loss of instrument air. Valve is Cooling System is only required to supportrequired open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminatedremoval. A Recovery Action may be required to open valve from the NSCA and will be addressed in theFCV-38-10 (PC031, OMCO01) NPO analysis.13 Screenhouse VFDR-13-007 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofinstrument air for a fire in any fire area ofthe plant. FCV-38- the Shutdown Cooling System. The Shutdown11 may fail closed on loss of instrument air. Valve is Cooling System is only required to supportrequired open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminatedremoval. A Recovery Action may be required to open valve from the NSCA and will be addressed in theFCV-38-11. (PC031, OMCO01) NPO analysis.13 Screenhouse VFDR-13-008 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofinstrument air for a fire in any fire area of the plant. BV-70- the Shutdown Cooling System. The Shutdown53 may fail closed on loss of instrument air. Valve is Cooling System is only required to supportrequired open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminatedremoval. A Recovery Action may be required to open valve from the NSCA and will be addressed in theBV-70-53 (PC031) NPO analysis.14 Diesel Fire VFDR-14-005 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofpump Room, el. instrument air for a fire in any fire area of the plant. BV-70- the Shutdown Cooling System. The Shutdown261 53 may fail closed on loss of instrument air. Valve is Cooling System is only required to supportrequired open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminatedremoval. A Recovery Action may be required to open valve from the NSCA and will be addressed in theBV-70-53. (PC031) NPO analysis.14 Diesel Fire VFDR-14-006 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofpump Room, el. instrument air for a fire in any fire area of the plant. FCV-38- the Shutdown Cooling System. The Shutdown261 09 may fail closed on loss of instrument air. Valve is Cooling System is only required to supportrequired open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminatedremoval. A Recovery Action may be required to open valve from the NSCA and will be addressed in theFCV-38-09. (PC031,OMC001) NPO analysis.14 Diesel Fire VFDR-14-007 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofpump Room, el. instrument air for a fire in any fire area of the plant. FCV-38- the Shutdown Cooling System. The Shutdown261 10 may fail closed on loss of instrument air. Valve is Cooling System is only required to supportrequired open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminatedremoval. A Recovery Action may be required to open valve from the NSCA and will be addressed in theFCV-38-10. (PC031, OMH0O1) NPO analysis.15 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 01-1: Cold Shutdown VFDRs Eliminated from Table B-3Fire AreaFire Area VFDR ID VFDR Title Details CommentsDescription14 Diesel Fire VFDR-14-008 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofpump Room, el. instrument air for a fire in any fire area of the plant. FCV-38- the Shutdown Cooling System. The Shutdown261 11 may fail closed on loss of instrument air. Valve is Cooling System is only required to supportrequired open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminatedremoval. A Recovery Action may be required to open valve from the NSCA and will be addressed in theFCV-38-11. (PC031,OMCO01) NPO analysis.15 Radwaste and VFDR-15-005 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofWaste Disposal instrument air. BV-70-53 may fail closed on loss of the Shutdown Cooling System. The ShutdownBuildings, el. instrument air. Valve is required open for cold shutdown to Cooling System is only required to support252 to 292 support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminatedrequired to open valve BV-70-53. (PC031) from the NSCA and will be addressed in theNPO analysis.15 Radwaste and VFDR-1S-006 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofWaste Disposal instrument air. FCV-38-09 may fail closed on loss of the Shutdown Cooling System. The ShutdownBuildings, el. instrument air. Valve is required open for cold shutdown to Cooling System is only required to support252 to 292 support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminatedrequired to open valve FCV-38-09.(PC031, OMC001) from the NSCA and will be addressed in theNPO analysis.is Radwaste and VFDR-15-007 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofWaste Disposal instrument air. FCV-38-10 may fail closed on loss of the Shutdown Cooling System. The ShutdownBuildings, el. instrument air. Valve is required open for cold shutdown to Cooling System is only required to support252 to 292 support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminatedrequired to open valve FCV-38-10.(PCO31, OMC001) from the NSCA and will be addressed in theNPO analysis.15 Radwaste and VFDR-15-008 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofWaste Disposal instrument air. FCV-38-11 may fail closed on loss of the Shutdown Cooling System. The ShutdownBuildings, el. instrument air. Valve is required open for cold shutdown to Cooling System is only required to support252 to 292 support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminatedrequired to open valve FCV-38-11.(PCO31, OMCO01) from the NSCA and will be addressed in theNPO analysis.16A Battery Board VFDR-16A-00S Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofRoom 12, el. instrument air for a fire in any fire area of the plant. FCV-38- the Shutdown Cooling System. The Shutdown261 09 may fail closed on loss of instrument air. Valve is Cooling System is only required to supportrequired open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminatedremoval. A Recovery Action may be required to open valve from the NSCA and will be addressed in theFCV-38-09. (PC031, OMC0O1) NPO analysis.16A Battery Board VFDR-16A-0O6 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofRoom 12, el. instrument air for a fire in any fire area of the plant. FCV-38- the Shutdown Cooling System. The Shutdown261 11 may fail closed on loss of instrument air. Valve is Cooling System is only required to supportrequired open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminatedremoval. A Recovery Action may be required to open valve from the NSCA and will be addressed in theFCV-38-11. (PC031, OMC001) NPO analysis.16A Battery Board VFDR-16A-007 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofRoom 12, el. instrument air for a fire in any fire area of the plant. BV-70- the Shutdown Cooling System. The Shutdown261 53 may fail closed on loss of instrument air. Valve is Cooling System is only required to supportrequired open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminatedremoval. A Recovery Action may be required to open valve from the NSCA and will be addressed in theBV-70-53. (PC031) NPO analysis.16A Battery Board VFDR-16A-008 Unavailability of Due to the unavailability of EDG 103 (Train 12, Path B), This VFDR is associated only with operation ofRoom 12, el. Shutdown Cooling power is not available to motor operated valve IV-38-02. IV- the Shutdown Cooling System. The Shutdown261 Valve IV-38-02 38-02 is a normally closed valve that fails as-is on loss of Cooling System is only required to supportpower. Valve is required open for cold shutdown to support CSD. On this basis, the VFDR is eliminateddecay heat removal. A Recovery Action may be required to from the NSCA and will be addressed in theopen valve IV-38-02. (PC031) NPO analysis.16B Battery Board VFDR-16B-005 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofRoom 11, el. instrument air for a fire in any fire area of the plant. FCV-38- the Shutdown Cooling System. The Shutdown261 10 may fail closed on loss of instrument air. Valve is Cooling System is only required to supportrequired open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminatedremoval. A Recovery Action may be required to open valve from the NSCA and will be addressed in theFCV-38-10. (PC031) NPO analysis.16 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 01-1: Cold Shutdown VFDRs Eliminated from Table B-3Fire AreaFire Area VFOR ID VFDR Title Details CommentsDescription16B Battery Board VFDR-16B-006 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofRoom 11, el. instrument air for a fire in any fire area of the plant. BV-70- the Shutdown Cooling System. The Shutdown261 53 may fail closed on loss of instrument air. Valve is Cooling System is only required to supportrequired open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminatedremoval. A Recovery Action may be required to open valve from the NSCA and will be addressed in theBV-70-53. (PC031) NPO analysis.17A Battery Room VFDR-17A-00S Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of12, el. 277 to instrument air for a fire in any fire area of the plant. FCV-38- the Shutdown Cooling System. The Shutdown291 09 may fail closed on loss of instrument air. Valve is Cooling System is only required to supportrequired open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminatedremoval. A Recovery Action may be required to open valve from the NSCA and will be addressed in theFCV-38-09. NPO analysis.17A Battery Room VFDR-17A-006 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of12, el. 277 to instrument air for a fire in any fire area of the plant. FCV-38- the Shutdown Cooling System. The Shutdown291 11 may fail closed on loss of instrument air. Valve is Cooling System is only required to supportrequired open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminatedremoval. A Recovery Action may be required to open valve from the NSCA and will be addressed in theFCV-38-11. NPO analysis.17A Battery Room VFDR-17A-007 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of12, el. 277 to instrument air for a fire in any fire area of the plant. BV-70- the Shutdown Cooling System. The Shutdown291 53 may fail closed on loss of instrument air. Valve is Cooling System is only required to supportrequired open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminatedremoval. A Recovery Action may be required to open valve from the NSCA and will be addressed in theBV-70-53. NPO analysis.17A Battery Room VFDR-17A-O08 Unavailability of Due to the unavailability of EDG 103 (Train 12, Path B), This VFDR is associated only with operation of12, el. 277 to Shutdown Cooling power is not available to motor operated valve IV-38-02. IV- the Shutdown Cooling System. The Shutdown291 Valve iV-38-02 38-02 is a normally closed valve that fails as-is on loss of Cooling System is only required to supportpower. Valve is required open for cold shutdown to support CSD. On this basis, the VFDR is eliminateddecay heat removal. A Recovery Action may be required to from the NSCA and will be addressed in theopen valve iV-38-02. NPO analysis.178 Battery Room VFDR-17B-OOS Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of11, el. 277 to instrument air for a fire in any fire area of the plant. FCV-38- the Shutdown Cooling System. The Shutdown291 10 may fail closed on loss of instrument air. Valve is Cooling System is only required to supportrequired open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminatedremoval. A Recovery Action may be required to open valve from the NSCA and will be addressed in theFCV-38-10. (PC031) NPO analysis.17B Battery Room VFDR-17B-006 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of11, el. 277 to instrument air for a fire in any fire area of the plant. BV-70- the Shutdown Cooling System. The Shutdown291 53 may fail closed on loss of instrument air. Valve is Cooling System is only required to supportrequired open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminatedremoval. A Recovery Action may be required to open valve from the NSCA and will be addressed in theBV-70-S3. (PC031) NPO analysis.18 Emergency VFDR-18-007 Flow Diversion from A spurious actuation concern exists impacting valves FCV- This VFDR is associated only with operation ofDiesel Containment Spray 93-74 and FCV-93-73 by diverting flow due to wire-to-wire the CTS/CTRWS, which provides torusGenerator 102 Raw Water System to shorts on the following cables. An internal wire-to-wire cooling. Torus cooling is required to supportMissile Containment Spray short on cable 171-163 spuriously opens FCV-93-74 CSD, and is not necessary when the primaryEnclosure, el. System diverting CTSRW flow to the CS system. (OP035) An internal decay heat removal method is achieved via271 wire-to-wire short on cable 171-160 spuriously opens FCV- the EC's. On this basis, the VFDR is eliminated93-73 diverting CTSRW flow to the CTS system. (OP034)19 Diesel VFDR-19-003 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofGenerator instrument air. BV-70-53 may fail closed on loss of the Shutdown Cooling System. The ShutdownRoom 103, el. instrument air. Valve is required open for cold shutdown to Cooling System is only required to support250 support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminatedrequired to open valve BV-70-53. (PC031) from the NSCA and will be addressed in theNPO analysis.19 Diesel VFDR-19-004 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofGenerator instrument air. FCV-38-11 may fail closed on loss of the Shutdown Cooling System. The ShutdownRoom 103, el. instrument air. Valve is required open for cold shutdown to Cooling System is only required to support250 support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminatedrequired to open valve FCV-38-11. (OMC001) from the NSCA and will be addressed in theNPO analysis.17 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 01-1: Cold Shutdown VFDRs Eliminated from Table B-3Fire AreaFire Area VFDR ID VFDR Title Details CommentsDescription19 Diesel VFDR-19-00S Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofGenerator instrument air. FCV-38-09 may fail closed on loss of the Shutdown Cooling System. The ShutdownRoom 103, el. instrument air. Valve is required open for cold shutdown to Cooling System is only required to support250 support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminatedrequired to open valve FCV-38-09. (PC031) from the NSCA and will be addressed in theNPO analysis.19 Diesel VFDR-19-O06 Unavailability of Due to the unavailability of EDG 103 (Train 12, Path B), This VFDR is associated only with operation ofGenerator Shutdown Cooling power is not available to motor operated valve IV-38-02. IV- the Shutdown Cooling System. The ShutdownRoom 103, el. Valve IV-38-02 38-02 is a normally closed valve that fails as-is on loss of Cooling System is only required to support250 power. Valve is required open for cold shutdown to support CSD. On this basis, the VFDR is eliminateddecay heat removal. A Recovery Action may be required to from the NSCA and will be addressed in theopen valve IV-38-02. (PC031) NPO analysis.20 Diesel VFDR-20-001 Unavailability of Due to the unavailability of EDG 103 (Train 12, Path B), This VFDR is associated only with operation ofGenerator Shutdown Cooling power is not available to motor operated valve IV-38-02. IV- the Shutdown Cooling System. The ShutdownEnclosed Valve IV-38-02 38-02 is a normally closed valve that fails as-is on loss of Cooling System is only required to supportCableway, el. power. Valve is required open for cold shutdown to support CSD. On this basis, the VFDR is eliminated250 decay heat removal. A Recovery Action may be required to from the NSCA and will be addressed in theopen valve IV-38-02. (PC031) NPO analysis.20 Diesel VFDR-20-004 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofGenerator instrument air. BV-70-53 may fail closed on loss of the Shutdown Cooling System. The ShutdownEnclosed instrument air. Valve is required open for cold shutdown to Cooling System is only required to supportCableway, el. support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminated250 required to open valve BV-70-53. (PC031) from the NSCA and will be addressed in theNPO analysis.20 Diesel VFDR-20-005 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofGenerator instrument air. FCV-38-11 may fail closed on loss of the Shutdown Cooling System. The ShutdownEnclosed instrument air. Valve is required open for cold shutdown to Cooling System is only required to supportCableway, el. support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminated250 required to open valve FCV-38-11. (OMC001) from the NSCA and will be addressed in theNPO analysis.20 Diesel VFDR-20-006 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofGenerator instrument air. FCV-38-09 may fail closed on loss of the Shutdown Cooling System. The ShutdownEnclosed instrument air. Valve is required open for cold shutdown to Cooling System is only required to supportCableway, el. support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminated250 required to open valve FCV-38-09. (PC031) from the NSCA and will be addressed in theNPO analysis.21 Below Power VFDR-21-001 Unavailability of Due to the unavailability of EDG 103 (Train 12, Path B), This VFDR is associated only with operation ofBoards 102/103, Shutdown Cooling power is not available to motor operated valve IV-38-02. IV- the Shutdown Cooling System. The Shutdownel. 250 Valve IV-38-02 38-02 is a normally closed valve that fails as-is on loss of Cooling System is only required to supportpower. Valve is required open for cold shutdown to support CSD. On this basis, the VFDR is eliminateddecay heat removal. A Recovery Action may be required to from the NSCA and will be addressed in theopen valve IV-38-02. (PC031) NPO analysis.21 Below Power VFDR-21-004 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofBoards 102/103, instrument air. BV-70-53 may fail closed on loss of the Shutdown Cooling System. The Shutdownel. 2S0 instrument air. Valve is required open for cold shutdown to Cooling System is only required to supportsupport decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminatedrequired to open valve BV-70-53. (PC031) from the NSCA and will be addressed in theNPO analysis.21 Below Power VFDR-21-O05 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofBoards 102/103, instrument air. FCV-38-11 may fail closed on loss of the Shutdown Cooling System. The Shutdownel. 250 instrument air. Valve is required open for cold shutdown to Cooling System is only required to supportsupport decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminatedrequired to open valve FCV-38-11. (OMC001, PC031) from the NSCA and will be addressed in theNPO analysis.21 Below Power VFDR-21-006 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation ofBoards 102/103, instrument air. FCV-38-09 may fail closed on loss of the Shutdown Cooling System. The Shutdownel. 250 instrument air. Valve is required open for cold shutdown to Cooling System is only required to supportsupport decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminatedrequired to open valve FCV-38-09.( OMC001,PC031) from the NSCA and will be addressed in theNPO analysis.18 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 01-1: Cold Shutdown VFDRs Eliminated from Table B-3Fire AreaFire Area VFDR ID VFDR Title Details CommentsDescription22 Emergency VFDR-22-006 Flow Diversion from A separation concern exists for a postulated fire in the area This VFDR is associated only with operation ofDiesel Containment Spray for the Containment Spray Raw Water system. CTSRW is the Shutdown Cooling System. The ShutdownGenerator 102 Raw Water System to required to support the decay heat removal function. Both Cooling System is only required to supportFoundation Containment Spray loops of the credited secondary decay heat removal CSD. On this basis, the VFDR is eliminatedRoom, el. 250 System function can be lost. An internal wire-to-wire short on cable from the NSCA and will be addressed in theand Diesel 171-163 spuriously opens FCV-93-74 diverting flow from NPO analysis.Generator pump PMP-93-03 to the CS system. An internal wire-to-wireRoom, el. 261 short on cable 171-160 spuriously opens FCV-93-73diverting flow from pump PMP-93-04 to the CTS system.Diversion of the CTSRW flow paths away from the CTSRWheat exchangers results in a loss of Decay Heat Removal.(OP034, OP035)19 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAT 01-2: New VFDRs Added to Table B-3 to Support Extended Hot Shutdown OperationFire Area VFDR ID VFDR Title Details CommentsThe Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions. After a This VFDR is associated with inventoryperiod of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the initiation of plant shutdown, inventory makeup to the Emergency Condensermakeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s04 VFDR-04-009 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeu isneededfromtheFireProtectionWater Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and make m t he availabilityion122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR isto periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the MainControl Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions. After a This VFDR is associated with inventoryperiod of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the initiation of plant shutdown, inventory makeup to the Emergency Condensermakeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s05 VFDR-05-047 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire ProtectionWater Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR isto periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the MainControl Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions. After a This VFDR is associated with inventoryperiod of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the initiation of plant shutdown, inventory makeup to the Emergency Condensermakeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s06 VFDR-06-019 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire ProtectionWater Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR isto periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the MainControl Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions. After a This VFDR is associated with inventoryperiod of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the initiation of plant shutdown, inventory makeup to the Emergency Condensermakeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s07 VFDR-07-014 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire ProtectionWater Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR isto periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the MainControl Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions. After a This VFDR is associated with inventoryperiod of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the initiation of plant shutdown, inventory makeup to the Emergency Condensermakeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s09 VFDR-09-022 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire ProtectionWater Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR isto periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the MainControl Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions. After a This VFDR is associated with inventoryperiod of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the initiation of plant shutdown, inventory makeup to the Emergency Condensermakeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s10 VFDR-10-029 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire ProtectionWater Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR isto periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the MainControl Room is not assured.20 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 01-2: New VFDRs Added to Table B-3 to Support Extended Hot Shutdown OperationFire Area VFDR ID VFDR Title Details CommentsThe Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions. After a This VFDR is associated with inventoryperiod of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the initiation of plant shutdown, inventory makeup to the Emergency Condensermakeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s11 VFDR-11-037 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeu is needed from the Fire ProtectionWater Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and make m toe availabilityion122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR isto periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the MainControl Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions. After a This VFDR is associated with inventoryperiod of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the initiation of plant shutdown, inventory makeup to the Emergency Condensermakeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s12 VFDR-12-009 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire ProtectionWater Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR isto periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the MainControl Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions. After a This VFDR is associated with inventoryperiod of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the initiation of plant shutdown, inventory makeup to the Emergency Condensermakeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s13 VFDR-13-011 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire ProtectionWater Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR isto periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the MainControl Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions. After a This VFDR is associated with inventoryperiod of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the initiation of plant shutdown, inventory makeup to the Emergency Condensermakeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s14 VFDR-144009 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire ProtectionWater Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR isto periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the MainControl Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions. After a This VFDR is associated with inventoryperiod of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the initiation of plant shutdown, inventory makeup to the Emergency Condensermakeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s1s VFDR4lS-O09 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire ProtectionWater Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR isto periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the MainControl Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions. After a This VFDR is associated with inventoryperiod of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the initiation of plant shutdown, inventory makeup to the Emergency Condensermakeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s16A VFDR-16A-009 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is neededfromthe Fire ProtectionWater Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR isto periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the MainControl Room is not assured.21 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 01-2: New VFDRs Added to Table B-3 to Support Extended Hot Shutdown OperationFire Area VFDR ID VFDR Title Details CommentsThe Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions. After a This VFDR is associated with inventoryperiod of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the initiation of plant shutdown, inventory makeup to the Emergency Condensermakeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of g hours16B VFDR-16B-007 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire ProtectionWater Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR isto periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the MainControl Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions. After a This VFDR is associated with inventoryperiod of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the initiation of plant shutdown, inventory makeup to the Emergency Condensermakeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s17A VFDRJ17A-009 Long-Term EC Make-up Tank water tanks isnecessaryto assure continued availability ofthe EC makeup is needed from the Fire ProtectionWater Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR isto periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the MainControl Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions. After a This VFDR is associated with inventoryperiod of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the initiation of plant shutdown, inventory makeup to the Emergency Condensermakeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s17B VFDR-17B-007 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire ProtectionWater Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR isto periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the MainControl Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions. After a This VFDR is associated with inventoryperiod of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the initiation of plant shutdown, inventory makeup to the Emergency Condensermakeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s18 VFDR-18-011 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability ofthe EC makeup is needed from the Fire ProtectionWater Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR isto periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the MainControl Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions. After a This VFDR is associated with inventoryperiod of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the initiation of plant shutdown, inventory mTakeup to the Emergency Condensermakeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s19 VFDR-19-007 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire ProtectionWater Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR isto periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the MainControl Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions. After a This VFDR is associated with inventoryperiod of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the initiation of plant shutdown, inventory makeup to the Emergency Condensermakeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s20 VFDR-20-008 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability ofthe EC makeup is neededfromthe Fire ProtectionWater Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR isto periodically refill the EC condenser makeup tanks. Additionally, classified a tiSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the MainControl Room is not assured.22 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 01-2: New VFDRs Added to Table B-3 to Support Extended Hot Shutdown OperationFire Area VFDR ID VFDR Title Details CommentsThe Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions. After a This VFDR is associated with inventoryperiod of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the initiation of plant shutdown, inventory makeup to the Emergency Condensermakeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s21 VFDR-21-008 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire ProtectionWater Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR isto periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the MainControl Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions. After a This VFDR is associated with inventoryperiod of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the initiation of plant shutdown, inventory makeup to the Emergency Condensermakeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s22 VFDR-22-008 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire ProtectionWater Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR isto periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the MainControl Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions. After a This VFDR is associated with inventoryperiod of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the initiation of plant shutdown, inventory makeup to the Emergency Condensermakeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s23 VFDR-23-008 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire ProtectionWater Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR isto periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the MainControl Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions. After a This VFDR is associated with inventoryperiod of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the initiation of plant shutdown, inventory makeup to the Emergency Condensermakeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s24 VFDR-24-007 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire ProtectionWater Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR isto periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the MainControl Room is not assured.23 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Safe Shutdown / Circuit Analysis RAI 03LAR Section 4.3 and Attachment D describe the methods and results of the non-power operations (NPO)evaluation, including references to the applicable outage programs, procedures, and NPO analyses.Additional information is requested asfollows:a. Provide "Appendix B: NMPJ NPO Pinch Point Assessment" in the NPO fire area reviewsincluding a summary level identification of unavailable paths in each fire area and the resolutionfor each pinch point.b. During NPO modes, spurious actuation of valves can have a significant impact on the ability tomaintain decay heat removal and inventory control. Provide a description of any actions beingcredited to minimize the impact of fire-induced spurious actuations on power operated valves(e.g., air operated valves (AOVs) and motor operated valves (MO Vs)) during NPO either as pre-fire conditioning or as required during the fire response recovery (e.g., pre-fire rack-out, locallypinning of valves, and isolation of air supplies).For example, it appears to the NRC staff that the Technical Specifications (TS) allow theshutdown cooling isolation valves 38-01 and 38-13 to be inoperable in the open position forgreater than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months under certain specific conditions. During higher risk evolutions such as ashort time to boil, preventing the spurious closure of any of these valves would be advantageous.Provide justification for not invoking the TS allowed flexibility for maintaining these valves openduring higher risk evolutions (HREs).c. Identify locations where key safety functions (KSFs) are achieved via RAs or for whichinstrumentation not already included in the at-power analysis is needed to support RAs requiredto maintain safe and stable conditions. Identify those RAs and instrumentation relied upon inNPO and describe how RA feasibility is evaluated. Include in the description whether thesevariables have been or will be factored into operator procedures supporting these actions.For instance, during outage conditions when there is a short time to boil, describe the operatorresponse to a spurious closure of one of the shutdown cooling system motor operated isolationvalves 38-01 or 38-13. Describe how any RAs are feasible (e.g., can be reliably accomplished inthe available time frame).Response to Safe Shutdown / Circuit Analysis RAI 03GeneralThe following is background information and other details of the non-power operations (NPO) analysisthat form the baseline for the specific responses to Parts a through c of this RAI.NMPNS has elected to modify its NFPA 805 transition analysis for NMP1 to revise the approach fordemonstrating the ability to reach and maintain safe and stable conditions, as specified by NFPA 805. Theoriginal Nuclear Safety Capability Assessment (NSCA) established as its basis for demonstrating safe andstable conditions the requirement to maintain Keff < 0.99 with a reactor coolant temperature at or belowthe requirements for hot shutdown and then subsequently cool down and maintain the plant in a cold24 of 47 ENCLOSURE1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)shutdown condition. Consistent with NFPA 805 and supplemental guidance, NMPNS is revising its basisfor the NMP1 NSCA to include only the requirement to establish hot shutdown conditions, includinglong-term hot shutdown capability. The impact of this change is primarily limited to the NSCA, which isaddressed in the response to Safe Shutdown / Circuit Analysis RAI 01. The change to safe and stableconditions does not impact the NPO Plant Operating States (POSs), Key Safety Functions (KSFs), orpinch point analysis. Hence, the response to this RAI does not depend on results or conclusions describedin the response to Safe Shutdown / Circuit Analysis RAI 01.As discussed in the response to Safe Shutdown / Circuit Analysis RAI 01, demonstration of the nuclearsafety performance criteria for safe and stable conditions is performed in two analyses based on the plantoperating modes, as defined in the NMP1 TS. These analyses are defined as follows:* At-Power analysis for potential fires while in either: (i) the Power Operating Condition (Reactormode switch is in "Startup" or "Run" position and the reactor is critical or criticality is possible due tocontrol rod withdrawal), or (ii) the Shutdown Condition -Hot operating condition (Reactor modeswitch is in "Shutdown" position and reactor coolant temperature is greater than 212'F), with theShutdown Cooling (SDC) system not aligned for decay heat removal. (Refer to the response to SafeShutdown / Circuit Analysis RAI 01 for further discussion of this analysis and its results.)* Non-Power analysis for potential fires while in Shutdown Condition -Hot operating condition andlower operating conditions.A copy of TS Section 1.1 containing the definitions of the NMP 1 reactor operating conditions is providedas Figure SSD/CA RAI 01-1 in the response to Safe Shutdown / Circuit Analysis RAI 01. Table SSD/CARAI 03-1 below provides a correlation between the three POSs identified in FAQ 07-0040 and plantoperating modes defined in the NMPI TS. Note that the reference to "RHR" in the FAQ 07-0040descriptions of POS is analogous to the SDC system at NMP 1.25 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 03-1: POS to TS Operating Condition CorrelationPOS Number and Description(from FAQ 07-0040)NMP1 TS Operating Condition and Description" This POS starts when the RHR -Reactor mode switch is insystem is placed into service. "Shutdown" or "Refueling"" The vessel head is on and the positionRCS is closed such that an m No core alterations leadingP05 1 extended loss of the decay heat Shutdown Condition -Hot to an addition of reactivityremoval (DHR) function without Shutdown Condition -Cold are being performedoperator intervention could result m Reactor coolant temperaturein a RCS re-pressurization above is greater than 212'F (Hot)the shutoff head for the RHR or equal to or less thanpumps. 212°F (Cold)This POS represents the shutdowncondition when:(1) The vessel head is removedand reactor pressure vesselwater level is less than thePOS 2 minimum level required for Major Maintenance No fuel is in the Reactormovement of irradiated fuel Conditionassemblies within the reactorpressure vessel as defined byTechnical Specifications, OR(2) A sufficient RCS vent pathexists for decay heat removal.m This POS represents the -Reactor mode switch is inshutdown condition when the "Refueling" positionreactor pressure vessel water m Fuel may be loaded orlevel is equal or greater than the unloadedPOS 3 minimum level required for Refueling Condition m Reactor coolant temperaturemovement of irradiated fuel is less than 212'Fassemblies within the reactor -No more than one operablepressure vessel as define by control rod is withdrawnTechnical Specificationsm This POS occurs during Mode 5As described in LAR Attachment D, procedure NIP-OUT-01, "Shutdown Safety," defines higher riskevolutions (HREs) and establishes KSFs and defense-in-depth (DID) strategies to protect the KSFs. HREsare defined as:"Outage activities, plant configurations, or conditions during shutdown where the plant is moresusceptible to an event causing the loss of a key safety function or the number of key safety systemsdrops below the shutdown safety criteria."NIP-OUT-01 ensures that HREs are identified and communicated to plant personnel with applicableprecautions and / or contingency plans clearly identified; e.g., on the Outage Schedule Shutdown SafetyReview (SSR) reports. KSFs considered for HREs, as required by NIP-OUT-01, include:26 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)1. Decay Heat Removal Capability. Assessments for maintenance activities affecting decay heatremoval capability should consider that the ability of systems and components to remove decay heatis dependent on a variety of factors, including the plant configuration, availability of other key safetysystems and components, and the ability of operators to diagnose and respond properly to an event.For example, assessment of maintenance activities that impact the decay heat removal key safetyfunction should consider:* Initial magnitude of decay heat.* Time to boil.* Time to core uncover.* Initial RCS water inventory condition (for example, filled, reduced, reactor cavity flooded, etc.).* RCS configurations (for example, reactor vessel open/closed, recirculation nozzle plugs installedor loop isolation valves closed, vent paths available, temporary covers installed, main steam lineplugs installed, etc.).When any fuel is offloaded to the spent fuel pool during the refueling outage, the decay heat removalfunction will be at least partially shifted from the RCS to the spent fuel pool (SFP). When the core iscompletely offloaded with the SFP gates installed, the decay heat removal function in the RCS can bemarked "Not Applicable."2. Inventory Control. Assessments for maintenance activities should address the potential for creatinginventory loss flow paths in both the RCS and the SFP. For example:Maintenance activities associated with the main steam lines (for example, safety or relief valveremoval, automatic depressurization system testing, main steam isolation valve maintenance, andso forth) can create a drain down path for the reactor cavity and fuel pool. This potential issignificantly mitigated through the use of main steam plugs.When the core is completely offloaded with the SFP gates installed, the reactor Inventory Controlfunction can be marked "Not Applicable."3. Power Availability. Assessments should consider the impact of maintenance activities on availabilityof electrical power. Electrical power is required during shutdown conditions to maintain cooling tothe reactor core and the SFP, to transfer decay heat to the heat sink, to achieve containment closurewhen needed, and to support other important functions.* Assessments for maintenance activities involving AC power sources and distribution systemsshould address providing defense in depth that is commensurate with the plant operating mode orconfiguration.* Assessments for maintenance activities involving the switchyard and transformer yard shouldconsider the impact on offsite power availability.27 of 47 ENCLOSURE1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)AC and DC instrumentation and control power is required to support systems that provide keysafety functions during shutdown. As such, maintenance activities affecting power sources,inverters, or distribution systems should consider their functionality as an important element inproviding appropriate defense in depth.4. Reactivity Control. The main aspect of this key safety function involves maintaining adequateshutdown margin in the reactor core and the SFP. During periods of cold weather, RCS temperaturescan also decrease below the minimum value assumed in the shutdown margin calculation. When inpower operation or startup conditions, availability of the Liquid Poison system must be considered.KSFs identified in NIP-OUT-01 associated with POSs specifically excluded from consideration by FAQ07-0040 are not discussed in the response to this RAI.Updates to Section 4.3 and Attachment F of the LAR Transition Report that are associated with theresponse to this RAI are provided in Enclosures 3 and 4.Part aTable SSD/CA RAI 03-2 below, from EIR 51-9171174, Appendix B -NMP1 NPO Pinch PointAssessment, provides summary level identification of KSF losses and pinch points on a fire zonebasis. The table identifies each KSF associated with a pinch point and the recommendations foraddressing the pinch points.As described in Section 4.3 and Attachment D of the LAR, the following KSFs are evaluated in eachfire zone:" Decay Heat Removal (DHR) for both the Reactor Vessel (RX) and the Spent Fuel Pool (SFP).* Inventory Control (INV) for both the Reactor Vessel and the Spent Fuel Pool.* Power (PWR) availability.The Reactivity Control KSF is not included in the NPO analysis because it is administrativelycontrolled in accordance with procedure NIP-OUT-01.Referring to Table SSD/CA RAI 03-2, the KSFs are categorized with codes assigned to each KSF -Fire Zone pair. Three codes have been established to summarize the fire impacts:' "1" (Impacted): At least one of the KSF paths associated with a given KSF is affected; i.e., acomponent of a specific KSF path or any of the component's required cables within the fire zoneare impacted, whereby that path can no longer be assured of being functional. However, at leastone other KSF path for the KSF remains available.* "L" (Lost): All available success paths for a given KSF are impacted.* "N" (None): No impacts to the KSF are identified.28 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)"Pinch Points" are then identified (on a fire zone basis), based on the loss of a KSF. An "N" in thepinch point column of Table SSD/CA RAI 03-2 indicates that no KSFs are lost in this fire zone. A"Y" in this column indicates that one or more KSFs are potentially lost in the fire zone, and thereforea pinch point is considered to exist. Fire zones are then categorized as follows:* Category 1 fire zones are not pinch points as they were found to have at least one success path foreach KSF. No recommendations for additional fire protection measures during HREs are madefor these zones. Standard DID strategies, as specified by procedure NIP-OUT-01, "ShutdownSafety," are adequate to address risk." Category 2 fire zones are pinch points as every success path is potentially lost for at least oneKSF. These KSF success paths can be preserved through fire protection/fire prevention actions,including the verification of functionality of available fire detection and suppression duringHREs.FAQ 07-0040 provides a listing of standard fire risk management methods that have been found to beacceptable for managing fire risk during HREs. During periods of NPO that are not defined as HREs,the standard fire protection DID actions are considered sufficient to minimize fire risk. During HREs,recommendations from FAQ 07-0040 have been identified for additional measures to consider as partof a comprehensive program to reduce fire risk. Each Category 2 fire zone includes one or morerecommendations from the list provided in Table SSD/CA RAI 03-3 to minimize fire risk to theKSFs, as described in Table SSD/CA RAI 03-2. Note that Recommendations 2B, 4, 6, and 7 fromFAQ 07-0040 are not used.29 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTILE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 03-2: Summary Level Identification of KSF Losses and Pinch Points(from EIR 51-9171174, Appendix B -NMP1 NPO Pinch Point Assessment)Fire KSFs Lost or Impacted Pinch RecommendationsFire Zone Area Fire Zone Description DHR INV Point Category (from Table SSD/CA RAI 03-3) Suppression DetectionAraPWR ?RX SFP RX SFPAl NA ADMINISTRATION BUILDING I I I I I N 1 Not a pinch point. No action needed.EL 250-0A2 NA ADMINISTRATION BUILDINGA2 NA EL 248-0 I I I I I N 1 Not a pinch point. No action needed.ABIA 12 RECORDS STORAGE AREA EL N N N N N N 1 Not a pinch point. No action needed.250-0ABIB 12 SAS EQUIPMENT AREA EL N N N N N N 1 Not a pinch point. No action needed.252-0AB1C 12 CPU EQUIPMENT AREA EL I I I I I N 1 Not a pinch point. No action needed.252-0ABID 12 GENERAL AREA EL 250-0 I I I I I N 1 Not a pinch point. No action needed.ABlIE 12 LOCKER AREA, LUNCH ROOM, N N N N N N 1 Not a pinch point. No action needed.OFFICES EL 261-0AB1F 4 FOAM ROOM EL 261-0 I I I I I N 1 Not a pinch point. No action needed.AB2A 12 ACCESS PASSAGEWAY EL N N N N N N 1 Not a pinch point. No action needed.248-0TECHNICAL SUPPORT AREAAB2B 12 EL 248-0 N N N N N N 1 Not a pinch point. No action needed.AB2C 12 RADIATION RECORDS AREA N N N N N N 1 Not a pinch point. No action needed.EL 248-0AB2D 12 WAREHOUSE AREA EL 248-0 N N N N N N 1 Not a pinch point. No action needed.AB3A 12 WAREHOUSE AREA EL 261-0 N N N N N N 1 Not a pinch point. No action needed.AB3B 12 OIL STORAGE ROOM EL 261-0 N N N N N N 1 Not a pinch point. No action needed.AB3C 12 STOREROOM TRUCK DOCK N N N N N N 1 Not a pinch point. No action needed.EL 261-0ELECTRICAUMECHANICALAB3D 12 SHOP AREA, OFFICE AREAS, N N N N N N 1 Not a pinch point. No action needed.LOCKER ROOMS EL 261-030 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)KSFs Lost or Impacted Pinch RecommendationsFire Zone Fre Fire Zone Description DHR INV Point Category (from Table SSDICA RAI 03-3) Suppression DetectionAraPWR ?RX SFP RX SFPAB3E 12 TELEPHONE ROOM 1 EL 261-0 N N N N N N 1 Not a pinch point. No action needed.AB3F 12 TELEPHONE ROOM 2 EL 261-0 N N N N N N 1 Not a pinch point. No action needed.AB4A 12 GENERAL OFFICE AREA EL N N N N N N 1 Not a pinch point. No action needed.277-0AB4B 12 FILE ROOM EL 277-0 N N N N N N 1 Not a pinch point. No action needed.AB4C 12 RECORDS PROCESSING N N N N N N 1 Not a pinch point. No action needed.AREA EL 277-0 1AB4D 12 GENERAL OFFICE AREA EL N N N N N N 1 Not a pinch point. No action needed.277-0PENTHOUSE VENTILATIONAB5 12 ROOM EL 290-0 N N N N N N 1 Not a pinch point. No action needed.ROOMTERY BOA90 OO-10EB1A 16A BATTERY BOARD ROOM 12 EL L L L I Y 2 1A and/or 3B and/or 5 None Yes261-0B1B 16B BATTERY BOARD ROOM 11 EL L L I I Y 2 1A and/or 3B and/or 5 None Yes261-0B2A 17A BATTERY ROOM 12 EL 277-0 I L L L I Y 2 1A and/or 3B and/or 5 None YesB2B 17B BATTERY ROOM 11 EL 277-0 I L L I I Y 2 1A and/or 3B and/or 5 None YesC1 10 CABLE SPREADING ROOM EL L L L L L Y 2 1A and/or 2A and/or 3A Yes Yes250-0C2 11 AUXILIARY CONTROL ROOM, L L L L L Y 2 1A and/or 2A and/or 3A and/or 5 Yes YesCOMPUTER ROOM 261-0C3 11 CONTROL ROOM EL 277-0 L L L L L Y 2 9 Yes YesEDG 103 FOUNDATION ROOMD1A 19 EL 20 I L I L I Y 2 IA and/or 2A and/or 3A Yes YesEL 250-0D1B 22 EG12FUDTOROM I L L I I Y 2 1A and/or 2A and/or 3A Yes YesEL 250-0D1C 20 EDG 103 CABLE ROUTING I L L L L Y 2 1A and/or 2A and/or 3A Yes YesAREA EL 250-0DID 21 ROOM BELOW PB'S 102& 103 I L L L L Y 2 1A and/or 2A and/or 3A Yes YesEL 250-0D2A 19 EDG 103 ROOM EL 261-0 I L I L I Y 2 1A and/or 2A and/or 3A Yes Yes31 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)FireZone Fire FiKSFs Lost or Impacted Pinch RecommendationsF nea DHR INV Point Category (from Table SSD/CA RAI 03-3) Suppression DetectionRX SFP RX SFPD2B 22 EDG 102 ROOM EL 261-0 L L L L I Y 2 1A and/or 2A and/or 3A Yes YesD2C 23 POWER BOARD 102 ROOM EL L L L L I Y 2 1A and/or 2A and/or 3A Yes Yes261-0D2D 24 POWERBOARD103ROOMEL L L L L I Y 2 1A and/or 2A and/or 3A Yes Yes261-0EDG 102 CONTROL CABLED3 18 MISSILE ENCLOSURE EL 271- L L L L I Y 2 1A and/or 5 None Yes0EXT EXT EXTERNAL TO PLANT L L L L L Y 2 1A and/or 2A and/or 3A and/or 10 Yes YesF1 4 FOAM ROOM EL 261-0 N N N N N N 1 Not a pinch point. No action needed.FBZ-1 1 REACTOR BUILDING FIRE N N N N N N 1 Not a pinch point. No action needed.BREAK ZONEFBZ-2 2 REACTOR BUILDING FIRE N N N N N N 1 Not a pinch point. No action needed.BREAK ZONEFBZR237N-1 1 REACTOR BUILDING EL 237-0 I I I I N 1 Not a pinch point. No action needed.COL N-Q, ROW 8-9FBZR237N-2 2 REACTOR BUILDING EL 237-0 1 1 1 1 N 1 Not a pinch point. No action needed.COL N-Q, ROW 8-9FBZR261N-1 1 REACTOR BUILDING EL 261-0 L L I I I Y 2 1Aand/or3Band/or5 Yes NoneCOL N-Q, ROW 8-9FBZR261N-2 2 REACTOR BUILDING EL 261-0 L L I I I Y 2 1A and/or 3B and/or 5 Yes NoneCOL N-Q, ROW 8-9FBZR281N-1 1 REACTOR BUILDING EL 281-0 L L I I I Y 2 1A and/or 3B and/or 5 Yes NoneCOL M-Q, ROW 6-7FBZR281N-2 2 REACTOR BUILDING EL 281-0 L L I I I Y 2 1A and/or 3B and/or 5 Yes NoneCOL M-0, ROW 6-7FBZR281S-1 1 REACTOR BUILDING EL 281-0 1 1 1 1 N 1 Not a pinch point. No action needed.COL K-L, ROW 7-8FBZR281 S-2 2 REACTOR BUILDING EL 281-0 1 1 1 1 N 1 Not a pinch point. No action needed.COL K-L, ROW 7-8FBZR298N-1 1 REACTOR BUILDING EL 298-0 1 1 1 1 N 1 Not a pinch point. No action needed.COL N-Q, ROW 7.5-8.532 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Fire KSFs Lost or Impacted Pinch RecommendationsFire Zone Area Fire Zone Description DHR INV Point Category (from Table SSD/CA RAI 03-3) Suppression DetectionPWR ?RX SFP RX SFPFBZR298N-2 2 REACTOR BUILDING EL 298-0 1 1 1 1 N 1 Not a pinch point. No action needed.COL N-Q, ROW 7.5-8.5FBZR298S-1 1 REACTOR BUILDING EL 298-0 1 1 1 1 N 1 Not a pinch point. No action needed.COL K-L, ROW 7-8FBZR298S-2 2 REACTOR BUILDING EL 298-0 1 1 1 1 N 1 Not a pinch point. No action needed.COL K-L, ROW 7-8FBZR318N-1 1 REACTOR BUILDING EL 318-0 1 1 1 1 N 1 Not a pinch point. No action needed.COL M-Q, ROW 6-7FBZR318N-2 2 REACTOR BUILDING EL 318-0 1 I I I I N 1 Not a pinch point. No action needed.COL M-Q, ROW 6-7FBZR318S-1 1 REACTOR BUILDING EL 318-0 I I I I N 1 Not a pinch point. No action needed.COL K-M, ROW 6-7FBZR318S-2 2 REACTOR BUILDING EL 318-0 1 I I I I N 1 Not a pinch point. No action needed.COL K-M, ROW 6-7 1_ ____FBZR340N-1 1 REACTOR BUILDING EL 340-0 1 I I I I N 1 Not a pinch point. No action needed.COL M-Q, ROW 6-7FBZR34ON-2 2 REACTOR BUILDING EL 340-0 1 1 1 1 N 1 Not a pinch point. No action needed.COL M-Q, ROW 6-7FBZR340S-1 1 REACTOR BUILDING EL 340-0 N N N N N N 1 Not a pinch point. No action needed.COL L-N, ROW 7-8FBZR340S-2 2 REACTOR BUILDING EL 340-0 N N N N N N 1 Not a pinch point. No action needed.COL L-N, ROW 7-8TURBINE BUILDING FIREFBZT261N 5 BREAK ZONE NORTH EL 261-0 I I I I I N 1 Not a pinch point. No action needed.TURBINE BUILDING FIREFBZT261S 5 BR NE SUTH E L L L L L Y 2 1B and/or 3B and/or 5 and/or 8 None NoneBREAK ZONE SOUTH EL 261-0OG1 GENERAL FLOOR AREA EL I I I I I N 1 Not a pinch point. No action needed.232-0OG2 5 GENERAL FLOOR AREA EL I I I I N 1 Not a pinch point. No action needed.247-0OG3 5 GENERAL FLOOR AREA EL I L I L I Y 2 1A and/or 2A and/or 3A Yes Yes261-033 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Fire FieZoeDecKSFs Lost or Impacted Pinch RecommendationsFireeoneecom m enPoi taCa egorFire Zone Area Fire Zone Description DHR INV Point Category (from Table SSD/CA RAI 03-3) Suppression DetectionRX SFP RX SFPR1 3 DRYWELL EL 237 -318 I I L I I Y 2 1B and/or 3B and/or 5 None NoneCTS PUMP ROOM ANDR1A 1 GENERAL FLOOR AREA EAST I L L L I Y 2 1A and/or 2A and/or 3A Yes YesEL 198-0 & 237-0CTS PUMP ROOM, CS PUMPR1B 2 ROOM, GENERAL FLOOR L L L I I Y 2 1A and/or 2A and/or 3A Yes YesAREA WEST EL 198-0 & 237-0ACCESS STAIRWELLRIC 1 SOUTHEAST EL 237-0 & 261-0 I I I Not a pinch point. No action needed.CS PUMP ROOM ANDPROTECTIVE CLOTHINGR1D 1 CHANGE AREA EL 198-0 & 237- I I I I Not a pinch point. No action needed.0R2A 1 GENERAL FLOOR AREA EAST L L L L I Y 2 1A and/or 2A and/or 3A Yes YesEL 261-0R2B 2 GENERAL FLOOR AREA WEST L L L I I Y 2 1A and/or 2A and/or 3A Yes YesEL 261-0R2C 2 SHUTDOWN COOLING ROOM I L L I I Y 2 1A and/or 3B and/or 5 Yes YesEL 261-0R2D 2 REACTOR BUILDING TRACK I I I I I N 1 Not a pinch point. No action needed.BAY EL 261-0R3A 1 GENERAL FLOOR AREA EAST L L L L I Y 2 1A and/or 2A and/or 3A Yes YesEL 281-0R3B 2 GENERAL FLOOR AREA WEST I L L I I Y 2 1A and/or 2A and/or 3A Yes YesEL 281-0R4A 1 GENERAL FLOOR AREA EAST L L I L I Y 2 1A and/or 3B and/or 5 None YesEL 298-0R4B 2 GENERAL FLOOR AREA WEST L L I I I Y 2 1A and/or 2A and/or 3A Yes YesEL 298-0EMERGENCY CONDENSERR4C-1 1 ISOLATION VALVE ROOM EL I I I I I N 1 Not a pinch point. No action needed.298-034 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Fire KSFs Lost or Impacted PinchReomnaisFire Zone Fire Fire Zone Description DHR INV Point Category Recommendations Suppression DetectionArea PW PontRateor (from Table SSD/CA RAI 03-3)RX SFP RX SFP PWR ?EMERGENCY CONDENSERR4C-2 2 ISOLATION VALVE ROOM EL I I I I I N 1 Not a pinch point. No action needed.298-0R5A 1 GENERLFLOORAREAEST I L I L I Y 2 1A and/or 3B and/or 5 None YesEL 318-0R51 2 GENERAL FLOOR AREA WEST I I I I N 1 Not a pinch point. No action needed.EL 318-0 1R6A 1 GENERAL FLOOR AREA EAST I I I I N 1 Not a pinch point. No action needed.EL 340-0R6B 2 GENERAL FLOOR AREA WEST N N N N N N 1 Not a pinch point. No action needed.EL 340-01RS1A 15 DRUM WASTE STORAGE N N N N N N 1 Not a pinch point. No action needed.VAULTS EL 252-0RS1B 15 ELECTRICAL EQUIPMENT I I I I I N 1 Not a pinch point. No action needed.ROOM EL 252-0GENERAL FLOOR AREARSIC 15 SOUTH, DRUM STORAGE N N N N N N 1 Not a pinch point. No action needed.ROOM EL 252-0RS2A 15 TRUCK LOADING AREA, N N N N N N 1 Not a pinch point. No action needed.NORTH EL 261-0RS2B TRUCK LOADING AREA, WEST N N N N N N 1 Not a pinch point. No action needed.EL 261-0RS2C 15 GENERAL FLOOR AREA EL N N N N N N 1 Not a pinch point. No action needed.261-0RS2D) 15 RADWASTE CONTROL ROOM, N N N N N N 1 Not a pinch point. No action needed.WEST EL 261-0RS2E 15 GENERAL FLOOR AREA, N N N N N N 1 Not a pinch point. No action needed.SOUTH EL 261-0RS3A 15 GENERAL FLOOR AREA, N N N N N N 1 Not a pinch point. No action needed.WEST EL 281-0RS4A 15 GENERAL FLOOR AREA, N N N N N N 1 Not a pinch point. No action needed.NORTHWEST EL 292-035 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)FireZone Fire FirZKSFs Lost or Impacted Pinch RecommendationsFire Zone Areaie Fire Zone Description DHR INV Point Category (from Table SSD/CA RAI 03-3) Suppression DetectionRX SFP RX SFPRS5B 15 GENERAL FLOOR AREA, N N N N N N 1 Not a pinch point. No action needed.SOUTHWEST EL 292-01S1 13 SCREENHOUSE EL 225-0 -I L I I I Y 2 1A and/or 3B and/or 5 None Yes256-0S2 14 DIESEL FIRE PUMP ROOM EL N N N N N N 1 Not a pinch point. No action needed.256-0TURBINET1 5 CONDENSER/HEATER BAY I I I L I Y 2 1A and/or 2A and/or 3A Yes YesAREA EL 250-0TURBINE BUILDING EL 240-261TIA 5 MSIV ROOM & STEAM TUNNEL I I I I I N 1 Not a pinch point. No action needed.T2A 6 TURBINE BUILDING EL 250-0 L L L L L Y 2 1A and/or 2A and/or 3A Yes YesT2B 7 TURBINE BUILDING SOUTH L L L L L Y 2 1A and/or 2A and/or 3A Yes YesAND WEST EL 250-0T2C TURBINE BUILDING OFFGAS I I I I I N 1 Not a pinch point. No action needed.TUNNEL EL 250-0T2D 9 TURBINE BUILDING GENERAL L L L L L Y 2 1A and/or 2A and/or 3A Yes YesAREA EAST EL 250-0T2E 7 UPS BATTERY ROOM EL 250 N N N N N N 1 Not a pinch point. No action needed.GENERAL FLOOR AREA EASTT3A 5 OF MSIV ROOM AND FIRE L L L L L Y 2 1A and/or 2A and/or 3A Yes YesZONE T1 EL 261-318GENERAL FLOOR AREA WESTT3B OF MSIV ROOM; ALSO SOUTH L L L L L Y 2 1Aand/or 2A and/or 3A Yes YesAND WEST OF FIRE ZONE 1EL 237-0 & 261-0T4A 5 GENERAL FLOOR AREA EAST L L L L L Y 2 1A and/or 2A and/or 3A Yes YesOF FIRE ZONE T1 EL 277-0T4B 5 GENERAL FLOOR AREA WEST L L L L L Y 2 1A and/or 2A and/or 3A Yes YesOF FIRE ZONE T1 EL 277-0T4C HYDROGEN SEAL OIL UNIT N N N N N N 1 Not a pinch point. No action needed.ROOM EL 277-0T4D 5 BATTERY ROOM EL 277 I I I I I N 1 Not a pinch point. No action needed.36 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)reZone Fire FieoKSFs Lost or Impacted Pinch RecommendationsFire Zone Area Fire Zone Description DHR INV Point Category (from Table SSDICA RAI 03-3) Suppression DetectionRX SFP RX SFPT5A 5 GENERAL FLOOR AREA I L I L I Y 2 1A and/or 2A and/or 3A Yes YesNORTH EL 291-0 1 1T6A 5 GENERLFLOORAREA I L I L I Y 2 1A and/or 2A and/or 3A Yes YesNORTH EL 305-6T6B 5 TURBINELAYDOWNAREA I L I L I Y 2 1A and/or 5 None YesEAST EL 300-0T6C 5 GENERAL FLOOR AREA I I I I I N 1 Not a pinch point. No action needed.SOUTH EL 300-0T6D 5 MECHANICAL STORAGE AREA N N N N N N 1 Not a pinch point. No action needed.EL 320-0T7A 5 GENERAL FLOOR AREA I L I L I Y 2 1A and/or 3B and/or 5 None YesSOUTH EL 320-0GENERAL FLOOR AREANORTH EL 333-0, GENERALT8A 5 FLOOR AREA NORTH EL 351- I I I I 1 N 1 Not a pinch point. No action needed.0, GENERAL FLOOR AREAEAST EL 369GENERAL FLOOR AREA WESTT8B 5 EL 369-0 I I I I I N 1 Not a pinch point. No action needed.GEEL A 369-0 250WD1 15 GENERAL AREA El 225-0 & I I I I N 1 Not a pinch point. No action needed.229-0WD2 15 GENERAL AREA EL 247-0 N N N N N N 1 Not a pinch point. No action needed.WD3A 15 GENERAL AREA EL 261-0 N N N N N N 1 Not a pinch point. No action needed.WD3B 15 RADWASTE CONTROL ROOM N N N N N N 1 Not a pinch point. No action needed.EL 261-0WD3C 15 BALER ROOM EL 261-0 N N N N N N 1 Not a pinch point. No action needed.WD3D 15 DOW SOLIDIFICATION AREA N N N N N N 1 Not a pinch point. No action needed.EL 261-0WD3E 15 TRUCK BAY EL 261-0 N N N N N N 1 Not a pinch point. No action needed.WD4 1 WASTE BUILDINGVENTILATION AREA EL 277-0 N N N N N N 1 Not a pinch point. No action needed.37 of 47 ENCLOSURE1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 03-3: List of Recommendations for Identified Pinch Points(from FAQ 07-0040)No. FAQ 07-0040 NMP1 Specific Recommendation for Outage PlanningRecommendation Category 2 Fire Zones ConsiderationsLimit hot work in this fire zone Outage planning considers periods1A Lii o oki hsfr oe of increased vulnerability forProhibition or limitation of hot during HRE conditions. l h in thisefareizone.workin irearea duinglimiting hot work in this fire zone.work in fire areas duringperiods of increased Outage planning considers periodsvulnerability. 1B Prohibit hot work in this fire zone of increased vulnerability forduring HREs. prohibiting hot work in this firezone.Verify that the available firedetection systems located in the Detection systems should be2A fire zone are functional. Post verified to be functional; i.e., notfirewatch in affected fire zones feragied t, etc.Verification of operable prior to entering HRE conditions if tagged out, etc.detection and /or suppression system(s) are impaired.in the vulnerable areas. Verify that the available firesuppression systems located in thearea are functional. Post firewatch Suppression systems should be2B* in affected fire zones prior to verified to be functional; i.e., notentering HRE conditions if tagged out.system(s) are impaired.Limit transient combustible Outage planning considers3A storage in this fire zone during limiting the hazard of combustibleProhibition or limitation of HRE conditions. materials.combustible materials in fireareas during periods ofincreased vulnerability. Prohibit transient combustible Outage planning considers3B storage in this fire zone during prohibiting the hazard ofHRE conditions. combustible materials.Plant configuration changes Power can be removed from various Outage planning considers using4* (e.g., removing power from components and equipment as part of alternate equipment and/or theequipment once it is placed in outage configuration line-ups prior to equipment's position wheneverits desired position). entering HRE conditions. removing power.Provision of additional firepatrols at periodic intervals or Provide a firewatch (continuous or Outage planning considers theother appropriate periodic) in this fire area during HRE appropriate compensatorycompensatory measures (such coditins measures required during periodsas surveillance cameras) conditions. of increased vulnerability.during increased vulnerability.Use of recovery actions to RActivities that may impact KSFs6* e oftigateepotentialosss tof Recovery actions to restore at least one should be limited and strictly6* mitigate potential losses of KSF success path can be taken. sol elmtdadsrclKSFs. controlled to mitigate losses.38 of 47 ENCLOSURE1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)No. FAQ 07-0040 NMP1 Specific Recommendation for Outage PlanningRecommendation Category 2 Fire Zones ConsiderationsIdentification and monitoring Outage planning considers the7* in-situ ignition sources for hazards from the introduction of"fire precursors" (e.g., combustible materials and sourcesequipment temperatures). of fire precursors.Reschedule the work to a Activities in these fire zones should be Outage planning considers8 period with lower risk or rescheduled to a period of non-HRE limiting work during periods ofhigher defense in depth (DID). conditions. HRE conditions.Control Rooms are constantly manned9 N/A locations. No other actions are required N/Aduring HRE conditions.Existing controls on switchyard activities10 N/A during HRE conditions are adequate and N/Awill manage fire risk as well.* These recommendations are not used.Part bThe NMP1 NPO pinch point analysis was developed in accordance with the guidance contained inFAQ 07-0040. The FAQ 07-0040 endorsed "Recommendations" utilized at NMP1 to reduce fire riskduring HREs are identified in Part a of this RAI response. The additional reduction in risk offered bythe "Recommended" strategies provides additional assurance that fire risk is minimized in areassusceptible to a loss of one or more KSFs during plant HREs.As discussed in the response to Part a and depicted in Table SSD/CA RAI 03-2, additional actions(e.g., pre-fire rack-out, locally pinning of valves, isolation of air supplies) are not relied upon as astrategy to reduce fire risk during HREs, including the impact of fire-induced spurious operations(single or multiple). The assessment of potential risk reduction options (including input fromOperations personnel) concluded that the actual additional risk posed by fire during HREs is bestcontrolled through the methods identified in Table SSD/CA RAI 03-2. Specifically, the NMP1 NPOstrategy does not credit the following methods:" Recovery Actions -Reliance on recovery actions during an outage is difficult to characterize forfeasibility due to the many variables that could exist, such as blockage of normal routes,scaffolding impact on lighting, equipment/material staging and movement, contract personnelcontingent, unusual equipment line ups, etc. For this reason, recovery actions are viewed as lesspredictable with respect to reliability and uncertainty in comparison to the risk reduction optionsselected.* Configuration Changes -The use of limited configuration changes to address in a preemptivemanner certain high consequence fire-induced failures, most notably spurious operations of keyvalves, was considered. However, after discussions with Operations personnel it was concluded39 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)that the reduction in operational flexibility to respond to a broader range of potential accidentsand abnormal conditions outweighs the marginal improvement in risk reduction associated withfire-induced spurious operations.With specific reference to the potential vulnerability of shutdown cooling isolation valves 38-01, 38-02, and 38-13 to fire-induced spurious closure, deliberately entering a TS required action wasevaluated as undesirable when viewed from a broader perspective beyond just potential fire events.Thus, the recommendations contained in Table SSD/CA RAI 03-2 are considered the best options toaugment existing procedures for managing shutdown risk, including risk from fire, during HREs.Part cAs shown in the response to Part a and in Table SSD/CA RAI 03-2, NMP1 does not credit recoveryactions as a strategy to reduce shutdown fire risk during HREs. The rationale for not employingrecovery actions is provided in the response to Part b above.40 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Safe Shutdown / Circuit Analysis RAI 07Based on a review of the updated final safety analysis report (UFSAR), switchgear other than motorcontrol centers (MCCs) use 125 VDC power for control of the electrically-operated circuit breakers sothe breakers may be operated if AC power is lost. Dualfeeds are provided to the DC control bus on eachpower board for added reliability, one each from either battery 11, 12 or 14.A generic concern in regards to the Fort Calhoun fire that occurred on June 7, 2011 (NRC SpecialInspection Report, March 12, 2012, ADAMS Accession No. ML12072A128) involves 125 VDC circuitsfrom both DC buses inside the same switchgear. Both DC buses were impacted with "soft" grounds thatremained after the fire had been isolated by removing power.With respect to the Fort Calhoun event, it appears that tile power boards at NMPJ have dual controlpower feeds. Describe if this issue has been considered. Describe if there are any proposed plans toperform modifications or procedure changes to address this issue.Response to Safe Shutdown / Circuit Analysis RAI 07General Description of the NMP1 125 VDC SystemThe safety related 125 VDC system at NMP 1 consists of two physically separate and independent trains(Batteries 11 and 12). Each train includes one 125 VDC station battery, two parallel static batterychargers (one primary and the other a backup), and one DC power distribution board. The battery boardsinclude the fuses and the fuse blocks required for distribution of 125V DC to various system loads. Theaugmented quality 125 VDC system consists of one 125 VDC station battery (Battery 14), a static batterycharger, and one DC power distribution battery board.The 125 VDC batteries 11, 12, and 14 are part of NMP1 Safe Shutdown Equipment. Battery 11 and theassociated battery board are located in Fire Areas 17B and 16B, respectively. Battery 12 and theassociated battery board are located in Fire Areas 17A and 16A, respectively. Battery 14 and theassociated battery board are located in Fire Area 5.Ground Detection Design FeaturesThe 125 VDC electrical distribution trains are operated independently and ungrounded, and, as such, asingle ground does not generate a fault current or disable the system. The system is equipped with grounddetection devices to indicate the occurrence of the first ground which allows operators to locate andcorrect the first ground. NRC Information Notice (IN) 94-80 "Inadequate DC Ground Detection in DirectCurrent Distribution Systems," alerted licensees to the potential for operating with undetectable groundsin vital direct current (DC) distribution systems due to inadequate ground-detection equipment orinadequate ground-alarm setpoints, or both. The IN recommended that ground detectors be incorporatedin the DC systems so that, if a single ground does occur, personnel are aware of the ground and can takeimmediate steps to clear the ground from the system. Failure to promptly eliminate a single ground couldmask subsequent additional grounds. Multiple grounds could lead to unpredictable spurious operation ofequipment, inoperable equipment, unanalyzed loads on batteries, or unanalyzed equipment failure modes.In response IN 94-80, the 125 VDC system ground detection scheme at NMP 1 was modified.41 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)A ground on the 125 VDC power system will be determined by the Ground Detection Relay andannunciated in the Control Room at panel A3 windows A3-4-2 for Battery 11, A3-4-3 for Battery 12, andA3-2-2 for Battery 14. Alarm Response Procedure Ni-ARP-A3 directs operators to Operating ProcedureNl-OP-47A (125 VDC System), Section H.8.0, which provides direction to locate and clear a groundwithin the 125 VDC system.System Loads ConfigurationFor added reliability, dual DC feeds are provided to a number of Power Boards, Emergency DieselGenerators (EDGs) 102 and 103, and the Diesel Fire Pump (DFP). The dual feeds are selected via amechanically operated knife switch located at each load and aligned to the normal DC feed. TableSSD/CA RAI 07-1 below provides a summary of loads with dual feeds.Table SSD/CA RAI 07-1: 125 VDC Electrical Distribution System Dual Feed Loads on BatteryBoards 11, 12, and 14Battery Board Load Battery Board Battery Board Battery Board#11 #12 #14Motor Generator (MG) Set 167 Normal AlternateBreaker Control -Power Board 11 Normal AlternateBreaker Control -Power Board 12 Alternate NormalBreaker Control -Power Board 13 Alternate NormalBreaker Control -Power Board 14 Alternate NormalBreaker Control -Power Board 15 Alternate NormalBreaker Control -Power Board 16 Normal AlternateBreaker Control -Power Board 17 Alternate NormalBreaker Control -Power Board 18 Alternate NormalBreaker Control -Power Board 101 Normal AlternateBreaker Control -Power Board 102 Normal AlternateBreaker Control -Power Board 103 Alternate NormalDC Valve Board 11 Normal AlternateDC Valve Board 12 Alternate NormalDiesel Fire Pump Normal AlternateHydrogen and Seal Oil System Annunciation Alternate NormalStator Water System Annunciation Alternate NormalEmergency Diesel Generator 102 Starting and Control Normal AlternateEmergency Diesel Generator 103 Starting and Control Alternate Normal42 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)NMP 1 Electrical Maintenance Procedures for Breakers and SwitchiearAs documented in the NRC Special Inspection Report 05000285/2011014 (Accession No.ML12072A128), the fire at Fort Calhoun Station occurred due to a high impedance connection whichcaused failure of a 480 VAC Breaker. The high impedance connection was caused by hardened grease onthe secondary disconnects and dirty secondary contacts in GE model AKD-5 low voltage switchgear. Theroot cause analysis determined that the electrical maintenance procedure associated with the low voltageswitchgear was less than adequate, that preventive maintenance activities were inadequate to ensureproper cleaning of conductors, proper torquing of bolted conductor and bus bar connections, and thatinspections for abnormal temperatures were inadequate.At NMP1, procedure Ni-EPM-GEN-310 implements preventive maintenance on 4.16KV Switchgear,600VAC Switchgear, 600 and 480 VAC Motor Control Centers (MCCs), and 125 VDC Battery PowerBoards. Attachments 1 and 2 of this procedure include specific steps which address the issues identifiedin the Fort Calhoun root cause analysis, as follows:* When maintaining GE AKD-5 Load Master Switchgear, the breaker primary disconnecting studs andfingers are cleaned and greased with a thin coat of Mobil 28. Mobil 28 is selected based on itsperformance characteristics, which include resistance to friction oxidation (fretting) and hardeningunder various environmental conditions.* Inspect bolted connections, and torque any loose connections in accordance with specific torquerequirements.Procedure NI-EPM-GEN-151 for the inspection of TYPE AK-50 and ITE K-LINE breakers includes thefollowing precautions and steps:* Prevent the mixing of Mobil 28 with previously used GE D50H15 or GE D50H47 lubricants since itmay result in grease hardening and breaker failure.* Inspect main, intermediate, and arcing movable and stationary contacts for discoloration that mayhave been caused by overheating.Procedure N I-EPM-GEN-182 for the inspection of MCCs includes the following precaution:To prevent component insulation degradation, Trichloroethane (CRC Lectra Clean) based solventsshall not be used on the component insulating parts of the MCC or MCC Bucket. Denatured andisopropyl alcohol are acceptable substitutes for cleaning/degreasing of component insulated parts.THC-based solvents may be used for cleaning/degreasing of current carrying parts in metallic MCCsas well as on metallic MCC bucket components.A review of past Condition Reports associated with breakers and breaker maintenance at Nine Mile Pointwas performed to determine if there has been any Condition Reports initiated due hardening of grease,dirty contacts, or loose connections. The following provides a summary of this review:43 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Condition Report Description1994-000170 OE -Misadjustment between GE 4.16 KV circuit breakers and their associatedcubicles2000-003715 OE 11485, Cleaning of silver contacts caused silver to be removed2001-001049 INPO SEN 218, Circuit breaker fault results in fire, LOOP, Reactor scram, andsevere turbine damage2001-004474 INPO SEN 221, Circuit Breaker to Bus Connector Faults Results in ReactorScram. The corrective actions associated with this CR included revisions toElectrical Preventive Maintenance procedures NI -EPM-GEN-1 50 and NI -EPM-GEN-3 10 to incorporate steps/precautions to not use abrasive cleanerwhen cleaning silver plated contact surfaces.2003-001128 GE Service Information Letter 448, Rev. 1, Recommendation for lubrication ofType AK GE Breakers2004-001184 Catastrophic Failure of MCCB 19A in 2NHS-MCCO10 due to phase-to-groundfault near line side connection. A corrective action of this Condition Reportimplemented procedure revisions to include specific steps and directions forinspection of MCCB line side power wiring (procedures NI -EPM-GEN-1 82and 310).The above Condition Reports and the associated corrective actions have resulted in procedure changesthat address the potential causes of the fire event at Fort Calhoun.Differences Between NMP1 and Fort CalhounIn addition to lack of adequate preventive maintenance, the NRC Special Inspection Report05000285/2011014 identified two addition contributing causes to the overall event at Fort Calhoun, asfollows:Implementation of a plant modification in 2009 that replaced AK-50 480 V main and bus-tie breakerswith Molded Case Square-D Masterpact circuit breaker/cradle assemblies and digital trip devices.The differences in form, fit, and function resulted in high resistance connections between the cradleassembly and bus stabs due to oxidation built-up caused by dissimilar metal (copper and silver) whichcontributed to the fire.There has been no modification implemented at NMP1 to replace breakers with the type identifiedabove." Unlike Fort Calhoun, the NMP1 DC feeds to power boards, EDGs, and the DFP are equipped withfuses. These fuses function to effectively clear and isolate the affected battery board from anovercurrent condition caused by a hot short or multiple shorts to ground.44 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Analysis of the Postulated Fire Scenario at NMP 1A circuit analysis for the NMP1 NFPA 805 transition was performed in accordance with NEI 00-01,Revision 2, which has been endorsed by Regulatory Guide (RG) 1.189, Revision 2. In accordance withthe guidance provided in NEI 00-01, Revision 2, evaluation of a potential "soft" ground (i.e.; a groundthat does not result in sufficient fault current to cause the circuit protective/isolation device to open) is notrequired. The following are excerpts from NEI 00-01, Section 3.5, with respect to ungrounded circuits:* In the case of an ungrounded circuit, postulating only a single short-to-ground on any part of thecircuit may not result in tripping the electrical protective device. Another short-to-ground on thecircuit or another circuit from the same source would need to exist to cause a loss of control power tothe circuit." Consider an individual, single short-to-ground on each conductor in each affected cable in a groundedcircuit. Consider the combined effects of shorts-to-ground if conductors are located in the samemulti-conductor cable in the primary circuit." For ungrounded circuits, two shorts-to-ground are required for the loss of control power to theindividual circuit. The recommended approach either assumes or evaluates for a second short-to-ground causing a loss of control power in the components control circuit for ungrounded circuits." Additionally, either assume a second short-to-ground exists in an ungrounded circuit resulting in aloss of control power or evaluate for an actual fire-induced cable impact with the potential to causethe second short-to-ground in the fire area." Depending on the coordination characteristics between the protective device on the circuit andupstream circuits, the power supply to other circuits could be affected. If multiple grounds can occurin a single fire area, they should be assumed to occur simultaneously unless justification to thecontrary is provided.In summary, the concern with respect to a postulated short to ground on an ungrounded DC control circuitis multiple fire induced grounds that could result in a loss of control capability due to opening of theisolation devices.Similar to Fort Calhoun Station, NMP1 125 V DC battery boards 11, 12, and 14 provide redundantcontrol power to a number of power boards, EDG 102 and 103 start and control circuits, and the DFP, aslisted in Table SSD/CA RAI 07-1. A mechanically operated transfer switch allows the operators tomanually re-align control power from the normal to the alternate 125 VDC battery board. The batteryboard loads listed in Table SSD/CA RAI 07-1 are located in Fire Areas 1, 2, 4, 5, 14, 19, 22, 23, and 24.The NMP1 Nuclear Safety Capability Assessment (NSCA) for the above Fire Areas did not address thepotential for a "soft" ground on the 125 VDC system. Fundamental assumptions in the circuit analysisand fire area assessment are that:1. A single short to ground will not affect the ability of the credited DC system to accomplish itsintended safe shutdown function, and45 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)2. Multiple hot shorts or shorts to ground will result in sufficient fault current as to cause actuation ofthe protective devices.A postulated fire within the Power Boards will potentially cause a single ground in the alternate 125 VDCsupply. However, this single ground will not affect the safe shutdown function(s) of the credited 125VDC bus. This is consistent with the conclusions of the NRC Special Inspection Report for the FortCalhoun event, which states on page 26:"The team concluded that dc control power remained available to the safety-related 4160 VAC busesthroughout the event, and the grounds on the dc buses would not have prevented the dc system fromperforming its safety function. Because the system was normally ungrounded, a single ground oneither the positive or negative bus of the system did not result in the loss of a circuit, but did indicatea degraded condition."Thus, the postulated "soft" ground on the credited 125 VDC electrical distribution system does not affectthe assumptions in the NMP1 NSCA performed to support NFPA 805 transition with respect to separationrequirements for redundant trains of the 125 VDC system. The analysis demonstrates that sufficientseparation exists to ensure one train of the 125 VDC system remains free of fire damage. A postulated fireat each of the power board locations does not affect the capability to maintain battery charging to theunaffected train of 125 VDC and, as such, a minimal leakage current (i.e.; below the fuse opening) due toa "soft" ground would not affect battery capacity or charging capability.Electrical Separation and IndependenceThe safety related electrical distribution system at NMP1 is designed to provide two redundant andindependent trains of control and power to safety related loads during and following anticipated transientsand design basis accidents. The design basis requirement also includes a criterion for limiting firedamage to one train of the electrical distribution system.To prevent paralleling the two trains of the safety related 125 VDC electrical distribution system, therebylosing train independence and redundancy, the following interlocks are provided:" The 125 VDC circuit breakers feeding computer MG Set 167 from DC battery boards 11 and 12 arekey interlocked to prevent closing both breakers at the same time." The 125 VDC circuit breakers feeding DC valve boards 11 and 12 from 125 VDC battery boards 11and 12 are mechanically interlocked to prevent closing both breakers at the same time.The 125 VDC system design and configuration meets the electrical separation requirements and singlefailure criterion and remains in compliance with the existing plant licensing basis based on the following:" The system is equipped with a ground detection circuit," Each control power feed is equipped with isolation devices which will effectively isolate the affectedbattery board from an overcurrent condition or hot short, and46 of 47 ENCLOSURE 1NINE MILE POINT UNIT 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)* The dual feeds from the redundant 125 VDC power boards to each compartment are selected via amechanically operated knife switch located in each power board to ensure electrical separation ismaintained.ConclusionThe event at Fort Calhoun is considered Operating Experience (OE). Although the existing NRCguidance (RG 1.189, NUREG-6850) and industry guidance (NEI 001-01) do not require evaluation of a"soft" ground as part of the circuit analysis and fire area assessments performed for NFPA 805 transition,the Fort Calhoun fire scenario has been evaluated for applicability to NMP1. This evaluation hasconcluded that the occurrence of a fire caused by a lack of proper breaker preventive maintenance and theresulting consequences is not a likely fire scenario at NMP1. This is mainly due to differences in designconfiguration and maintenance activities at NMP 1. In addition, the fuses associated with each DC feed tothe loads are sized to ensure that any short to ground faults are effectively isolated from the affectedbattery board.However, as documented in Information Notice 94-80, multiple grounds could lead to unpredictablespurious operation of equipment, inoperable equipment, and unanalyzed battery loads or equipmentfailure modes. To enhance operator knowledge and plant response to the potential for "soft" grounds, achange to post-fire safe shutdown procedure N1-SOP-21.1 ("Fire In Plant") is being processed. Thischange will alert the operators to the potential for a fire induced ground in the DC system following aconfirmed fire in the plant, thereby enhancing reliability and defense-in-depth with respect to maintainingthe availability of 125 VDC control power. No plant modifications or other procedure changes aredeemed necessary to address this issue.47 of 47 ENCLOSURE 2NINE MILE POINT UNIT 1UPDATED RESPONSES TOPROBABILISTIC RISK ASSESSMENT RAI 05 AND RAI 08Nine Mile Point Nuclear Station, LLCApril 30, 2013 ENCLOSURE2NINE MILE POINT UNIT 1UPDATED RESPONSES TO PROBABILISTIC RISK ASSESSMENT RAI 05 AND RAI 08By letter dated February 27, 2013, Nine Mile Point Nuclear Station, LLC (NMPNS) provided responsesto requests for additional information documented in the NRC's letter dated January 3, 2013. In theFebruary 27, 2013 letter, NMPNS committed to provide updates to the responses for Probabilistic RiskAssessment RAI 05 and Probabilistic Risk Assessment RAI 08 (if needed) to reflect the response to SafeShutdown / Circuit Analysis RAI 01, in which the definition of the Nine Mile Point Unit 1 (NMP1) safeand stable condition has been revised. Each NRC RAI is repeated (in italics), followed by the updatedNMPNS response. Changes to the responses are identified by revisions bars drawn in the right margin.Probabilistic Risk Assessment RAI 05Section 10 of NUREG/CR-6850, Supplement 1, states that a sensitivity analysis should be performedwhen using the fire ignition frequencies in the supplement instead of the fire ignition frequencies providedin Table 6-1 of NUREG/CR-6850. Provide the sensitivity analysis of the impact on using the supplementI frequencies instead of the Table 6-1 frequencies on core damage frequency (CDF), large early releasefrequency (LERF), delta (4)CDF, and ALERF for all of those bins that are characterized by an alpha thatis less than or equal to one. If the sensitivity analysis indicates that the change in risk acceptanceguidelines would be exceeded using the values in Table 6-1, justify not meeting the guidelines.Updated Response to Probabilistic Risk Assessment RAI 05The NMP 1 Fire PRA uses the ignition frequencies from the latest guidance related to fire PRAs as givenin Supplement 1 to NUREG/CR-6850. Supplement 1 to NUREG/CR-6850 (Section 10.2) addresses theuse of the ignition frequencies therein as follows:"The NRC accepts use of these revised fire bin ignition frequencies for fire PRAs conducted forNFPA-805 transition for best-/point-estimate calculations of fire risk (core damage frequency [CDF]and large early release frequency [LERF]), including delta-risk values from plant change evaluations,with the following provision. The fire PRA, including plant change evaluations, must also evaluatethe sensitivity of the risk and delta-risk results to evaluations performed using the current fire binignition frequencies in EPRI 1011989, NUREG/CR-6850, Chapter 6, "Fire Ignition Frequencies,"Table 6-1, "Fire Frequency Bins and Generic Frequencies," and Appendix C, "Determination ofGeneric Fire Frequencies," Table C-3, "Generic Fire Ignition Frequency Model for U.S. NuclearPower Plants." For those cases where the results from this sensitivity analysis indicate a change inthe potential risk significance associated with elements of the fire PRA or plant change evaluationsthat affects the decisions being made (e.g., what is acceptable with the new frequencies from EPRI1016735 might not be acceptable with the current applicable set from EPRI 1011989, NUREG/CR-6850), the licensee must address this situation by considering fire protection, or related, measures thatcan be taken to provide additional defense in-depth."With respect to the required sensitivity analysis, a footnote provides the following clarification:"The sensitivity analyses should be performed for a fire ignition frequency bin using the mean of thefire ignition frequency bins contained in NUREG/CR-6850. Furthermore, sensitivity analyses onlyneed to be performed for those bins characterized by an alpha from the EPRI 1016735 analysis that isless than or equal to 1. Note that an alpha value less than or equal to 1 is characteristic of a reverse-Jshaped probability density function, i.e., the same shape as the non-informative prior distributionsused in EPRI 1016735. This reverse-J shape is indicative of the large uncertainty in the bin fire1 of 7 ENCLOSURE2NINE MILE POINT UNIT 1UPDATED RESPONSES TO PROBABILISTIC RISK ASSESSMENT RAI 05 AND RAI 08frequency due to the sparsity of data for that bin, and therefore, the potential for significant changesshould the post-2000 fire event data differ significantly from the 1991-2000 data. The requiredsensitivity analysis is, for the purpose of this interim solution, judged to provide an adequateindication of the effects on risk and delta-risk in such a case."Results of the Sensitivity AnalysisTable PRA RAI 05-1 lists the ignition frequencies with alpha values < 1. Table PRA RAI 05-2 lists therisk results when the ignition frequencies from NUREG/CR-6850 are used, as well as the risk results asreported in the updated LAR Transition Report, which are based on Supplement 1 to NUREG/CR-6850ignition frequencies. Table PRA RAI 05-3 lists, at the fire area level, the risk results using ignitionfrequencies from NUREG/CR-6850 with alpha values < 1.An evaluation of the sensitivity results against Regulatory Guide 1.174 indicates that the delta risks byfire area in Table PRA RAI 05-3 using the NUREG/CR-6850 ignition frequencies meet the riskacceptance guidelines illustrated for Regions II and III of Figures 4 and 5 in Regulatory Guide 1.174 onan individual fire area basis. However, the total increase in risk associated with the implementation ofNFPA 805 for the overall plant calculated by summing the risk increases exceeds the acceptanceguidelines, as summarized in Table PRA RAI 05-2.Excluding fire risk, the plant risks associated with internal events, seismic, and high winds are estimatedfrom Table 1 in the Fire Risk Evaluations (FRE) Report (NI -FRE-FOO 1, Revision 0) and are also shownin Table PRA RAI 05-2 below. Summing those risks with the fire risks gives the total plant CDF andLERF including fire and other risks (Table PRA RAI 05-2). Total CDF and LERF including non-firerisks remain below the critical levels of 10-4 for CDF and 10-5 for LERF. However, the ACDF andALERF results exceed the delta risk guidelines (10-5 for CDF and 10-6 for LERF).Table PRA RAI 05-3 below lists the contribution to delta CDF and delta LERF for each fire area whenthe Fire PRA model is quantified using the frequencies from NUREG/CR-6850. The results indicate thatmost of the contribution to delta CDF is generated by Fire Area 05 (-81%) and Fire Area 11 (- 1I%).These two areas are the top contributors to LERF as well. Consistent with the guidance in Section 10.2 ofSupplement 1 to NUREG/CR-6850, fire protection, or related, measures that can be taken to providedefense in-depth for these three areas are discussed in the following paragraphs.Justification for Not Meeting the Guidelines with the Higher Ignition FrequenciesAs suggested in Section 10.2 of Supplement 1 to NUREG/CR-6850, NMPNS has identified fireprotection and related measures that provide additional defense-in-depth (DID), as justification for thesensitivity analysis results not meeting the delta risk guidelines. These measures are presented in the FireRisk Evaluation Report (Ni-FRE-FO01) and are summarized below for the fire areas that contribute themost to the calculated delta risk.Defense in Depth Measures for Fire Area 05Fire Area 05 is relatively large covering most of the turbine building above elevation 261'. This firearea can be classified in two groups of fire zones. The first group gathers those fire zones where thereis no installed automatic fire suppression system, or there is such a system but no credit is taken for it2 of 7 ENCLOSURE2NINE MILE POINT UNIT 1UPDATED RESPONSES TO PROBABILISTIC RISK ASSESSMENT RAI 05 AND RAI 08in the Fire PRA. This group consists of Fire Zones FBZT261S, OGI, OG2, OG3, TiA, T4B, T4D,T5A, T6A, T6B, T6C, T6D, T7A, T8A, and T8B.* There is no credit in the Fire PRA for the installed fire detection systems, automatic firesuppression systems, and manual suppression in Fire Zones FBZT261 S, T5A, T6C, T6D andT8A. The following systems are available in these zones and are credited for DID:Zone Detection Systems for DID Suppression Systems for DIDFBZT261S DA-2161E, DA-2161M WP-2161T5A D-2294, D-2304 SP-2314, SP-2324T6C D-2385, D-2395, D-2395FL, D- WD-2395FL2395PLT6D DA-2375 WP-2375T8A D-2445, D-2485 SP-2465* In Fire Zone TiA, there are no installed fire detection systems or automatic fire suppressionsystems, and no credit is taken in the Fire PRA for manual suppression. Manual fire suppressionby the fire brigade is credited for DID.* In Fire Zones OGI, OG2, T4D, T6B, T7A and T8B, no credit is taken for the installed firedetection system and subsequent manual suppression (no automatic suppression system isinstalled in these fire zones). The following systems are available and are credited for DID:Zone Detection Systems for DIDOGI D-7013OG2 D-7013T4D D-2194T6B D-2355, D-2405T7A DA-2425T8B D-2485" In Fire Zones OG3, T4B, and T6A, no credit is taken for the installed automatic fire suppressionsystem, but the Fire PRA credits the installed fire detection systems and subsequent manualsuppression (manual suppression in T6A is credited only for the structural steel fire scenario).The following systems are available and are credited for DID:Zone Suppression System(s) for DIDOG3 SP-7053T4B SP-2224, WP-2092T6A SP-2314, C-2365The second group of fire zones is made up of the balance of fire zones in the fire area; i.e.: TI, T3A,T3B, T4A, T4C, and FBZT261N. For these fire zones, the Fire PRA takes credit for installed firedetection systems and automatic suppression systems, as well as manual suppression. The local CO2fire suppression system is credited in TI, but with manual actuation only.3 of 7 ENCLOSURE 2NINE MILE POINT UNIT 1UPDATED RESPONSES TO PROBABILISTIC RISK ASSESSMENT RAI 05 AND RAI 08Defense in Depth Measures for Fire Area 11Fire Area II is equipped with a manual CO2 system (C-303 1) credited for DID in Fire Zone C2. FireZone C2 is the Auxiliary Control Room (or Relay Room) located under the main control room. Thissystem is not credited in the Fire PRA.Defense in Depth Measures Applicable for All Fire AreasThe governing procedures for fire protection activities are GAP-INV-02, "Control of Material StorageAreas," and GAP-FPP-02, "Control of Hot Work." These procedures are not credited explicitly in theFire PRA (i.e., the Fire PRA does not include failure probabilities to follow the requirements of theseprocedures) for postulating transient fires within Fire Area 05. The procedures are considered in theFire PRA consistent with the guidelines in NUREG/CR-6850 for selecting the appropriate credit forprompt suppression and hotwork manual suppression curve for the appropriate scenarios and fordetermining the influence factors serving as weighting factors for transient fire ignition frequencies.Consequently, the specific provisions of these procedures are credited for DID for: (1) controllingtransient combustibles throughout the plant; and (2) assigning compensatory measures to maintenanceactivities that may temporarily change the plant configuration.Effect of Response to Safe Shutdown / Circuit Analysis RAI 01LAR Section 4.2.1.2 originally defined the NMP1 safe and stable condition as "the ability to maintainK~ff< 0.99 with a reactor coolant temperature at or below the requirements for hot shutdown and thensubsequently cool down and maintain NMP1 in a cold shutdown condition." In the response to SafeShutdown / Circuit Analysis RAI 01, the definition of the NMP1 safe and stable condition for the"At-Power" analysis has been revised to hot shutdown. Analyses performed to support the response toSafe Shutdown / Circuit Analysis RAI 01 indicate an improvement to the delta risk numbers. Inparticular, the plant-level ACDF decreased from 1.52E-05/yr to 1.11E-05/yr, and the ALERFdecreased from 1.71 E-06/yr to 1.29E-06/yr.Table PRA RAI 05-1: Ignition Frequencies with Alpha Less than or Equal to 1T. 1 NUREG/CR FrequencySupp. 1 UE/R Ratio:Supp. 1 Ignition Source (Location) Supp. 1 Mean -6850 Mean NUREG/CRBin Alpha Frequency Frequency -6850 to(1 / y) (1 / y) -6850 toSupp.11 Batteries (Battery Room) 0.5 3.26E-04 7.5E-04 2.34 Main control board (Control Room) 1 8.24E-04 2.5E-03 3.09 Air Compressors (Plant-Wide) 0.5 4.65E-03 2.4E-03 0.511 Cable fires caused by welding and cutting 1 9.43E-04 2.OE-03 2.1(Plant-Wide)13 Dryers (Plant-Wide) 0.5 4.20E-04 2.6E-03 6.215.1 Electrical Cabinets Non-HEAF (Plant- 0.453 2.36E-02 4.5E-02 1.9Wide)22 RPS MG sets (Plant-Wide) 0.92 9.33E-04 1.6E-03 1.731 Cable fires caused by welding and cutting 0.5 4.50E-04 1.6E-03 3.6(Turbine Building)4 of 7 ENCLOSURE2NINE MILE POINT UNIT 1UPDATED RESPONSES TO PROBABILISTIC RISK ASSESSMENT RAI 05 AND RAI 08Table PRA RAI 05-2: Sensitivity Study ResultsResult with Supplement I Sensitivity Result withRisk Measure Ignition Frequencies NUREG/CR-6850Rkere (Ref. Updated LAR Table Ignition Frequencies(/yr) W-3, N1-FRE-F001,Revision 1)CDF (fire) 2.06E-05 3.29E-05CDF (other) 5.26E-06 5.26E-06CDF (total) 2.59E-05 3.82E-05LERF (fire) 2.23E-06 4.60E-06LERF (other) 2.07E-06 2.07E-06LERF (total) 4.30E-06 6.66E-06ACDF 8.47E-06 1.11E-05ALERF 7.05E-07 1.29E-06Table PRA RAI 05-3: Delta Risks by Fire AreaFire Irea ACDF ACDF ALERF ALERFFyr) Contribution (/yr) Contribution01 2.62E-07 2.37% 5.32E-08 4.12%02 1.08E-07 0.97% 3.20E-08 2.48%04 O.OOE+00 0.00% 1.60E-15 0.00%05 8.92E-06 80.69% 4.11E-07 31.83%06 1.26E-07 1.14% 1.27E-08 0.98%07 2.26E-07 2.04% 1.11E-08 0.86%09 2.32E-08 0.21% 1.48E-09 0.11%10 3.25E-08 0.29% 4.01E-09 0.31%II 1.26E-06 11.43% 7.01E-07 54.29%12 3.17E-11 0.00% 1.63E-09 0.13%13 4.34E-09 0.04% 3.87E-09 0.30%14 0.OOE+00 0.00% 0.OOE+00 0.00%15 1.29E-09 0.01% 6.OOE- 12 0.00%16A 0.OOE+00 0.00% 0.OOE+00 0.00%16B 0.OOE+00 0.00% 0.OOE+00 0.00%17A 0.OOE+00 0.00% 0.OOE+00 0.00%17B 0.00E+00 0.00% O.OOE+00 0.00%18 6.08E-10 0.01% 4.04E-10 0.03%19 8.84E-12 0.00% 1.49E-11 0.00%20 5.54E-11 0.00% 1.03E-11 0.00%21 5.82E-1 1 0.00% 1.04E-1 1 0.00%22 8.13E-08 0.74% 5.43E-08 4.21%23 6.50E-09 0.06% 4.34E-09 0.34%24 6.95E-10 0.01% 2.79E- 11 0.00%EXT N/A N/A N/A N/ASum 1.11E-05 100.00% 1.29E-06 100.00%5 of 7 ENCLOSURE2NINE MILE POINT UNIT 1UPDATED RESPONSES TO PROBABILISTIC RISK ASSESSMENT RAI 05 AND RAI 08Probabilistic Risk Assessment RAI 08The transition report describes and justifies an initial coping time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , after which, actions arenecessary to maintain safe and stable beyond 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Provide a discussion of the actions necessaryduring and beyond 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to maintain safe and stable conditions beyond 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months such as refillingfluidtanks or re-aligning systems. Evaluate quantitatively or qualitatively the risk associated with theseactions and equipment necessary to maintain safe and stable beyond 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months given the post-firescenarios during which they may be required.Updated Response to Probabilistic Risk Assessment RAI 08The PRA model uses a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months mission time for success criteria, similar to other PRAs and consistentwith ASME/ANS RA-Sa-2009. The plant must be in a safe stable state (e.g., hot shutdown condition)during this timeframe. Decay heat levels are lower after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of success (safe stable state withinventory control and heat removal) in the PRA model and offsite resources and recoveries are availablein case of any failures after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The probability of failures that may occur after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months andbeyond 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is considered negligible when other capabilities to recover are included. Those supportsystem dependencies in the PRA that are potentially sensitive to time have been evaluated. The followingsummarizes these considerations:" Condensate Storage Tank: This tank supports reactor pressure vessel (RPV) makeup from thefeedwater (high pressure coolant injection -HPCI) system and control rod drive (CRD) pumps, and isa source of emergency condenser makeup. Considering RPV makeup without the emergencycondensers, this tank is judged inadequate to last 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . With the emergency condensers and noloss of coolant accident (LOCA) condition, the tank may be adequate for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . There are 40,000gallons of makeup water available to the condensate storage tanks via gravity feed from thecondensate demineralizer water storage tank. Additional makeup would be required for a 72 hourmission time. Fire water make up to the emergency condenser makeup tanks is available. Damagerepair procedure Nl-DRP-OPS-001 has instructions for supplying the demineralized water storagetank from city water, service water, or fire water. The demineralized water storage tank can then bedrained to the condensate storage tanks by opening manual valve 57-31 located at Turbine Buildingelevation 305', column line J-1. The Fire PRA was also updated to model the recovery actions aimedat ensuring long-term EC makeup tank water supply via the diesel-driven fire pump andquantitatively evaluate their risk." 125V DC Power: Since emergency AC power is required, the batteries need only be available ondemand to support emergency diesel generator (EDG) starting and other initial start loads. As long asthe static charger and AC power are available after this battery demand, the batteries are not requiredin the long term. The batteries cannot supply DC loads for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months without AC power support.* EDG Fuel Supply: At full load, one EDG consumes 228 gallons of fuel oil per hour. Each EDG has a12,000 gallon fuel oil storage tank and a 400 gallon day tank. This would allow operation for morethan 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The fuel oil storage tanks can be cross connected to allow operation of one EDG atfull load for 4 days. Therefore, the EDG fuel oil supply will last for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .* Room Cooling: The only areas of concern in the PRA are the two EDG areas (roof fans and the rolldoor in each EDG room). All other areas were judged to have slow heat up rates and/or maximumtemperatures were sufficiently low.6 of 7 ENCLOSURE 2NINE MILE POINT UNIT 1UPDATED RESPONSES TO PROBABILISTIC RISK ASSESSMENT RAI 05 AND RAI 08The risks associated with activities occurring after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months have been evaluated qualitatively and areconsidered to be negligible and, thus, acceptable.LAR Section 4.2.1.2 originally defined the NMP 1 safe and stable condition as "the ability to maintain Kef< 0.99 with a reactor coolant temperature at or below the requirements for hot shutdown and thensubsequently cool down and maintain NMP1 in a cold shutdown condition." In response to SafeShutdown / Circuit Analysis RAI 01, the definition of the NMP1 safe and stable condition for the "AtPower" analysis has been revised to hot shutdown. This change led to the elimination of several VFDRswhich pertained to cold shutdown, and also led to the creation of new VFDRs associated with long-termwater supply to the EC makeup tanks. These new VFDRs were addressed by taking credit for recoveryactions whose feasibility was evaluated and the risk quantitatively evaluated in the Fire PRA.7 of 7 ENCLOSURE 3REVISIONS TO THE LAR TRANSITION REPORTWITH CHANGES HIGHLIGHTEDThe following are revisions to the Transition Report (included with the License Amendment Request(LAR) submitted by Nine Mile Point Nuclear Station, LLC (NMPNS) letter dated June 11, 2012)resulting from the responses to NRC requests for additional information (RAI) Safe Shutdown / CircuitAnalysis RAI 01 and Safe Shutdown / Circuit Analysis RAI 03. The revised Transition Report pages,with the changes highlighted to facilitate their identification, are as noted below.* Sections 4.2 and 4.3 (Pages 14 through 29a)* Table 4-3 (Pages 54 through 60)* Attachment A (Pages A-42 through A-44)" Attachment B (Pages B-I through B-102)" Attachment F (Pages F-6 and F-7)" Attachment G (Pages G-1 through G-41)Nine Mile Point Nuclear Station, LLCApril 30, 2013 REVISIONS TO TRANSITION REPORTSECTION 4.2, NUCLEAR SAFETY PERFORMANCE CRITERIASECTION 4.3, NON-POWER OPERATIONAL MODESPages 14 through 29a with changes highlighted.

Constellation Energy Nuclear Group4.0 Compliance with NFPA 805 Requirementsstructures housing equipment required for nuclear plant operations are considered as"power block" structures.These structures are listed in Attachment I and define the "power block" and "plant".4.2 Nuclear Safety Performance CriteriaThe Nuclear Safety Performance Criteria are established in Section 1.5 of NFPA 805.Chapter 4 of NFPA 805 provides the methodology to determine the fire protectionsystems and features required to achieve the performance criteria outlined in Section1.5. Section 4.3.2 of NEI 04-02 provides a systematic process for determining theextent to which the pre-transition licensing basis meets these criteria and for identifyingany necessary fire protection program changes. NEI 04-02, Appendix B-2 providesguidance on documenting the transition of Nuclear Safety Capability AssessmentMethodology and the Fire Area compliance strategies.4.2.1 Nuclear Safety Capability Assessment MethodologyThe Nuclear Safety Capability Assessment (NSCA) Methodology review consists of fourprocesses:" Establishing compliance with NFPA 805 Section 2.4.2" Establishing the Safe and Stable Conditions for the Plant" Establishing Recovery Actions" Evaluating Multiple Spurious OperationsThe methodology for demonstrating reasonable assurance that a fire during non-poweroperational (NPO) modes will not prevent the plant from achieving and maintaining thefuel in a safe and stable condition is an additional requirement of 10 CFR 50.48(c) andis addressed in Section 4.3.4.2.1.1 Compliance with NFPA 805 Section 2.4.2Overview of ProcessNFPA 805 Section 2.4.2 Nuclear Safety Capability Assessment states:"The purpose of this section is to define the methodology for performing anuclear safety capability assessment. The following steps shall be performed:(1) Selection of systems and equipment and their interrelationships necessary toachieve the nuclear safety performance criteria in Chapter 1(2) Selection of cables necessary to achieve the nuclear safety performancecriteria in Chapter 1(3) Identification of the location of nuclear safety equipment and cables(4) Assessment of the ability to achieve the nuclear safety performance criteriagiven a fire in each fire area"The NSCA methodology review evaluated the existing post-fire safe shutdown analysis(SSA) methodology against the guidance provided in NEI 00-01, Revision 2, Chapter 3,"Deterministic Methodology," as discussed in Appendix B-2 of NEI 04-02. NMP1 usedthe guidance provided in NEI 00-01, Revision 2 because it is endorsed as anNMI, pil213Pge1INMP1, April 2013Page 14 1 Constellation Energy Nuclear Group4.0 Compliance with NFPA 805 Requirementsacceptable methodology in NRC RG 1.205 and due to feedback received as a result ofNRC requests for additional information on other post-pilot plant LARs.The methodology is depicted in Figure 4-2 and consisted of the following activities:" Each specific subsection of NFPA 805 Section 2.4.2 was correlated to thecorresponding section of Chapter 3 of NEI 00-01, Revision 2. Based upon thecontent of the NEI 00-01 methodology statements, a determination was made ofthe applicability of the section to the station." The plant-specific methodology was compared to applicable sections of NEI00-01 and one of the following alignment statements and its associated basiswere assigned to the section:o Alignso Aligns with Intento Not in Alignmento Not in Alignment, but Prior NRC Approvalo Not in Alignment, but no adverse consequences" For those sections that do not align, an assessment was made to determine if thefailure to maintain strict alignment with the guidance in NEI 00-01 could haveadverse consequences. Since NEI 00-01 is a guidance document, portions of itstext could be interpreted as 'good practice' or intended as an example of anefficient means of performing the analyses. If the section has no adverseconsequences, these sections of NEI 00-01 can be dispositioned without furtherreview.In addition, a review of NEI 00-01, Revision 3 was conducted against the guidance fromNEI 00-01, Revision 2. There were no gaps relative to MSOs identified.The comparison of the NMP1 existing post-fire SSA to NEI 00-01 Chapter 3 (NEI 04-02Table B-2) was performed and documented in EIR 51-9133191, Nine Mile Point Unit I -Nuclear Safety Capability Assessment.Results from Evaluation ProcessThe method used to perform the NSCA with respect to selection of systems andequipment, selection of cables, and identification of the location of equipment andcables, either meets the NRC endorsed guidance from NEI 00-01, Revision 2, Chapter3 (as supplemented by the gap analysis to Revision 3) directly or meets the intent of theendorsed guidance with adequate justification as documented in Attachment B.Referenced documents are planned as being retained as post-transition documents.NEI 00-01, Revision 2, Chapter 3 contains guidance criteria concerning identifyingrequired and important to SSD components. These specific guidance criteria are notapplicable to plants transitioning to NFPA 805; therefore, they were not addressed forNMPI.NMPI, April 2013 Pagel1 IA Constellation Energy Nuclear Group4.0 Compliance with NFPA 805 RequirementsConstellation Energy Nuclear Group 4.0 Compliance with NFPA 805 RequirementsStep I Assemble DocumentatlonLIpDeternnne and DocumentStep 2 Applicability of NEI 00-01SectionsFrApialFor Applicable NEI 00-01Sections, Perform Compa idsonof SSD Method vs. NEI 00-01No No Yese N4.2.1.2 Safe and Stable Conditions for the PlantOverview of ProcessThe nuclear safety goals, objectives and performance criteria of NFPA 805 allow moreflexibility than the previous deterministic programs based on 10 CFR 50 Appendix Rand NUREG-0800, Section 9.5.1 (and NEI 00-01, Chapter 3), since NFPA 805 onlyrequires the licensee to maintain the fuel in a safe and stable condition rather thanachieve and maintain cold shutdown.NFPA 805, Section 1.6.56, defines Safe and Stable Conditions as follows"For fuel in the reactor vessel, head on and tensioned, safe and stable conditionsare defined as the ability to maintain Keff <0.99, with a reactor coolant temperature ator below the requirements for hot shutdown for a boiling water reactor and hotstandby for a pressurized water reactor. For all other configurations, safe and stableconditions are defined as maintaining Keff <0.99 and fuel coolant temperature belowboiling."The nuclear safety goal of NFPA 805 requires "...reasonable assurance that a fireduring any operational mode and plant configuration will not prevent the plant fromachieving and maintaining the fuel in a safe and stable condition" without a specificreference to a mission time or event coping duration.For the plant to be in a safe and stable condition, it may not be necessary to perform atransition to cold shutdown as currently required under 10 CFR 50, Appendix R.Therefore, the unit may remain at or below the temperature defined by a hotstandby/hot shutdown plant operating state for the event.NMPI, April 2013 Page 16 IINMP1, April 2013Page 16 1 Constellation Energy Nuclear Group4.0 Compliance with NFPA 805 RequirementsConstellation Energy Nuclear Group 4.0 Compliance with NFPA 805 RequirementsResultsDemonstration of the Nuclear Safety Performance Criteria for safe and stable conditionswas performed in two analyses." At-Power analysis for potential fires while in the Power Operating Conditions(Reactor mode switch is in "Startup" or "Run" position and the reactor is critical orcriticality is possible due to control rod withdrawal) or Shutdown Condition -Hot(Reactor mode switch is in "Shutdown" position and reactor coolant temperatureis greater than 212°F), but not on shutdown cooling mode of decay heatremoval.incl!udin*g .ta.tup and run. This analysis is discussed in Section 4.2.4." Non-Power analysis for potential fires while in Shutdown Condition -Hot andlower operating conditions., which 'includoc Hot Shutdown and below.. Thisanalysis is discussed in Section 4.3.Based on the EIR 51-9133191, Nine Mile Point Unit I -Nuclear Safety CapabilityAssessment, the NFPA 805 licensing basis for a safe and stable condition is defined asthe ability to maintain Keff < 0.99 with a reactor coolant temperature at or below therequirements for hot shutdown. and- subsoquontly cool down -and- Maintain PIAI120in a cold ,hutdown condition.The At-Power analysis includes a primary and alternate means of achieving andmaintaining safe and stable conditions. The primary means of achieving andmaintaining Hot Shutdown (HSD) is via the Emergency Cooling system. Either of thetwo redundant Emergency Cooling decay heat removal loops can achieve and maintainHSD. The Emergency Cooling system operates by natural circulation where steamflows upward to the condenser(s) and returns as condensate to the Reactor PressureVessel (RPV). Decay heat is removed through the transfer of heat from the reactorcoolant to the shell side water of the Emergency Condenser(s),.G...l. which vents thedeveloped steam to atmosphere. Operation of either Emergency Cooling loop cansustain HSD conditions for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months without the need for makeup. from the CondncsatoSto.ago Tank (-ST). Upon achieving HSD conditions, the plant is able to maintain safeand stable operation for an extended period of time using the Emergency Coolingsystem. After 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , the Emergency Condenser makeup tanks can be replenished asneeded using the diesel driven fire pump (DFP), which draws water from Lake Ontario(effectively an infinite source). Periodic refueling of the DFP is accomplished inaccordance with existing plant procedures using an on-site fuel source. Reactor coolantmakeup is required after 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , assuming a nominal Technical Specification leakagerate of 25 gpm. Makeup is provided via the Control Rod Drive (CRD) system using oneof the CRD pumps drawing suction from the CST. AC power is required to operate aCRD pump. The diesel driven fire pump may be aligned to provide primary makeup inthe event no CRD pump is available.The Emergency Cooling system can be initiated either manually or automatically. TheRPS instruments and logic that automatically initiate the Emergency Cooling system onhigh reactor pressure or low-low reactor level have been included in the analysis.Manual initiation of the Emergency Cooling system can be accomplished from either theControl Room or Remote Shutdown Panel (RSP) if Control Room abandonment isNMPI, April 2013 Page 17 IINIWIPI, April 2013Page 17 1 Constellation Energy Nuclear Group4.0 Compliance with NFPA 805 Requirementsnecessary, depending on the fire location. AC power is not required to manually initiateDecay Heat Removal (DHR) via the Emergency Cooling system.In the event the primary means of decay heat removal during HSD method-is notavailable (i.e., Emergency Cooling system not available), the plant can be maintained inHSD by opening three Electromagnetic Relief Valves (ERVs) in the automaticdepressudzation system (ADS) and blowing steam to the Torus to reduce pressure.When reactor pressure reaches approximately 365 psig, Core Spray (CS) may beutilized to provide core cooling. AC and DC electrical power are required for thismethod of decay heat removal.The ADS system can be initiated either manually or automatically. The RPSinstruments and logic automatically initiate the ADS system on a combination of low-low-low reactor level and high drywell pressure.The CS system can be initiated either manually or automatically. The RPS instrumentsand logic automatically initiate the CS system on low-low reactor level or high drywellpressure.The instrumentation and logic circuitry that automatically initiates ADS and CS havebeen included in the analysis. Manual initiation of the ERVs and CS system isaccomplished from the Control Room. In the event spurious actuations of the ERVsNMPI, April 2013 Page 17a IINMP1, April 2013Page 17a I Constellation Energy Nuclear Group4.0 Compliance with NFPA 805 Requirementstake place due to a Control Room fire, CS can be initiated manually from outside theControl Room if Control Room abandonment is necessary.The oif .aWhiviAng and maintaining Cold Shutdown (CSD) i6s via theS-hubt-d-OWA Cooling SYMStm (SDC). WAhen roactor presr isrduced to 120 psig andrnn_ r* a.rnG # ~ nm

e-n, ;- e-A'AGO +kn i +n a kn +rr r +;m^Ainn, *n t" k ~Iinima~rRa[I*3f TkA Qflr' n ni4,nA ki., Danr Ei, inun inet f'Innn,q I e-en Mn M(RBCLC=) and Eimergoncy Sendaio Wiato W.OoSCpm nascainwtRBCLCG and ESW ~srqio oachioVo and maintain C-SD cnitions. AC Gpoworiroquirod to 8ntitoGD.Undor thsscnriractor coolan-t makeupis reud after R hoursm with an assumedDove.' (CRD) system usng9 of the C-RD1 pumps draw~ing suctionn fr-m. t-ho CsT. AC-rrqu o or a iE6 pump. iT no iesel nHr pump may; aso healgedt proid makeup in the e':ent no CRD pump is, av.ailaboThe alternate means of decay heat removal can be used to maintain safe and stableconditions until such time that the Shutdown Cooling (SDC) system is placed in service.in the event the pnimar'; CSD method is not av.ailable, the plant can be coolo-d dowAMn to-C.S.D using the CS syr.. m. CS is a two loop system. Operation of one loop is adequateto ensure core cooling.ahiove-CSD-. When utilizing CS, the reactor vessel eventuallyfloods to the point where the ERVs are passing fluid to the Torus rather than steam, inessence placing the RCS in recirculation through the Torus. During this process, decayheat is removed by operation of the Containment Spray (CTS) system in conjunctionwith the Containment Spray Raw Water (CTSRW) system. This mo'thod will bring thplant to CSG .Fully flooding the RPV negates the need for another system toprovide inventory makeup. AC power is required to initiate this method.For either the primary or alternate means of achieving and maintaining safe and stableconditions, AC power is available from either the station Emergency Diesel Generators(EDGs) or offsite power. Actions required to achieve and initially maintain hot shutdownconditions can be performed by the minimum shift complement consisting of reactoroperators, senior reactor operators, and non-licensed plant operators. The EDGs canbe refueled in accordance with existing plant procedures using an on-site fuel source(tanker truck), until such time that offsite power is restored. Additional resources fromthe emergency response organization will be available to support EDG refuelingactivities.LGIng TerM Safe and rStabl s ill baien;ntaiinl using either the preferred otalIte-rna teA CSD m ethod. Wit AC h e~e availableM froinm either the stdationn Em-FeFgencDiesel Generato (EDGs) Vfite power, C5[D nscan be maintained4.2.1.3 Establishing Recovery ActionsOverview of ProcessNEI 04-02 and RG 1.205 suggest that a licensee submit a summary of its approach foraddressing the transition of OMAs as recovery actions in the LAR (Regulatory Position2.21 and NEI-04-02, Section 4.6). As a minimum, NEI 04-02 suggests that theINMPI, April 2013Page 18 1 Constellation Energy Nuclear Group4.0 Compliance with NFPA 805 Requirementsassumptions, criteria, methodology, and overall results be included for the NRC todetermine the acceptability of the licensee's methodology.The discussion below provides the methodology used to transition pre-transition OMAsand to determine the population of post-transition recovery actions. This process isbased on FAQ 07-0030 (ML1 10070485) and consists of the following steps:* Step 1: Clearly define the primary control station(s) and determine which pre-transition OMAs are taken at primary control station(s) (Activities that occur in theMain Control Room are not considered pre-transition OMAs). Activities that takeplace at primary control station(s) or in the Main Control Room are not recoveryactions, by definition.NMPI, April 2013 Page 18a IINIVIPI, April 2013Page 18a I Constellation Energy Nuclear Group4.0 Compliance with NFPA 805 Requirements" Step 2: Determine the population of recovery actions that are required to resolvevariances from deterministic requirements (VFDRs) (to meet the risk acceptancecriteria or maintain a sufficient level of defense-in-depth)." Step 3: Evaluate the additional risk presented by the use of recovery actionsrequired to demonstrate the availability of a success path." Step 4: Evaluate the feasibility of the recovery actions." Step 5: Evaluate the reliability of the recovery actions.ResultsThe review results are documented in EIR 51-9156521, Recovery Action Review forNine Mile Point Nuclear Power Station Unit I Transition to NFPA 805. Refer toAttachment G for the detailed evaluation process and summary of the results from theprocess.4.2.1.4 Evaluation of Multiple Spurious OperationsOverview of ProcessNEI 04-02 suggests that a licensee submit a summary of its approach for addressingpotential fire-induced MSOs for NRC review and approval. As a minimum, NEI 04-02suggests that the summary contain sufficient information relevant to methods, tools, andacceptance criteria used to enable the NRC to determine the acceptability of thelicensee's methodology. The methodology utilized to address MSOs for NMP1 issummarized below.As part of the NFPA 805 transition project, a review and evaluation of NMP1susceptibility to fire-induced MSOs was performed. The process was conducted inaccordance with NEI 04-02 and RG 1.205, as supplemented by FAQ 07-0038 Revision3 (ML1 10140242). The BWR Generic MSO list in NEI 00-01, Revision 2, dated June 5,2009 (including a gap analysis to the Revision 3 Generic MSO list) was utilized.The approach outlined in Figure 4-3 (based on Revision 3 from FAQ 07-0038) is oneacceptable method to address fire-induced MSOs. This method used insights from theFire PRA developed in support of transition to NFPA 805 and consists of the following:" Identifying potential MSOs of concern." Conducting an expert panel to assess plant specific vulnerabilities (e.g., per NEI00-01, Rev. 1 Section F.4.2)." Updating the Fire PRA model and existing post-fire SSA to include the MSOs ofconcern." Evaluating for NFPA 805 compliance." Documenting results.This process is intended to support the transition to a new licensing basis. Post-transition changes would use the RI-PB change process. The post-transition changeprocess for the assessment of a specific MSO would be a simplified version of thisprocess, and may not need the level of detail shown in the following section (e.g., Anexpert panel may not be necessary to identify and assess a new potential MSO.lNMPI, April 2013Page 19 1 Constellation Energy Nuclear Group4.0 Compliance with NFPA 805 RequirementsConstellation Energy Nuclear Group 4.0 Compliance with NFPA 805 RequirementsIdentification of new potential MSOs may be part of the plant change review processand/or inspection process).Identify Potential MVSOs of Concern*SSAStep 1 Generic List of MSOsSelf AssessmentsPRA InsightsOperating ExperienceExpert PanelStep 2 Identify and Document MSOs ofConcernUpdate PRA model & NSCA (asappropriate) to include MSOs ofconcernStep 3

ID equipment* ID logical relationships* ID cables* ID cable routingStep 4 ComolianNo Pursue other resolution optionsStep 5 Document ResultsFigure 4-3 -Multiple Spurious Operations -Transition Resolution Process(Based on FAQ 07-0038)ResultsRefer to Attachment F for the process used for NMP1 and the results from the process.4.2.2 Existing Engineering Equivalency Evaluation TransitionOverview of Evaluation ProcessThe EEEEs that support compliance with NFPA 805 Chapter 3 or Chapter 4 (both thosethat existed prior to the transition and those that were created during the transition)were reviewed using the methodology contained in NEI 04-02. The methodology forperforming the EEEE review includes the following determinations:0 The EEEE is not based solely on quantitative risk evaluations,INMP1, April 2013Page 20 1 Constellation Energy Nuclear Group4.0 Compliance with NFPA 805 Requirements" The EEEE is an appropriate use of an engineering equivalency evaluation," The EEEE is of appropriate quality," The standard license condition is met," The EEEE is technically adequate," The EEEE reflects the plant as-built condition, and" The basis for acceptability of the EEEE remains validIn accordance with the guidance in RG 1.205, Regulatory Position 2.3.2, and NEI 04-02,as clarified by FAQ 07-0054, Demonstrating Compliance with Chapter 4 of NFPA 805,EEEEs that demonstrate that a fire protection system or feature is 'adequate for thehazard' are summarized in the LAR as follows:" If not requesting specific approval for 'adequate for the hazard' EEEEs, then theEEEE was referenced where required and a brief description of the evaluatedcondition was provided." If requesting specific NRC approval for 'adequate for the hazard' EEEEs, thenthe EEEE was referenced where required to demonstrate compliance and wasincluded in Attachment L for NRC review and approval.In all cases, reliance on EEEEs to demonstrate compliance with NFPA 805requirements is documented in the LAR.ResultsThe review results for EEEEs are documented in EIR 51-9077683, NFPA 805Fundamental Fire Protection Program and Design Elements Transition Review.In accordance with the guidance provided in RG 1.205, Regulatory Position 2.3.2, andNEI 04-02, as clarified by FAQ 07-0054, EEEEs used to demonstrate compliance withChapters 3 and 4 of NFPA 805 are referenced in Attachments A and C as appropriate.None of the transitioning EEEEs require NRC approval.4.2.3 Licensing Action TransitionOverview of Evaluation ProcessThe existing licensing actions (exemptions / safety evaluations) review was performedin accordance with NEI 04-02. The methodology for the licensing action reviewincluded the following:" Determination of the bases for acceptability of the licensing action." Determination that these bases for acceptability are still valid and required forNFPA 805.ResultsAttachment K contains the detailed results of the Licensing Action Review.The following licensing actions will be transitioned into the NFPA 805 fire protectionprogram as previously approved (NFPA 805 Section 2.2.7). These licensing actions areconsidered compliant under 10 CFR 50.48(c).lNMP1, April 2013Page 21 1 Constellation Energy Nuclear Group4.0 Compliance with NFPA 805 Requirements* NoneThe following licensing actions are no longer necessary and will not be transitioned intothe NFPA 805 fire protection program:" An exemption from the requirements of Section III.G.2 of Appendix R for thebattery board rooms (FA 16A and FA 16B), since their boundary walls do notprovide the required 3-hour rated barriers." An exemption from the requirements of Section III.G.2 of Appendix R for thebattery rooms (FA 17A and FA 177B), since their boundary walls do not providethe required 3-hour rated barriers." An exemption from the requirements of Section III.G of Appendix R for thecontrol room (FA 11), since the control room ceiling does not have a 3-hourrating from the control room side due to unprotected structural steel members." An exemption from the requirements of Section Ill.G.2 of Appendix R for the wallbetween the reactor building and the turbine building above elevation 340' (FA 1,FA 2, and FA 5), since the wall is not a 3-hour rated barrier." An exemption from the requirements of Section III.G.2 of Appendix R for the firebreak zone separating FA 1 and FA 2 in the reactor building upper level(elevation 340'), since the wall is not a 3-hour rated barrier.These exemptions are no longer required because the subject boundaries have beendemonstrated adequate for the hazard in an EEEE.Since the exemptions are either compliant with 10 CFR 50.48(c) or no longernecessary, in accordance with the requirements of 10 CFR 50.48(c)(3)(i), NMPNSrequests that the exemptions listed in Attachment K be rescinded as part of the LARprocess. See Attachment 0, Orders and Exemptions.4.2.4 Fire Area TransitionOverview of Evaluation ProcessThe Fire Area Transition (NEI 04-02 Table B-3) was performed using the methodologycontained in NEI 04-02 and FAQ 07-0054. The methodology for performing the FireArea Transition, depicted in Figure 4-4, is outlined as follows:Step 1 -Assembled documentation. Gathered industry and plant-specific fire areaanalyses and licensing basis documents.Step 2 -Documented fulfillment of nuclear safety performance criteria." Assessed accomplishment of nuclear safety performance goals. Documentedthe method of accomplishment, in summary level form, for the fire area." Documented evaluation of effects of fire suppression activities. Documented theevaluation of the effects of fire suppression activities on the ability to achieve thenuclear safety performance criteria." Performed licensing action reviews. Performed a review of the licensing aspectsof the selected fire area and documented the results of the review. See Section4.2.3.INMP1, April 2013Page 22 1 Constellation Energy Nuclear Group4.0 Compliance with NFPA 805 Requirements" Performed existing engineering equivalency evaluation reviews. Performed areview of existing engineering equivalency evaluations (or created newevaluations) documenting the basis for acceptability. See Section 4.2.2." Performed a review of pre-transition OMAs to determine those actions takingplace outside of the main control room or outside of the primary control station(s).See Section 4.2.1.3.Step 3 -VFDR identification, characterization and resolution considerations. Identifiedvariances from the deterministic requirements of NFPA 805, Section 4.2.3.Documented variances as either a separation issue or a degraded fire protectionsystem or feature. Developed VFDR problem statements to support resolution.Step 4 -Performance-Based evaluations (Fire Modeling or Fire Risk Evaluations). SeeSection 4.5.2 for additional information.Step 5 -Final Disposition." Documented final disposition of the VFDRs in Attachment C (NEI 04-02Table B-3)." For recovery action compliance strategies, ensured the manual action feasibilityanalysis of the required recovery actions was completed. Note: If a recoveryaction cannot meet the feasibility requirements established per NEI 04-02, thenalternate means of compliance were considered." Documented the post transition NFPA 805 Chapter 4 compliance basis.Step 6 -Documented required fire protection systems and features. Reviewed theNFPA 805 Section 4.2.3 compliance strategies (including fire area licensing actions andengineering evaluations) and the NFPA 805 Section 4.2.4 compliance strategies(including simplifying deterministic assumptions) to determine the scope of fireprotection systems and features 'required' by NFPA 805 Chapter 4. The 'required' fireprotection systems and features are subject to the applicable requirements of NFPA805 Chapter 3.NMPI, April 2013 Page 23 IINMP1, April 2013Page 23 1 Constellation Energy Nuclear Group4.0 Compliance with NFPA 805 RequirementsIdentify INITIAL VariancesFromDeterminisic Requirementsof NFPA 805 § 4.2.3(6-3 Table)Document Final Disposition ofVFDRCompliance options include:Accept As IsRequire FP systemslfeaturesRequire Recovery ActionRequire ProgrammaticEnhancementsRequire Plant Modifications(B-3 Table)Figure 4-4 -Summary of Fire Area Review[Based on FAQ 07-0054 Revision 1]NMPI, April 2013 Page 24 IINMP1, April 2013Page 24 1 Constellation Energy Nuclear Group4.0 Compliance with NFPA 805 RequirementsResults of the Evaluation ProcessAttachment C contains the results of the Fire Area Transition review (NEI 04-02 TableB-3). On a fire area basis, Attachment C summarizes compliance with Chapter 4 ofNFPA 805.NEI 04-02 Table B-3 includes the following summary level information for each fire area:" Regulatory Basis -NFPA 805 post-transition regulatory bases." Performance Goal Summary -An overview of the method of accomplishment ofeach of the performance criteria in NFPA 805 Section 1.5." Reference Documents -Specific references to Nuclear Safety CapabilityAssessment Documents." Licensing Actions -Specific references to safety evaluations that will remain partof the post-transition licensing basis. A brief description of the condition and thebasis for acceptability of the licensing action are provided. In addition,summaries of Fire Risk Evaluations performed for variances from thedeterministic requirements are also provided." EEEE -Specific references to EEEE that rely on determinations of "adequate forthe hazard" that will remain part of the post-transition licensing basis. A briefdescription of the condition and the basis for acceptability are provided." VFDRs -Specific variances from the deterministic requirements of NFPA 805Section 4.2.3. Refer to Section 4.5.2 for a discussion of the performance-basedapproach.4.3 Non-Power Operational Modes4.3.1 Overview of Evaluation ProcessNMP1 implemented the process outlined in NEI 04-02 and FAQ 07-0040, "Non-PowerOperations Clarifications." The goal (as depicted in Figure 4-5) is to ensure thatcontingency plans are established when the plant is in a Non-Power Operational (NPO)mode where the risk is intrinsically high. During low risk periods, normal riskmanagement controls and fire prevention/protection processes and procedures will beutilized.The process to demonstrate that the nuclear safety performance criteria are met duringNPO modes involved the following steps:" Review of the existing Outage Management Processes" Identification of Equipment/Cables:o Review of plant systems to determine success paths that support each of thedefense-in-depth Key Safety Functions (KSFs), ando Identification of cables required for the selected components anddetermination of their routing." Perform Fire Area Assessments (identify pinch points -plant locations where asingle fire may damage all success paths of a KSF)." Manage pinch-points associated with fire-induced vulnerabilities during theoutage.!NMPI, April 2013Page 25 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 RequirementsThe process is depicted in Figures 4-5 and 4-6. The results are presented in Section4.3.2.Figure 4-5 Review POSs, KSFs, Equipment, and Cables, and Identify Pinch PointsNMPI, April 2013 Page 26 IINIVIP1, April 2013Page 26 1 Constellation Energy Nuclear Group4.0 Compliance with NFPA 805 RequirementsConstellation Energy Nuclear Group 4.0 Compliance with NFPA 805 RequirementsHigher Risk Evolution as Defined by Plant SpecificOutage Risk Criteria for example1) Time to Boil2) Reactor Coolant System and Fuel Pool Inventory3) Decay Heat RemovalFigure 4-6 Manage Pinch PointsNMPI, April 2013 Page 27 IlNMP1, April 2013Page 27 1 Constellation Energy Nuclear Group4.0 Compliance with NFPA 805 RequirementsConstellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements4.3.2 Results of the Evaluation ProcessBased on FAQ 07-0040, the Plant Operating States (POSs) considered for equipmentand cable selection are defined in EIR 51-9171174, Nine Mile Point 1, Non-PowerOperations KSF Equipment List. The methodology for determination of KSFs, successpaths, components required to achieve the success paths and their associated cablingarewere defined in EIR 51-9137629, Nine Mile Point 1 -Nuclear Power Station -NFPA805 Transition -Non Power Operations Component Pinch Point Analysis. Thesedocuments provide the component selection information for the POSs included in theanalysis (HSD, CSD, Refueling and Defueled conditions). Systems and componentswere identified to support the following KSFs:* Reactor Vessel Decay Heat Removal Capability* Spent Fuel Pool Cooling* Reactor Vessel Inventory Control* Spent Fuel Pool Inventory Control* Electrical Power Availability* Reactivity ControlKSFs are evaluated in each fire zone. Each KSF has one or more paths that satisfy thespecific function. The process for selection and treatment of components is consistentwith the methodology in the NSCA. Where new components are required to supportKSFs, the components are included in NPO equipment list and cable identification isperformed using the same methodology as that employed for the NSCA. Where NSCAequipment is also relied upon in the NPO analysis but the NPO functional requirementsdiffered from that in the NSCA, additional reviews are performed to ensurecomprehensive cable selection. Inherent in the process is identification of componentspotentially vulnerable to single and multiple spurious operation concerns during NPO.Note that the Reactivity Control KSF is not included in the NPO analysis because it isadministratively controlled in accordance with procedure NIP-OUT-01.: DHR for boththe R-e-mctor and the Spent Fuel Pool (SFP), ... ..,,tr.y Control for- bth theVeosel and the SF12, and Po wo r avai lability. Each m1ay have one or moIre 15SMpathr, that satisfy thatoifcKFNo effort iswas made to eliminate or reduce fire impact by circuit analysis; therefore, aconservative estimate of damage is provided, including hot-short induced spuriousoperation of equipment. By assuming that a single fire impacts any and all componentsin a fire zone, (whether the individual component or its associated cables are physicallylocated within the fire zone), the assumption is made that the entire contents of the firezone are lost.EIR 51-9137629 contains the fire zone evaluations comprising the 'KSF pinch point'analysis. If a component that is part of a particular KSF flew-path is impacted, it isassumed that the KSF path is lost. -However, there are normallymay-be one or moreother flew-paths within the particular KSF that are not impacted; therefore, the KSF isnot considered lost and does not constitute a pinch point. Only when all paths for aparticular KSF are impacted-, is the KSF itself considered lost and identified is-as a pinchpoint.NMPI, April 2013 Page 28 IINMPI, April 2013Page 28 1 Constellation Energy Nuclear Group4.0 Compliance with NFPA 805 RequirementsThe NPO analysis results (EIR 51-9137629 and EIR 51-9171174 ) categorize each KSF(in each fire zone), as either 'T", "L" or "N" as follows:T "I" (Impacted): At least one of the KSF paths associated with a giventhe KSF isaffected, i.e., a component of a specific the-KSF path or any of the component'srequired cablesits aSSOciatod cables within the fire zone are impacted, wherebythat path can no longer be assured of being functional. However, at least oneother KSF path for the within that-KSF remains is-still-available." "L" (Lost): All available success paths for a given KSF are impacted." "N" (None): No impacts to the KSF are identified."Pinch Points" are were-thenidentified (on a fire zone basis), based on the completeloss of a KSF. In accordance with FAQ 07-0040, any evaluated area in which all of thecredited success paths for a given KSF are lost is considered a KSF pinch point. EachKSF for all Fire Zones is evaluated and documented in the pinch point analysis. TheFire Zone is labeled as AP,-"N" if in the pinch POint ,olumn indicates that no KSFsarewere lost. in th;s firAe zone. The Fire Zone is labeled with aA "Y' in this "w'-mnifddioates4hat-if one or more KSFs arewere lost, thereby identifying that the Fire Zonecontains one or more pinch points.in this fire Zone and therefeo-e, constitutes, a pinchNUMARC 91-06 discusses the development of outage plans and schedules. A keyelement of that process is to ensure the KSFs perform as needed during the variousoutage evolutions. During outage planning, the NPO Fire Zone Assessment is reviewedto identify areas of single-point KSF vulnerability during HRE to develop neededcontingency plans/actions. Depending upon the significance of the damage for thoseareas, combinations of the following options to reduce fire risk are considered at aminimum:INMPI, April 2013 Page 28a I!NMP1, April 2013Page 28a I'A Constellation Energy Nuclear Group4.0 Compliance with NFPA 805 RequirementsConstellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements" Prohibition or limitation of hot work in fire areas during periods of increasedvulnerability;" Verification of functional fire detection and/or suppression in the vulnerableareas;" Prohibition or limitation of combustible materials in fire areas during periods ofincreased vulnerability;" Plant configuration changes (e.g., removing power from equipment once it isplaced in its desired position);" Provision of additional fire patrols at periodic intervals or other appropriatecompensatory measures (such as surveillance cameras) during increasedvulnerability;" Use of recovery actions to mitigate losses of Key Safety Functions;" Reschedule the work to a period with lower risk or higher DID;" Identification and monitoring in-situ ignition sources for fire precursors.In addition, KSF equipment removed from service during the HREs is evaluated. Theevaluation is based on KSF equipment availability and NPO Fire Zone Assessment forany necessary contingency plans/actions. The recommendations for reducing risk canbe applied as appropriate during the implementation phase (as part of the technicaldocument and procedure review). NMP1 strategies for reducing NPO fire risk do notrely on Recovery Actions or Pre-Fire Actions (i.e., Plant Configuration Changes) asstrategies for reducing NPO fire risNote that roc .'"or; acti.n. wor. not. uod- toFiiAtolmno firomd, ucod failuroc.See Attachment D for more complete details. Based on incorporation of therecommendations from the KSF pinch point evaluations into appropriate plantprocedures prior to implementation of the NFPA 805 fire protection program, theperformance goals for NPO modes are fulfilled and the requirements of NFPA 805 willbe met. See Implementation Items in Table S-2 of Attachment S.4.4 Radioactive Release Performance Criteria4.4.1 Overview of Evaluation ProcessThe review of the fire protection program against NFPA 805 requirements for firesuppression related radioactive release was performed using the methodologycontained in EIR 51-9085686, NMP-1 NFPA 805 Radiological Release TransitionReview. The methodology consists of the following:" Screened the fire zones based on the potential for the presence of contaminatedmaterials during all plant operating modes, including full power and non-powerconditions. The screening process considered input from radiation protectionpersonnel and review of the NMP1 fire pre-plans. The evaluation focused onradioactive release to any unrestricted area due to firefighting activities." Reviewed fire pre-plans and fire brigade training materials to identify fireprotection program elements (e.g., systems / components / procedural controlactions / flow paths, etc.) that are being credited to meet the radioactive releaseNMPI, April 2013 Page 29 IINMPI, April 2013Page 29 1 Constellation Energy Nuclear Group4.0 Compliance with NFPA 805 Requirementsgoals, objectives, and performance criteria during all plant operating modes,including full power and non-power conditions.Reviewed engineering controls to ensure containment of gaseous and liquideffluents (e.g., smoke and fire fighting agents). This review included all plantNMPI, April 2013 Page 29a IINMP1, April 2013Page 29a I REVISIONS TO TRANSITION REPORTTABLE 4-3, SUMMARY OF NFPA 805 COMPLIANCE BASIS ANDREQUIRED FIRE PROTECTION SYSTEMS AND FEATURESPages 54 through 60 with changes highlighted.

Constellation Energy Nuclear Group4.0 Compliance with NFPA 805 RequirementsConstellation Energy Nuclear Group 4.0 Compliance with NFPA 805 RequirementsTable 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and FeaturesNFPA 805 Required Required Required FireFire Suppression Detection Protection Required Fire Protection FeatureArea Fire Zone Description Regulatory System System Feature and System Details1Basis (S, E, R, D)2 (S, E, R, D)2 (S, E, R, D)21 Reactor Building East EL 198-0 thru EL 340-0 4.2.4.23FBZR237NFBZR261NFBZR281NFBZR281SFBZR298NFBZR298SFBZR318NFBZR318SFBZR340NFBZR340SReactor Building EL 237-0 COL N-Q, ROW 8-9Reactor Building EL 261-0 COL N-Q, ROW 8-9Reactor Building EL 281-0 COL M-Q, ROW 6-7Reactor Building EL 281-0 COL K-L, ROW 7-8Reactor Building EL 298-0 COL N-Q, ROW 7.5-8.5Reactor Building EL 298-0 COL K-L, ROW 7-8Reactor Building EL 318-0 COL M-Q, ROW 6-7Reactor Building EL 318-0 COL K-M, ROW 6-7Reactor Building EL 340-0 COL M-Q, ROW 6-7Reactor Building EL 340-0 COL L-N, ROW 7-8S,RS,RS,RS,RS,RS,RS,RS,RNoneNoneS,RS,RS,RS,RS,RS,RS,RS,RDDNoneNoneNoneNoneNoneNoneNoneNoneNoneNoneWater Pre-Action Sprinkler, DetectionWater Pre-Action Sprinkler, DetectionWater Pre-Action Sprinkler, DetectionWater Pre-Action Sprinkler, DetectionWater Pre-Action Sprinkler, DetectionWater Pre-Action Sprinkler, DetectionWater Pre-Action Sprinkler, DetectionWater Pre-Action Sprinkler, DetectionDetectionDetection1 R1A CTS Pump Room And General Floor Area East EL 198-0 D R None Water Pre-Action Sprinkler, Detection& 237-01 RiC Access Stairwell Southeast EL 237-0 & 261-0 None D None Detection1 R1D CS Pump Room And Protective Clothing Change Area EL D D None Water Pre-Action Sprinkler, Detection198-0 & 237-01 R2A General Floor Area East EL 261-0 D R None Water Pre-Action Sprinkler, Detection1 R3A General Floor Area East EL 281-0 None RD None Detection1 R4A General Floor Area East EL 298-0 None D None DetectionHalon Suppression System,1 R4C Emergency Condenser Isolation Valve Room EL 298-0 D D None DtconDetection1 R5A General Floor Area East EL 318-0 None D None DetectionI R6A General Floor Area East EL 340-0 None D None Detection2 Reactor Building West EL 198-0 thru EL 340-0 4.2.4.232 FBZR237N Reactor Building EL 237-0 COL N-Q, ROW 8-9 S,R S,R None Water Pre-Action Sprinkler, Detection2 FBZR261N Reactor Building EL 261-0 COL N-Q, ROW 8-9 S,R S,R None Water Pre-Action Sprinkler, Detection2 FBZR281N Reactor Building EL 281-0 COL M-Q, ROW 6-7 S,R S,R None Water Pre-Action Sprinkler, DetectionINMPI, April 2013Page 54 1 Constellation Energy Nuclear Group4.0 Compliance with NFPA 805 RequirementsTable 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and FeaturesFire NFPA 805 Required Required Required FireFire Suppression Detection Protection Required Fire Protection FeatureArea Fire Zone Description Regulatory System System Feature and System Details1Basis (S, E, R, D)2 (S, E, R, D)2 (S, E, R, D)22 FBZR281S Reactor Building EL 281-0 COL K-L, ROW 7-8 S,R S,R None Water Pre-Action Sprinkler, Detection2 FBZR298N Reactor Building EL 298-0 COL N-Q, ROW 7.5-8.5 S,R S,R None Water Pre-Action Sprinkler, Detection2 FBZR298S Reactor Building EL 298-0 COL K-L, ROW 7-8 S,R S,R None Water Pre-Action Sprinkler, Detection2 FBZR318N Reactor Building EL 318-0 COL M-Q, ROW 6-7 S,R S,R None Water Pre-Action Sprinkler, Detection2 FBZR318S Reactor Building EL 318-0 COL K-M, ROW 6-7 S,R S,R None Water Pre-Action Sprinkler, Detection2 FBZR340N Reactor Building EL 340-0 COL M-Q, ROW 6-7 None D None Detection2 FBZR340S Reactor Building EL 340-0 COL L-N, ROW 7-8 None D None Detection2 RIB CTS Pump Room, CS Pump Room, General Floor Area None R None DetectionWest EL 198-0 & 237-02 R2B General Floor Area West EL 261-0 D R None Water Pre-Action Sprinkler, Detection2 R2C Shutdown Cooling Room EL 261-0 None D None Detection2 R2D Reactor Building Track Bay EL 261-0 D None None Dry Pipe System2 R3B General Floor Area West EL 281-0 None E,R None Detection2 R4B General Floor Area West EL 298-0 None D None Detection2 R4C Emergency Condenser Isolation Valve Room EL 298-0 D D None Halon Suppression System,Detection2 R5B General Floor Area West EL 318-0 D D None Water Pre-Action Sprinkler, Detection2 R6B General Floor Area West EL 340-0 None D None Detection3 Drywell EL 237-0 thru 318-0 4.2.3.13 R1 Drywell EL 237 -318 None None None4 Foam Room EL 261-0 4.2.4.234 AB1F Foam Room EL 261-0 None RD None Detection5 Turbine Building EL 240-0 thru 369-0 4.2.4.235 FBZT261N Turbine Building Fire Break Zone North EL 261-0 E,R E,R None Water Pre-Action Sprinkler, Detection5 FBZT261S Turbine Building Fire Break Zone South EL 261-0 E,RD E,RD None Water Pre-Action Sprinkler, Detection5 OG1 General Floor Area EL 232-0 None E,D None DetectionINMPI, April 2013 Page 55 I!NMP1, April 2013Page 55 1 Constellation Energy Nuclear Group4.0 Compliance with NFPA 805 RequirementsTable 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and FeaturesNFPA 805 Required Required Required FireFire Suppression Detection Protection Required Fire Protection FeatureAire Fire Zone Description Regulatory System System Feature and System Details'Basis (S, E, R, D)f (S, E, R, D)2 (S, E, R, D)O5 OG2 General Floor Area EL 247-0 None E,D None Detection5 OG3 General Floor Area EL 261-0 E,D ER None Wet Pipe System, Detection5 T1 Turbine Condenser/Heater Bay Area EL 250-0 E,R E,R None Deluge System, CO2 required for risk,Detection5 TIA Turbine Building EL 240-261 MSIV Room & Steam Tunnel None None NoneGeneral Floor Area East of MSIV Room and Fire Zone T1 Water Pre-Action Sprinkler, Wet and5 T3A EL 261-318 E,R,D E,R None Dry Pipe System, CO required forDID, DetectionWater Pre-Action Sprinkler, CO25 T3B General Floor Area West of MSIV Room; Also South And E,R,D E,R NeeS required for DID, Detection, Promat-HWest Of Fire Zone 1 EL 237-0 & 261-0 enclosure for the HVAC ductWater Pro-Action Sprinkler, Wet Pipe5 T4A General Floor Area East Of Fire Zone T1 EL 277-0 E,R E,R None and Deluge Sytem, D etionand Deluge System, DetectionWater Pre-Action Sprinkler, Wet Pipe5 T4B General Floor Area West Of Fire Zone T1 EL 277-0 E,D E,R None Sytem DetectionSystem, Detection5 T4C Hydrogen Seal Oil Unit Room EL 277-0 E,R,D E,R None Deluge System, Co 2 required forDID, Detection5 T4D Battery Room EL 277 None E,D None Detection5 T5A General Floor Area North EL 291-0 E,D E,D None Wet Pipe System, Detection5 T6A General Floor Area North EL 305-6 ED E,R None Wet Pipe System, CO2 required forDID, Detection5 T6B Turbine Laydown Area East EL 300-0 None E,D None Detection5 T6C General Floor Area South EL 300-0 E,D E,D None Deluge System, Detection5 T6D Mechanical Storage Area EL 300-0 E,D E,D None Water Pre-Action Sprinkler, Detection5 T7A General Floor Area South EL 320-0 None E,D None DetectionT8A General Floor Area North EL 333-0, General Floor Area E,D E,D None Wet Pipe System, DetectionNorth EL 351-0, General Floor Area East EL 3695 T8B General Floor Area West EL 369-0 None E,D None DetectionT .... EL 20 42"6 T2A Turbine Building EL 250-0 R E,R None Water Pre-Action Sprinkler, DetectionNMPI, April 2013 PageSe IINMPI, April 2013Page 56 1 Constellation Energy Nuclear Group4.0 Compliance with NFPA 805 RequirementsConstellation Energy Nuclear Group 4.0 Compliance with NFPA 805 RequirementsTable 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and FeaturesNFPA 805 Required Required Required FireFire Suppression Detection Protection Required Fire Protection FeatureArea Fire Z Desption Reulatory System System Feature and System Details'Basis (S, E, R, D)2 (S, E, R, D) (S, E, R, D)27 Tu7 T2B Turbine Building South And West EL 250-0 R R None Water Pre-Action Sprinkler, Detection7 T2E UPS Battery Room EL 250 None D None Detection9 T2C Turbine Building Offgas Tunnel EL 250-0 None D None Detection9 T2D Turbine Building General Area East EL 250-0 R R None Water Pre-Action Sprinkler, Detection10~~Wae Pre-cmo Sprinkler, RoCO 5.04A.10 C1 Cable Spreading Room EL 250-0 E,R,DWater Pre-Action Sprinkler, 02required for DID, DetectionConro Cmplx L 61- AidFEL27-04.2A11C2 Auxiliary Control Room, Computer Room 261-0R,DERNoneSHalon Suppression System, C02required for DID, Detection, Promat-Henclosure for the HVAC duct I11 C3 Control Room EL 277-0 None E,R None Detection12 Amfbd uitE 5 4.42"12 AB1A Records Storage Area EL 250-0 None D None Detection12 AB1B SAS Equipment Area EL 252-0 D D None Halon Suppression System,Detection12 ABIC CPU Equipment Area EL 252-0 D D None Halon Suppression System,Detection12 ABID General Area EL 250-0 D D None Wet Pipe System, Detection12 ABlE Locker Area, Lunch Room, Offices EL 261-0 D D None Wet Pipe System, Detection12 AB2A Access Passageway EL 248-0 D None None Wet Pipe System12 AB2B Technical Support Area EL 248-0 D D None Wet Pipe System, Detection12 AB2C Radiation Records Area EL 248-0 D D None Wet Pipe System, Detection12 AB2D Warehouse Area EL 248-0 D None None Wet Pipe System12 AB3A Warehouse Area EL 261-0 D None None Wet Pipe SystemNMPI, April 2013 Page 57INMP1, April 2013Page 57 1 Constellation Energy Nuclear Group4.0 Compliance with NFPA 805 RequirementsTable 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and FeaturesNFPA 805 Required Required Required FireFire Suppression Detection Protection Required Fire Protection FeatureArea Fire Zone Description Regulatory System System Feature and System Details'Basis (S. E, R, D)2 (S, E, R, D)2 (S, E, R, D)212 AB3B Oil Storage Room EL 261-0 D None None Wet Pipe System12 AB3C Storeroom Truck Dock EL 261-0 D None None Dry Pipe System12 AB3D Electrical/Mechanical Shop Area, Office Areas, Locker OR OR None Wet Pipe System, DetectionRooms EL 261-012 AB3E Telephone Room 1 EL 261 -0 D D None Halon Suppression System,Detection12 AB3F Telephone Room 2 EL 261-0 D D None Halon Suppression System,Detection12 AB4A General Office Area EL 277-0 D D None Wet Pipe System, Detection12 AB4B File Room EL 277-0 D D None Water Pre-Action Sprinkler, Detection12 AB4C Records Processing Area EL 277-0 R R None Water Pre-Action Sprinkler, Detection12 AB4D General Office Area EL 277-0 R R None Wet Pipe System, Detection12 AB5 Penthouse Ventilation Room EL 290-0 D D None Deluge System, Detection13 S1 Screenhouse EL 225-0 -256-0 SD R None Dry Pipe System, Detection14 Disl iePm o L2104.Z4jsWet Pipe System required for14 S2 Diesel Fire Pump Room EL 256-0 SD SD None Chapter 3, Section 3.9.4,complianceD.y .-"pe -System- ,Detection15 RS1A Drum Waste Storage Vaults EL 252-0 None None None15 RS1B Electrical Equipment Room EL 252-0 D D None Halon Suppression System,Detection15 RSIC General Floor Area South, Drum Storage Room EL 252-0 None D None Detection15 RS2A Truck Loading Area, North EL 261-0 D None None Dry Pipe System15 RS2B Truck Loading Area, West EL 261-0 D None None Dry Pipe System15 RS2C General Floor Area EL 261-0 D D None Dry-pipe System, DetectionINMPI, April 2013 PageS8 IINMPI, April 2013Page 58 1 Constellation Energy Nuclear Group4.0 Compliance with NFPA 805 RequirementsTable 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and FeaturesNFPA 805 Required Required Required Fireirea Fire Zone Description Regulatory Suppression Detection Protection Required Fire Protection FeatureFArea Fr oeDsrpinRgltr System System )2 Feature and System Details'Basis (S, E, R, D)2 (S, E, R, D)2 (S. E, R, D)2Halon Suppression System,15 RS2D Radwaste Control Room, West EL 261-0 D D None DetonDetection15 RS2E General Floor Area, South EL 261-0 None None None15 RS3A General Floor Area, West EL 281-0 None D None Detection15 RS4A General Floor Area, Northwest EL 292-0 D D None Deluge System, Detection15 RS5B General Floor Area, Southwest EL 292-0 None D None Detection15 WD1 General Area EL 225-0 & 229-0 None D None Detection15 WD2 General Area EL 247-0 None D None Detection15 WD3A General Area EL 261-0 D D None Water Pre-Action Sprinkler, Detection15 WD3B Radwaste Control Room EL 261-0 None D None Detection15 WD3C Baler Room EL 261-0 D D None Dry Pipe System, Detection15 WD3D Dow Solidification Area EL 261-0 D D None Dry Pipe System, Detection15 WD3E Truck Bay EL 261-0 E,D E,D None Dry Pipe System, Detection15 WD4 Waste Building Ventilation Area EL 277-0 None D None Detection16A BIA Battery Board Room 12 EL 261-0 None E,D None Detection168~I IalyBadRo I i I EL 2iii4216B BIB Battery Board Room 11 EL 261-0 None E,D None Detection1IA Batter Ro I E I17A B2A Battery Room 12 EL 277-0 None E,D None Detection1713Ba~ey Rom II E 277ZII42Ae17B B2B Battery Room 11 EL 277-0 None E,D None Detection18 D3 EDG 102 Control Cable Missile Enclosure EL 271-0 None E,D NoneS Detection, Promat-H enclosureNMPI, April 2013 Page 59 IINMP1, April 2013Page 59 1 Constellation Energy Nuclear Group4.0 Compliance with NFPA 805 RequirementsTable 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and FeaturesNFPA 805 Required Required Required FireFire Suppression Detection Protection Required Fire Protection FeatureArea Fire Zone Description Regulatory System System Feature and System Details1Basis (S, E, R, D) (S, E, R, D)2 (S, E, R, D)219 Emergency Diesel Generator 103 Foundation Room 4.2.4.23EL 250-0 and Diesel Generator Room El 261-019 D1A EDG 103 Foundation Room EL 250-0 E,D E,D None Water Pre-Action Sprinkler, Detection19 D2A EDG 103 Room EL 261-0 R E,R None C02 System, Detection20 Diesel Generator Enclosed Cableway EL 250-0 4.2.4.2320 D1C EDG 103 Cable Routing Area EL 250-0 D D None Water Pre-Action Sprinkler, Detection21 Below Powerboards 102 & 103 EL 250-0 4.2.4.2321 D1D Room Below PB's 102 & 103 EL 250-0 D D None Water Pre-Action Sprinkler, Detection22 Emergency Diesel Generator 102 Foundation Room 4.2.4.23EL 250-0 and Diesel Generator Room EL 261-022 D1B EDG 102 Foundation Room EL 250-0 E,D E,D None Water Pre-Action Sprinkler, Detection22 D2B EDG 102 Room EL 261-0 R E,R None C02 System, Detection23 Power Board 102 Room EL 261-0 4.2.4.2323 D2C Power Board 102 Room EL 261-0 D E,R None C02 System, Detection24 Power Board 103 Room EL 261 4.2.4.2324 D2D Power Board 103 Room EL 261-0 D E,R None C02 System, DetectionEXT External to Plant 4.2.3.13EXT EXT External to Plant E E None Deluge System, DetectionNotes:1. Refer to Attachment C for each area and additional information2. NR -Not Required; S -Required for Separation; E -Required for Engineering Evaluation; R -Required for Risk; D -Required for Defense-in-Depth3. Compliance includes reliance on simplifying deterministic assumptionsNMPI, April 2013 Page 60 I!NMPI, April 2013Page 60 1 REVISIONS TO TRANSITION REPORT ATTACHMENT ANEI 04-02 TABLE B-l, TRANSITION OF FUNDAMENTALFIRE PROTECTION PROGRAM & DESIGN ELEMENTSPages A-42 through A-44 with changes highlighted.

Attachment

A -NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design ElementsConstellation Energy Nuclear Groupa.a.-I u I I itsl rt: pruL;tmUviI wdtt~rsupply system shall bededicated for fire protection useonly.%.,Um pllest WitlIClarification based onException No. IIII dUUILIUII LU lilt! PFULUL.UII UbU, 1'4 I-OL.-U 10 dIIU LII1UFSAR describe the following non-fire protection,nuclear safety / emergency uses of the fire protectionwater supply system:Exception No. 1: Fire protectionwater supply systems shall bepermitted to be used to providebackup to nuclear safetysystems, provided the fireprotection water supply systemsare designed and maintained todeliver the combined fire andnuclear safety flow demands forthe duration specified by theapplicable analysis.Exception No. 2: Fire protectionwater storage can be providedby plant systems serving otherfunctions, provided the storagehas a dedicated capacitycapable of providing themaximum fire protectiondemand for the specifiedduration as determined in thissection.a. Provide a source of make-up water to theemergency condenser make-up tanks.b. Provide an emergency source of water forcontainment and reactor vessel flooding.c. Provide an emergency source of make-up water tothe spent fuel pool.d. Provide a back-up water source for the emergencyservice water system.e. Provide a back-up water source for the dieselgenerator cooling water system.The following additional considerations apply to non-fire protection uses described above:a. Use of the electric motor-driven or diesel engine-driven fire pump as a source of emergency make-upto the emergency condenser make-up tanks wouldnot be required for a minimum of 8-hours4A hae-wgafter depletion of the emergency condenser inventory,assuming worst case conditions in which the CSTinventories are unavailable due to fire damage to theCST transfer system. If the CST transfer system isavailable, emergency condenser makeup using a firepump is not required for a minimum of 48-hours. aRdCST in'-ntorioc. Thorof-rm, In either case, concurrentfire protection use is unlikely.UFSAR Sec. X.N, Appendix 10A, Rev.21, Sec. 2.5, pg. 1OA-29UFSAR Sec. X.N, Appendix 10B, Rev.21, Appendix D, Sec. 1.c, pg. 1OB-220Calculation $13.1-100F005, "DieselFire Pump / Reactor Vessel Flooding,"Rev. 0, AllCalculation S13.1-100F006," PressureDrop Calculation, NMP2 Main FirePumps Supply to NMP1 Fire WaterDistribution System," Rev. 0, AllCalculation 13.1-100F007, "HydraulicAnalysis of Diesel Fire Pump Supplyto ESW #11 and Emergency DieselCooling Water Systems," Rev. 0, AllNMPI, April 2013 Page A-42 IINMPI, April 2013Page A-42 I Constellation Energy Nuclear GroupAttachment A -NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements,bhaptei. renceand Compliance Statement Compliance Bases Reference DocumentsReqgu irementslGUidance3.5.16 (cont.) (cont.)b. Use of the electric motor-driven or diesel engine-driven fire pump for emergency containment andreactor vessel flooding would only be required in theunlikely event that all other means of core injectionare lost. On this basis, concurrent fire protection useis unlikely.c. Use of the electric motor-driven or diesel engine-driven fire pump as a source of emergency spent fuelpool inventory would only be required upon loss of theFuel Pool Make-up System, condensate transfer viahoses, and demineralized water with hoses fromrefueling service connections. On this basis,concurrent fire protection use is unlikely.d. and e. Use of the diesel engine-driven fire pump asa source of ESW and EDG cooling water would onlybe required upon occurrence of a fire in theScreenhouse that could disable all otherScreenhouse pumps.Clarification: The required flow rate and pressuredemand for each of the non-fire protection, nuclearsafety / emergency uses of the fire protection watersupply system described above can be supplied byuse of one (1) of the redundant fire pumps. Therefore,in the unlikely event that concurrent fire protectionand non-fire protection use is required, either theremaining Unit 1 fire pump, or the Unit 2 fireprotection water supply system will be available. Toensure adequate water supply for fire suppressionactivities concurrent with other uses, a modificationfor a cross-connection will be installed. Seemodifications in Attachment S.NMI pi 01 aeA4INIVIPI, April 2013Page A-43 I

Attachment

A -NEI 04-02 Table 13-1 Transition of Fundamental Fire Protection Program & Design ElementsConstellation Energy Nuclear Group3.5.16 (cont.) (cont.)Additionally, since the source of the Unit I and Unit 2fire protection water supply is Lake Ontario, the wateravailable to supply fire protection and/or non-fireprotection demands is not a concern.3.6 Standpipe and Hose N/A N/A -Section title, no technical requirements. SeeStations. sub-sections for specific compliance statementsand references.3.6.1 For all power block Complies with use of Standpipe and hose systems are provided for all NI-SD-018, Rev. 05, Sec. 2.2.5buildings, Class III standpipe EEEE power block structures. Per N1-SD-018, standpipeand hose systems shall be risers are located at various points throughout Nine UFSAR Sec. X.N, Appendix 10A, Rev.installed in accordance with Mile Point Unit I to serve hose stations. The 21, Sec. 2.5.3.4, Table 1.2.2,NFPA 14, Standard for the standpipes/hose stations are so spaced as to permit pgs. I OA-46, 1OA-47, 1OA-86Installation of Standpipe, hose stream coverage of all points in the buildingsPrivate Hydrant, and Hose including primary containment. Hose stations are FPEE 0-98-003, "Acceptable Use ofSystems. equipped with 100 feet of 1-1/2 inch hose with Aluminum Fire Hose Couplings," Rev.adjustable spray nozzles. Hose stream coverage is in 0accordance with NFPA 14 Class III systems.EIR 51-9077284-000, "NMP-1 CodeReviews," Sec. 4.0, Appendix F3.6.2 A capability shall be Complies Based on the fire protection water supply design $13.1-100F002, -Fire Protectionprovided to ensure an adequate information contained in $13.1-100F002, the NMP1 Water Supply," Rev. 02water flow rate and nozzle fire protection water supply system is capable ofpressure for all hose stations. providing adequate flow and pressure for all hose UFSAR Sec. X.N, Appendix 1OA, Rev.This capability includes the stations and exceeds NFPA 14-1963 design 21, Sec. 2.5.3.4, Table 1.2.2,provision of hose station requirements. pgs. 1OA-46, 1OA-47, 1OA-86pressure reducers wherenecessary for the safety of plant Pressure reduction devices are not installed for hoseindustrial fire brigade members stations. This has been deemed acceptable becauseand off-site fire department fire hoses connected to the standpipe system arepersonnel. intended for use exclusively by trained fire brigadeI personnel. INMPI, April 2013 Page A-44 IINMP1, April 2013Page A-" I REVISIONS TO TRANSITION REPORT ATTACHMENT BNEI 04-02 TABLE B-2, NUCLEAR SAFETY CAPABILITYASSESSMENT -METHODOLOGY REVIEWPages B-1 through B-102 with changes highlighted.

Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology ReviewB. NEI 04-02 Table B-2 -Nuclear Safety Capability Assessment -Methodology Review101 Pages AttachedNMPI, April 2013 Page B-IINMP1, April 2013Page B-1 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionA comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire event shall be developed. The equipment list shall contain an inventory of thosecritical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions andcomponents whose fire-induced failure could prevent the operation or result in the mal-operation of those components needed to meet the nuclear safety criteria shall be included.Availability and reliability of equipment selected shall be evaluated.NEI 00-01 Ref3 Deterministic MethodologyNEI 00-01 GuidanceThis section discusses a generic deterministic methodology and criteria that licensees can use to perform a post-fire safe shutdown analysis toaddress regulatory requirements. For a complete understanding of the deterministic requirements, work this section in combination with theinformation in Appendix C, High/Low Pressure Interfaces, Appendix D, Alternative and Dedicated Shutdown Requirements, Appendix E,Acceptance Criteria for Operator Manual Actions and repairs, and Appendix H, Hot Shutdown versus Important to Safe ShutdownComponents. To resolve the industry issue related to MSOs, refer to Section 4, Appendix B, Appendix F and Appendix G. The plant specificanalysis approved by NRC is reflected in the plant's licensing basis. The methodology described in this section is an acceptable method ofperforming a post-fire safe shutdown analysis. This methodology is depicted in Figure 3-1. Other methods acceptable to NRC may also beused. Regardless of the method selected by an individual licensee, the criteria and assumptions provided in this guidance document mayapply. The methodology described in Section 3 is based on a computer database oriented approach, which is utilized by several licensees tomodel Appendix R data relationships. This guidance document, however, does not require the use of a computer database orientedapproach.ApplicabilityApplicableCommentsNoneAlignment StatementAlignsAlignment BasisThis document addresses the comparison of the deterministic methodology used for the existing the Nine Mile Point 1 (NMP1) Safe Shutdown Analysis and the requirements of10CFR50 Appendix R, Sections IlI.G.1, III.G.2 and Ill.G.3 against the compliance requirements and criteria specified in NFPA 805. The subsequent sections determine the extent theanalysis meets the requirements as described in NEI 04-02. This documents a line by line review and comparison against the methodology and criteria provided in Chapter 3 of NEI00-01, Revision 2. The deterministic methodology described in Section 3 and Figure 3-1 of NEI 00-01 was utilized as documented in the following Table B-2 sections. The NMP1 safeshutdown methodology utilizes a computer oriented database to model data relationships for systems, components, and cables used to comply with the requirements for post fire safeshutdown. This review of modifications, procedural controls, repair procedures and previously approved configurations and boundaries demonstrates that the safe shutdown analysisgenerally meets the Nuclear Safety Performance Criteria including the information provided in Appendix B of NFPA 805 related to circuit criteria and Multiple Spurious Operations(MSOs).Reference DocumentEIR 51-9133191, NSCA, Section 9.0NMPI, April 2013 Page B-2 IINMP1, April 2013Page B-2 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.1 Safe Shutdown Systems andPath DevelopmentNEI 00-01 GuidanceThis section discusses the identification of systems necessary to perform the required safe shutdown functions. It also providesinformation on the process for combining these systems into safe shutdown paths. Appendix R Section III.G.1 .a requires that thecapability to achieve and maintain hot shutdown be free of fire damage. Appendix R Section Ill.G.1 .b requires that repairs to systems andequipment necessary to achieve and maintain cold shutdown be completed within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . This section provides some guidance onclassifying components as either required or important to SSD circuit components. It also provides some guidance on the tools availablefor mitigating the effects of fire-induced circuit failures to each of these classes of equipment. For a more detailed discussion of the topicof required and important to SSD components refer to Appendix H.The goal of post-fire safe shutdown is to assure that a one train of shutdown systems, structures, and components remains free of firedamage for a single fire in any single plant fire area. This goal is accomplished by determining those functions required to achieve andmaintain hot shutdown. Safe shutdown systems are selected so that the capability to perform these required functions is a part of eachsafe shutdown path. The functions required for post-fire safe shutdown generally include, but are not limited to the following:* Reactivity control* Pressure control systemsInventory control systemsDecay heat removal systemsProcess monitoring (as defined in NRC Information Notice 84-09)Support systems* Electrical power and control systems* Component Cooling systems* Component Lubrication systemsThese functions are of importance because they have a direct bearing on the safe shutdown goal of being able to achieve and maintainhot shutdown, which ensures the integrity of the fuel, the reactor pressure vessel and the primary containment. If these functions arepreserved, then the plant will be safe because the fuel, the reactor and the primary containment will not be damaged. By assuring thatthis equipment is not damaged and remains functional, the protection of the health and safety of the public is assured.The components required to perform these functions are classified as required for hot shutdown components. These components arenecessary and sufficient to perform the required safe shutdown functions assuming that fire-induced impacts to other plantequipment/cables do not occur. Since fire-induced impacts to other plant equipment/cables can occur in the fire condition, these impactsmust also be addressed. The components not necessary to complete the required safe shutdown functions, but which could be impactedby the fire and cause a subsequent impact to the required safe shutdown components are classified as either required for hot shutdown orimportant to SSD components. Depending on the classification of the components, the tools available for mitigating the effects of fire-induced damage vary. The available tools are generally discussed in this section and in detail in Appendix H. The classification of acomponent or its power or control circuits may vary from fire area to fire area. Therefore, the required safe shutdown path for any givenfire area is comprised of required for hot shutdown components and important to SSD components. The distinction and classification foreach required safe shutdown path for each fire area should be discernible in the post-fire safe shutdown analysis.Generic Letter 81-12 specifies consideration of associated circuits of concern with the potential for spurious equipment operation and/orloss of power source, and the common enclosure failures. As described above, spurious operations/actuations can affect theaccomplishment of the required safe shutdown functions listed above. Typical examples of the effects of the spurious operations ofconcern are the following:NMPI, April 2013 Page B-3 I!NMPI, April 2013Page B-3 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review* A loss of reactor pressure vessel/reactor coolant inventory in excess of the safe shutdown makeup capability.* A flow loss or blockage in the inventory makeup or decay heat removal systems being used for the required safe shutdownpath.Spurious operations are of concern because they have the potential to directly affect the ability to achieve and maintain hot shutdown,which could affect the fuel and cause damage to the reactor pressure vessel or the primary containment. To address the issue of multiplespurious operations, Section 4 provides a Resolution Methodology for developing a Plant Specific List of MSOs for evaluation. AppendixB provides the circuit failure criteria applicable to the evaluation of the Plant Specific list of MSOs.Common power source and common enclosure concerns could also affect the safe shutdown path and must be addressed.In addition to the tools described for components classified as required for hot shutdown, fire-induced impacts to cables and componentsclassified as important to SSD may be mitigated using some additional tools. For important to SSD component failures, operator manualactions, fire modeling and/or a focused-scope fire PRA may be used to mitigate the impact. (If the use of a Focused-Scope Fire PRAs isnot permitted in the Plants Current License Basis, then, a License Amendment Request (LAR) will be necessary to use the Focused-Scope Fire PRA).Applicability CommentsApplicable NoneAlignment StatementAlignsAlignment BasisTo achieve post-fire safe shutdown, the Safe Shutdown Systems, their functions, and components required to support the safe shutdown functions were identified. P&IDs andElectrical drawings were marked up and annotated to select equipment and specific flow paths for each system required to support safe shutdown. This information was populated intoa computer database to provide a database oriented approach to model Appendix R data relationships.Safe shutdown systems, components, and cables were selected using the criteria and assumptions of NEI 00-01, Rev. 2, Sections 3.1.1 and 3.1.2 to satisfy each safe shutdownfunction performance goals, including process monitoring and support systems for each path. NEI 00-01 was used as the guidance in developing the Safe Shutdown Equipment Listand the Safe Shutdown success paths. Safe Shutdown paths were designed based on the combination of systems in the respective fire area.The ability to achieve post-fire Safe Shutdown (SSD) is assured by having at least one safe shutdown path of the required systems, structures, and components to remain free of firedamage. This assurance that the safe shutdown equipment is available supports the required performance goals identified in the guidance, maintains the integrity of the fuel, reactorpressure vessel and primary containment. The SSD path used to achieve post-fire safe shutdown is comprised of SSD systems and components that remain free of fire damage.Reference DocumentEIR 51-9133191, NSCA, Sections 4.0, 5.0, 5.1, 5.2, 8.6.2, and 8.6.3HNP RAI 3-6, RAI 3-9, RAI 3-10, and RAI 3-11, NRC Requests for Additional Information dated August 6, 2009 (ML092170715)Closure of National Fire Protection Association 805 Transition Program Frequently Asked Question Number 06-0006 (ML070030117)ONS RAI 3-35, NRC Request for Additional Information dated November 18, 2009 (ML092920347)NMPI, April 2013 Page B-4 IINMP1, April 2013Page B-4 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.1.1 Criteria/AssumptionsNEI 00-01 GuidanceThe following criteria and assumptions should be considered, as applicable, when identifying systems available and necessary to performthe required safe shutdown functions and combining these systems into safe shutdown paths. This list provides recognized examples ofcriteria/assumptions but should not be considered an all-inclusive list. The final set of criteria/assumptions should be based on regulatoryrequirements and the performance criteria for post-fire safe shutdown for each plant.ApplicabilityApplicableAlignment StatementNot RequiredCommentsNoneAlignment BasisThis paragraph provides introductory information and contains no specific requirements. Discussion is provided in subsequent sub-sections.Reference DocumentNMPI, April 2013 Page B-5 IINMP1, April 2013Page B-5 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.1.1.1 Safe Shutdown Paths ForBWRsNEI 00-01 Guidance[BWR] GE Report GE-NE-T43-00002-00-01-RO1 entitled "Original Safe Shutdown Paths For The BWR" addresses the systems andequipment originally designed into the GE boiling water reactors (BWRs) in the 1960s and 1970s, that can be used to achieve andmaintain safe shutdown per Section III.G.1 of 10CFR 50, Appendix R. Any of the shutdown paths (methods) described in this report areconsidered to be acceptable methods for achieving redundant safe shutdown.ApplicabilityApplicableCommentsNoneAlignment StatementAlignsAlignment BasisThe primary means of achieving and maintaining hot shutdown following a fire coincident with a Loss of Offsite Power (LOOP) is via the Emergency Condenser (EC) system. The ECsystem consists of two redundant emergency cooling loops; each loop is capable of independently accomplishing hot shutdown. Therefore, this option provides two redundant pathsfor obtaining hot shutdown.The EC system operates by natural circulation. Steam flows from the vessel through the EC tubes. Condensate returns to the vessel through a reactor recirculation loop. Boiling ofwater in the secondary side of the ECs, which is vented to the atmosphere, provides the necessary cooling.In the event the preferred shutdown method is not available, the plant can be shut down by opening three Electromatic Relief Valves (ERVs) and discharging steam to the Torus toreduce pressure. When reactor pressure reaches approximately 365psig, Core Spray (CS) may be initiated. CS is a two loop system. Operation of one LOOP is adequate to achieveshutdown. Eventually, the reactor vessel floods to the point where the ERVs are passing fluid to the Torus rather than steam, in essence placing the Reactor Coolant System inrecirculation through the Torus. During this process, decay heat is removed by operation of the Containment Spray System in conjunction with the Containment Spray Raw Watersystem. This shutdown method will bring the plant directly to CSD. Fully flooding the Reactor Pressure Vessel negates the need for another system to provide inventory makeup. ACpower is required to initiate this shutdown method.Reference DocumentEIR 51-9133191, NSCA, Sections 5.1 and 5.2NMPI, April 2013 Page B-6INMP1, April 2013Page B-6 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.1.1.2 SRVs and LPCI/CSNEI 00-01 Guidance[BWR] GE Report GE-NE-T43-00002-00-03-RO1 provides a discussion on the BWR Owners' Group (BWROG) position regarding the useof Safety Relief Valves (SRVs) and low pressure systems (LPCI/CS) for safe shutdown. The BWROG position is that the use of SRVsand low pressure systems is an acceptable methodology for achieving redundant safe shutdown in accordance with the requirements of10 CFR 50 Appendix R Sections III.G.1 and III.G.2. The NRC has accepted the BWROG position and issued an SER dated Dec. 12,2000.ApplicabilityApplicableAlignment StatementAlignsCommentsNoneAlignment BasisFor NMP1, using the Electromatic Relief Valves (ERVs) to remove decay heat to the torus and low pressure Core Spray system is an acceptable method for achieving safe shutdown.The Emergency Condensers provide one shutdown flow path while the ERVs provide an Alternate Shutdown method.Reference DocumentEIR 51-9133191, NSCA, Section 4.0NMP1 Safety Evaluation 84-18, ADS Logic ModificationNMPI, April 2013 Page B-7 IINMP1, April 2013Page B-7 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.1.1.3 Pressurizer HeatersNEI 00-01 Guidance[PWR] Generic Letter 86-10, Enclosure 2, Section 5.3.5 specifies that hot shutdown can be maintained without the use of pressurizerheaters (i.e., pressure control is provided by controlling the makeup/charging pumps). Hot shutdown conditions can be maintained vianatural circulation of the RCS through the steam generators. The cooldown rate must be controlled to prevent the formation of a bubble inthe reactor head. Therefore, feedwater (either auxiliary or emergency) flow rates as well as steam release must be controlled.ApplicabilityNot ApplicableAlignment StatementNot RequiredAlignment BasisNMP1 is a BWR plant.Reference DocumentCommentsNMP1 is a BWR plant.NMPI, April 2013 Page B-8 IINMP1, April 2013Page B-8 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.1.1.4 Alternative ShutdownClassificationNEI 00-01 GuidanceThe classification of shutdown capability as alternative/dedicated shutdown is made independent of the selection of systems used forshutdown. Alternative/dedicated shutdown capability is determined based on an inability to assure the availability of a redundant safeshutdown path. Compliance to the separation requirements of Sections II.G.1 and III.G.2 may be supplemented by the use of operatormanual actions to the extent allowed by the regulations and the licensing basis of the plant (see Appendix E), repairs (cold shutdownonly), exemptions, deviations, GL 86-10 fire hazards analyses or fire protection design change evaluations permitted by GL 86-10, asappropriate. These may also be used in conjunction with alternative/dedicated shutdown capability. A discussion of time zero for the firecondition, as it relates to operator manual actions and repairs, is contained in Appendix E.ApplicabilityApplicableCommentsNoneAlignment StatementAlignsAlignment BasisThe criteria of 10CFR50 Appendix R, Section III.G.3, requires alternate or dedicated shutdown capability for all plant areas where the protection of systems for hot shutdown does notsatisfy the requirements of Sections Ill.G.1 and IlI.G.2 of Appendix R. The hot shutdown system to be used for a control room evacuation event is the EC system.NMP1 has two Remote Shutdown Panels installed for the purpose of monitoring the Plant Shutdown in the event of a Control Room evacuation. Modifications were performed whichhardened the EC system from spurious isolations due to the effects of a control complex fire. Upon receiving either a high reactor pressure signal or a low-low reactor water levelsignal, the ECs will automatically initiate, independent of the control complex, due to the shutdown supervisory control system redundant initiation logic located in the reactor building,although Operator action will initiate the safe shutdown systems prior to its automatic initiation to conserve reactor vessel inventory.Reference DocumentEIR 51-9133191, NSCA, Section 2.1.1NMP1 Safety Evaluation 83-29, Emergency CondensersNMPI, April 2013 Page B-9 IINMPI, April 2013Page B-9 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.1.1.5 Operable and AvailableNEI 00-01 GuidanceAt the onset of the postulated fire, all safe shutdown systems (including applicable redundant trains) are assumed operable and availablefor post-fire safe shutdown. Systems are assumed to be operational with no repairs, maintenance, testing, Limiting Conditions forOperation, etc. in progress. The units are assumed to be operating at full power under normal conditions and normal lineups.ApplicabilityApplicableCommentsNoneAlignment StatementAlignsAlignment BasisAs stated in the Fire Area Assessments at the onset of the fire, all systems not affected by the fire are considered to be available and capable of functioning as designed. All safeshutdown systems (including redundant trains) are assumed to be operational and available for post-fire safe shutdown. Systems are assumed to be operational with no repairs,maintenance, testing, Limiting Conditions for Operation, etc. The unit is assumed to be operating at full power under normal conditions and normal lineups.Reference DocumentEIR 51-9133191, NSCA, Section 4.0NMPI, April 2013 Page B-b IINMP1, April 2013Page B-10 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref NEI 00-01 Guidance3.1.1.6 No concurrent DBAs No Final Safety Analysis Report accidents or other design basis events (e.g. loss of coolant accident, earthquake), single failures ornonfire induced transients need be considered in conjunction with the fire.Applicability CommentsApplicable NoneAlignment StatementAlignsAlignment BasisNo accidents or other design basis events (e.g. loss of coolant accident, earthquake), single failures or non-fire induced transients are considered in conjunction with the fire.The fire does not occur simultaneously or coincident with any other transient or abnormal condition, except for loss of offsite power (LOOP) and those conditions resulting directly fromthe effects of a fire. No credit is taken for offsite power availability; however, offsite power is considered to be available if the fire effects produce more conservative results.Reference DocumentEIR 51-9133191, NSCA, Section 9.0NMPI, April 2013 Page B-lI IINMP1, April 2013Page B-11 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.1.1.7 Offsite Power AvailabilityNEI 00-01 GuidanceFor the case of redundant shutdown, offsite power may be credited if demonstrated to be free of fire damage. Offsite power should beassumed to remain available for those cases where its availability may adversely impact safety (i.e., reliance cannot be placed on firecausing a loss of offsite power if the consequences of offsite power availability are more severe than its presumed loss). No credit shouldbe taken for a fire causing a loss of offsite power. For areas where train separation cannot be achieved and altemative shutdowncapability is necessary, shutdown must be demonstrated both where offsite power is available and where offsite power is not available for72 hours.ApplicabilityApplicableCommentsONS RAI 3-40 is addressed; No credit is taken for offsite power.Alignment StatementAlignsAlignment BasisFor redundant safe shutdown, offsite power is presumed lost. However, offsiteOffsite power is assumed to remain available for those cases where its availability may adversely impactsafety (i.e., reliance cannot be placed on fire causing a loss of offsite power if the consequences of offsite power availability are more severe than its presumed loss). No credit istaken for a fire causing a loss of offsite power. For areas where train separation cannot be achieved and alternative shutdown capability is necessary, shutdown is demonstrated bothwhere offsite power is available and where offsite power is not available (NFPA 805 requires maintaining the fuel in a safe and stable condition, i.e., there is no requirement to achieveand maintain CSD, and therefore no 72-hour coping time requirement). -f9 72 hheur.-Offsite power has not been specifically analyzed. There are no fire areas where offsite power iscredited.Reference DocumentEIR 51-9133191, NSCA, Section 9.0ONS RAI 3-40, NRC Request for Additional Information dated July 30, 2010 (ML102110394)Closure of National Fire Protection Association 805 Transition Program Frequently Asked Question Number 08-0054 (ML110140183)NMPI, April 2013 Page 6-12 IINMP1, April 2013Page B-12 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref NEI 00-01 Guidance3.1.1.8 Safety Related Classification Post-fire safe shutdown systems and components are not required to be safety-related.Applicability CommentsApplicable NoneAlignment StatementAlignsAlignment BasisCredited safe shutdown systems and components are not always safety related. Most components are safety related due to their emergency functions, but they are not required to besafety-related.Reference DocumentEIR 51-9133191, NSCA, Section 4.0NMPI, April 2013 Page B-13 IINMP1, April 2013Page B-13 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.1.1.9 72-hour Coping PeriodNEI 00-01 GuidanceThe post-fire safe shutdown analysis assumes a 72-hour coping period starting with a reactor scram/trip. Fire-induced impacts thatprovide no adverse consequences to hot shutdown within this 72-hour period need not be included in the post-fire safe shutdownanalysis. At least one train can be repaired or made operable within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months using onsite capability to achieve cold shutdown.ApplicabilityApplicableCommentsNoneAlignment StatementAlignsAlignment BasisAppendix R requires cold shutdown of the reactor within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for fire events, with or without offsite power available. The original NMP1 The-Safe Shutdown Analysis identifies thesafe shutdown systems and components which are powered by on-site sources, where at least one train can be repaired or made operable within the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . However, NFPA 805does not require a 72-hour coping period. Rather, the requirement is to maintain the fuel in a safe and stable condition (i.e., there is neither a requirement nor timeframe to reach andmaintain cold shutdown). Refer to Section 4.2.1.2 for a description of safe and stable as applied to NMP1. NMP1 demonstrates the ability to maintain for This is ac..mplished for oachfire area the fuel in a safe and stable condition with one of four designated shutdown paths."mld Id- tains. Offsite power is not credited with providing any power or beneficialeffects.ahutdaum m.thods duri~n the 72 hMUMs. thereb mot--no,. thi ream'ro.--_.Shutdown Paths:Cold Shu8tdo-wn, Options:A.- Train 11 -Emergency cooling, Shutdo, n "coling, RBCLC, ESW, CRD system (makeup).B. Train 12 -Emergency cooling, Shutdown -coling, RSCLC, ESW, CRD system (makeup).C. Train 11 -ERVs, core spray, containment spray, containment spray raw water.D. Train 12 -ERVs, core spray, containment spray, containment spray raw water.Note: As part of the NMP1 defense-in-depth approach to fire protection, provisions have been made for permanent installation of a feedwater/fire protection water spool piece, whichwill provide an emergency makeup source from the diesel fire pump for cold shutdown inventory control.S-hutdownim Coo-ling with LOOP (Options A and 2)The primary means for aahieving and maintaining Gold shutdown following a fire coeinc-ld-ent ith a LOGOP is via the shutdown cooling system. The shutdown cooling system, supportedby the RBCLC and the SSW systems, rcmovoc decay heat from the reactor. vesscolet the UHIS. Em~ergency AG power (Oncite DOs) is required for this mode of Geld shutdown.Reference DocumentEIR 51-9133191, NSCA, Section 4.0 and 5.0HNP RAI 3-7 and RAI 3.8, NRC Request for Additional Information dated August 6, 2009 (ML092170715)Closure of National Fire Protection Association 805 Transition Program Frequently Asked Question Number 08-0054 (ML1 10140183)lNMP1, April 2013Page B-14 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 RefNEI 00-01 Guidance3.1.1.10 Manual Initiation of Systems Manual initiation from the main control room or emergency control stations of systems required to achieve and maintain safe shutdown isacceptable where permitted by current regulations or approved by NRC (See Appendix E); automatic initiation of systems selected forsafe shutdown is not required but may be included as an option, if the additional cables and equipment are also included in the analysis.Spurious actuation of automatic systems (Safety Injection, Auxiliary Feedwater, High Pressure Coolant Injection, Reactor Core IsolationCooling, etc.) due to fire damage, however, should be evaluated.ApplicabilityApplicableCommentsNoneAlignment StatementAlignsAlignment BasisManual initiation of safe shutdown related equipment from either the control room, emergency control stations or approved locations other than the primary control stations is anacceptable means for compliance based on the current regulations. The safe shutdown performance and design requirements for the reactivity control function can be met withoutautomatic scram capability. The post-fire safe shutdown analysis need only provide the capability to manually scram the reactor. Automatic functions of components have beenincluded for selected systems. Impacts due to spurious actuation of automatic systems are included in the evaluation of the analysis.Reference DocumentEIR 51-9133191, NSCA, Sections 4.0, 5.1, 5.2, and 5.4NMP1, April 2013Page B-15 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.1.1.11 Multi-unit PlantNEI 00-01 GuidanceWhere a single fire can impact more than one unit of a multi-unit plant, the ability to achieve and maintain safe shutdown for each affectedunit must be demonstrated.ApplicabilityApplicableCommentsNoneAlignment StatementAlignsAlignment BasisAs stated in the UFSAR, simultaneous fires affecting NMP1 and NMP2 are not anticipated, due to the spatial separation between the two units and the designed separation (firebarrier) at facility interfaces. NMP1 and NMP2 do not share common facilities for the support of reactor operation or generation of electricity. However, there is the capability to cross-tie the firewater system between NMP1 and NMP2 via manual ross-tie valves.Reference DocumentUFSAR, Appendix 10A, Section 2.1.9NM P1, April 2013 Page B-16 I!NMP1, April 2013Page B-16 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref NEI 00-01 Guidance3.1.2 Shutdown Functions The following discussion on each of these shutdown functions provides guidance for selecting the systems and equipment required for hotshutdown. For additional information on BWR system selection, refer to GE Report GENE-T43-00002-00-01-ROl entitled "Original SafeShutdown Paths for the BWR."Applicability CommentsApplicable NoneAlignment StatementNot RequiredAlignment BasisThis paragraph provides introductory information and contains no specific requirements. Discussion is provided in subsequent sub-sections.Reference DocumentNMPI, April 2013 Page B-17 IINMP1, April 2013Page B-17 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.1.2.1 Reactivity ControlNEI 00-01 Guidance[BWR] Control Rod Drive SystemThe safe shutdown performance and design requirements for the reactivity control function can be met without automatic scram/tripcapability. Manual scram/reactor trip is credited. The post-fire safe shutdown analysis must only provide the capability to manuallyscram/trip the reactor. Each licensee should have an operator manual action to either vent the instrument air header or to remove RPSpower in their post-fire safe shutdown procedures. The presence of this action precludes the need to perform circuit analysis for thereactivity control function and is an acceptable way to accomplish this function. If this action is a "time critical" action, the timing must bejustified.[PWR] Makeup/ChargingThere must be a method for ensuring that adequate shutdown margin is maintained from initial reactor SCRAM to cold shutdownconditions, by controlling Reactor Coolant System temperature and ensuring borated water is utilized for RCS makeup/charging.ApplicabilityApplicableCommentsNoneAlignment StatementAligns With IntentAlignment BasisAs documented in the current NMP1 Safe Shutdown Analysis, the reactivity control function is capable of achieving and maintaining cold shutdown reactivity conditions. The safeshutdown performance and design requirements for the reactivity control function can be met without automatic scram capability. The post-fire safe shutdown analysis provides thecapability to manually scram the reactor. Manual reactor scram is accomplished via the scram valves. Plant operators can manually scram the plant from the control room or theremote shutdown panel. The capability to manually scram/trip the reactor is provided in NMP1 Special Operating Procedures N1-SOP-21.1, Fire in Plant, and N1-SOP-21.2, ControlRoom Evacuation, with reference to the Emergency Operating Procedures for operators to manually vent the instrument air header or to remove RPS power. This action is notconsidered a "time critical" action because the Mode switch is placed in shutdown and all control rods inserted prior to evacuating the control room. Also, NMP1 complies with theposition in BWROG document BWROG-TP-1 1-011 entitled "BWROG Assessment of Generic Multiple Spurious Operations (MSOs) in Post-Fire Safe Shutdown Circuit Analysis for theOperation of BWR Plants" for manual scram; thereby, supporting that this is not a time critical action.Reference DocumentEIR 51-9133191, NSCA, Sections 4.0 and 5.3NMP1 Emergency Operating Procedure N1-EOP-2, RPV ControlNMP1 Emergency Operating Procedure 3, Failure to ScramNMP1 Emergency Operating Procedure 3.1, Alternate Control Rod InsertionNMP1 Special Operating Procedure, N1-SOP-21.2, Control Room Evacuation, pg. 3NMP1 Special Operating Procedure, N1-SOP-21.1, Fire in Plant, pg. 2NMPI, April 2013 Page B-18 IINMP1, April 2013Page B-18 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref NEI 00-01 Guidance3.1.2.2 Pressure Control Systems The systems discussed in this section are examples of systems that can be used for pressure control. This does not restrict the use ofother systems for this purpose.[BWR] Safety Relief Valves (SRVs)Initial pressure control may be provided by the SRVs mechanically cycling at their setpoints (electrically cycling for EMRVs).Mechanically-actuated SRVs require no electrical analysis to perform their overpressure protection function. The SRVs may also beopened to maintain hot shutdown conditions or to depressurize the vessel to allow injection using low pressure systems. These areoperated manually. Automatic initiation of the Automatic Depressurization System (ADS) is not a required function. Automatic initiation ofthe ADS may be credited, if available. If automatic ADS is not available and use of ADS is desired, an alternative means of initiation ofADS separate from the automatic initiation logic for accomplishing the pressure control function should be provided.[PWR] Makeup/ChargingRCS pressure is controlled by controlling the rate of charging/makeup to the RCS. Although utilization of the pressurizer heaters and/orauxiliary spray reduces operator burden, neither component is required to provide adequate pressure control. Pressure reductions aremade by allowing the RCS to cool/shrink, thus reducing pressurizer level/pressure. Pressure increases are made by initiatingcharging/makeup to maintain pressurizer level/pressure. Manual control of the related pumps is acceptable.Applicability CommentsApplicable NoneAlignment StatementAlignsAlignment BasisThe pressure control function is capable of safely reducing reactor vessel pressure.For the preferred shutdown method, pressure control is achieved through control of the cooldown rate of the emergency condensers. The Electromatic Relief Valves (ERVs) aremaintained closed.In the event the preferred shutdown method is not available, the plant can be shut down by opening three ERVs and discharging steam to the Torus to reduce pressure. For thissecondary method, the ERVs are opened manually to depressurize the vessel to allow injection using low pressure systems. Automatic initiation of the ADS is not a required function.Reference DocumentEIR 51-9133191, NSCA, Sections 4.0, 5.1, 5.2, and 5.3NMPI, April 2013 Page 6-19 IINMP1, April 2013Page B-19 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.1.2.3 Inventory ControlNEI 00-01 Guidance[BWR] Systems selected for the inventory control function should be capable of supplying sufficient reactor coolant to achieve andmaintain hot shutdown. Manual initiation of these systems is acceptable. Automatic initiation functions are not required. Spuriousactuation of automatic systems, however, should be evaluated (High Pressure Coolant Injection, High Pressure Core Spray, Reactor CoreIsolation Cooling, etc.).[PWR]: Systems selected for the inventory control function should be capable of maintaining level to achieve and maintain hot shutdown.Typically, the same components providing inventory control are capable of providing pressure control. Manual initiation of these systemsis acceptable. Automatic initiation functions are not required. Spurious actuation of automatic systems, however, should be evaluated(Safety Injection, High Pressure Injection, Auxiliary Feedwater, Emergency Feedwater, etc.).ApplicabilityApplicableAlignment StatementAlignsAlignment BasisCommentsNoneFor the preferred hot shutdown method, reactor vessel make-up is required 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after operation of the Emergency Condensers (EC) ensues. A Control Rod Drive pump is used toprovide vessel make-up. The inventory makeup provided by the CRD pump is also used part of Gold shuJtdo, to raise the Reactor water level at such time that and-make-theshutdown cooling system is placed in offec"Ps,. Restoration of the CRD pump is vi. r.pair a.... to ,ecur po.er SOUrcs, SIc.NMP1 has a defense-in-depth inventory control approach for fire protection by using the firewater to feedwater connection in accordance with NMP1 operating procedures, whichprovides an emergency makeup source from the diesel fire pump for Gold-shutdown inventory control.As a secondary method, the vessel can be flooded by the Core Spray system after depressurization. This fulfills the vessel inventory make-up by default.Spurious Actuations are addressed specifically in the Fire Area Assessments. They are considered to exist from the onset of the fire and for the duration of the shutdown process.Reference DocumentEIR 51-9133191, NSCA, Sections 5.3 and 8.6NMP1, April 2013Page B-20 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref NEI 00-01 Guidance3.1.2.4 Decay Heat Removal [BWR] Systems selected for the decay heat removal function(s) should be capable of:-Removing sufficient decay heat from primary containment, to prevent containment over-pressurization and failure.* Satisfying the net positive suction head requirements of any safe shutdown systems taking suction from the containment(suppression pool).* Removing sufficient decay heat from the reactor to achieve cold shutdown. (This is not a hot shutdown requirement).[PWR] Systems selected for the decay heat removal function(s) should be capable of:* Removing sufficient decay heat from the reactor to reach hot shutdown conditions. Typically, this entails utilizing naturalcirculation in lieu of forced circulation via the reactor coolant pumps and controlling steam release via the Atmospheric Dumpvalves.* Removing sufficient decay heat from the reactor to reach cold shutdown conditions. (This is not a hot shutdown requirement).This does not restrict the use of other systems.Applicability CommentsApplicable NoneAlignment StatementAlignsAlignment BasisThe decay heat removal function is capable of achieving and maintaining safe and stable conditions.hot and cGo-ld shutdoWn. The EC system operates by natural circulation wheresteam flows upward to the condenser(s) and returns as condensate to the Reactor Pressure Vessel. Decay heat is removed through the transfer of heat from the reactor coolant to theshell side water of the EC which vents the developed steam to atmosphere. Operation of either EC loop can sustain Hot Shutdown conditions for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months without the need for makeupfrom the Condensate Storage Tank or Fire Water System. The decay heat removal process also reduces reactor pressure. When reactor pressure is reduced to 120 psig and reactortemperature is reduced to 350 degrees F, the plant can be transitioned to Cold Shutdo.w by inifiating shutdown cooling. The EC system can be used to maintain Hot Shutdownconditions for an extended period of time by using the diesel fire pump to refill the Emergency Condensers as needed. The diesel fire pump draws water from Lake Ontario (essentiallyan infinite source) and need only be refueled periodically to maintain this method of decay heat removal. Note that Emergency Condenser level monitoring is a credited function.In the event the preferred shutdown method is not available, the plant can be shut down by opening three ERVs and discharging steam to the Torus to reduce pressure. When reactorpressure reaches approximately 365 psig, Core Spray (CS) may be initiated. CS is a two loop system. Operation of one loop is adequate to achieve shutdown. Eventually, thereactor vessel floods to the point where the ERVs are passing fluid to the Torus rather than steam, in essence placing the Reactor Coolant System in recirculation through the Torus.During this process, decay heat is removed by operation of the Containment Spray System in conjunction with the Containment Spray Raw Water system. Containment spray andcontainment spray raw water systems are utilized to remove heat from the torus water and maintain it within the core spray and containment spray pumps' net positive suction head(NPSH) requirements. This shutdown method will bring the plant directly to cold shutdown. Fully flooding the Reactor Pressure Vessel negates the need for another system to provideinventory makeup. AC power is required to initiate this shutdown method.Reference DocumentEIR 51-9133191, NSCA, Sections 4.0 and 5.1NMPI, April 2013 Page B-21 I!NMP1, April 2013Page B-21 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.1.2.5 Process MonitoringNEI 00-01 GuidanceThe process monitoring function is provided for all safe shutdown paths. IN 84-09, Attachment 1,Section IX "Lessons Learned from NRCInspections of Fire Protection Safe Shutdown Systems (10 CFR 50 Appendix R)" provides guidance on the instrumentation acceptable toand preferred by the NRC for meeting the process monitoring function. This instrumentation is that which monitors the process variablesnecessary to perform and control the functions specified in Appendix R Section 1I.L.1. Such instrumentation must be demonstrated toremain unaffected by the fire. The IN 84-09 list of process monitoring is applied to alternative/dedicated shutdown (III.G.3). The use ofthis same list for Ill.G.2 redundant Post-Fire Safe Shutdown is acceptable, but the analyst needs to review the specific license basis forthe plant under evaluation. In general, process monitoring instruments similar to those listed below are needed to successfully useexisting operating procedures (including Abnormal Operating Procedures).BWR* Reactor coolant level and pressure* Suppression pool level and temperature* Emergency or isolation condenser level* Diagnostic instrumentation for safe shutdown systems* Level indication for tanks needed for safe shutdownPWR* Reactor coolant temperature (hot leg / cold leg)* Pressurizer pressure and level* Neutron flux monitoring (source range)* Level indication for tanks needed for safe shutdown* Steam generator level and pressure* Diagnostic instrumentation for safe shutdown systemsThe specific instruments required may be based on operator preference, safe shutdown procedural guidance strategy (symptomatic vs.prescriptive), and systems and paths selected for safe shutdown.CommentsNoneApplicabilityApplicableAlignment StatementAlignsAlignment BasisThe process monitoring function is to be provided for all safe shutdown paths. NEI 00-01 refers to NRC IN 84-09, Attachment 1,Section IX, as providing guidance on theinstrumentation acceptable to and preferred by the NRC for meeting the process monitoring function. This instrumentation is that which monitors the process variables necessary toperform and control the functions specified in Appendix R Section III.L.1. Such instrumentation must be demonstrated to remain unaffected by the fire. The IN 84-09 list of processmonitoring is applied to alternative shutdown (III.G.3). In general, process monitoring instruments similar to those listed below are needed to successfully use existing operatingprocedures related to post-fire shutdown (including Abnormal Operating Procedures).INMP1, April 2013Page B-22 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review-Reactor coolant level and pressure-Suppression pool level and temperature-Level indication for tanks needed for safe shutdown-Diagnostic instrumentation for safe shutdown systemsThe Reactor Protection System is utilized to satisfy the process monitoring objectives throughout hot and cold shutdown. The specific instruments required may be based on operatorpreference, safe shutdown procedural guidance strategy (symptomatic vs. prescriptive), and systems and paths selected for safe shutdown.The following process monitoring functions are provided to support post-fire shutdown:Primary and Secondary Methods-Reactor coolant level-Reactor coolant pressure-Reactor coolant temperature-Emergency Diesel Generator parametersPrimary Method-Emergency Condenser levelSecondary Method-Torus level-Torus temperature-Drywell pressure-Drywell temperature-Containment Spray water temperature-Containment Spray pump discharge pressureReference DocumentEIR 51-9133191, NSCA, Sections 4.0 and 5.3NMPI, April 2013 Page B-23 I!NMP1, April 2013Page B-23 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref NEI 00-01 Guidance3.1.2.6 Support Systems Blank Heading -No Specific GuidanceApplicability CommentsApplicable NoneAlignment StatementNot RequiredAlignment BasisGeneric Heading: Alignment discussed in subsequent sub-sections.Reference DocumentNMPI, April 2013 Page B-24INMPI, April 2013Page B-24 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref NEI 00-01 Guidance3.1.2.6.1 Electrical Systems AC Distribution SystemPower for the Appendix R safe shutdown equipment is typically provided by a medium voltage system such as 4.16 KV Class 1E busseseither directly from the busses or through step down transformers/load centers/distribution panels for 600, 480 or 120 VAC loads. Forredundant safe shutdown performed in accordance with the requirements of Appendix R Section III.G.1 and 2, power may be suppliedfrom either offsite power sources or the emergency diesel generator depending on which has been demonstrated to be free of firedamage. No credit should be taken for any beneficial effects of a fire causing a loss of offsite power. Refer to Section 3.1.1.7.DC Distribution SystemTypically, the 125VDC distribution system supplies DC control power to various 125VDC control panels including switchgear breakercontrols. The 125VDC distribution panels may also supply power to the 120VAC distribution panels via static inverters. These distributionpanels may supply power for instrumentation necessary to complete the process monitoring functions.For fire events that result in an interruption of power to the AC electrical bus, the station batteries are necessary to supply any requiredcontrol power during the interim time period required for the diesel generators to become operational. Once the diesels are operational,the 125VDC distribution system can be powered from sources feed from the diesels through the battery chargers.[BWR] Certain plants are also designed with a 250VDC Distribution System that supplies power to Reactor Core Isolation Cooling and/orHigh Pressure Coolant Injection equipment.The DC control centers may also supply power to various small horsepower Appendix R safe shutdown system valves and pumps. If theDC system is relied upon to support safe shutdown without battery chargers being available, it must be verified that sufficient batterycapacity exists to support the necessary loads for sufficient time (either until power is restored, or the loads are no longer required tooperate).Applicability CommentsApplicable NoneAlignment StatementAlignsAlignment BasisNo credit has been taken for the loss of offsite power, however in the event that offsite power is lost, NMP1 has two Emergency Diesel Generators (EDGs). The EDGs will supply therequired medium and low voltage safe shutdown loads with AC power. The EDGs are designed to start automatically on loss of offsite power to re-energize emergency busses 102and 103.The 125V DC distribution panels supply power to the 120V AC distribution panels via static inverters. These distribution panels typically supply power for instrumentation necessary tocomplete process monitoring functions. The 125V DC distribution system supplies control power to various 125V DC control panels including switchgear breaker controls. Vital ACpower can be provided via RPS uninterruptible power supplies (UPS) 162A, 162B, 172A and 172B.This 125V DC distribution system is credited to support post fire safe shutdown. For fire events that result in an interruption of power to the AC electrical bus, the station batteriessupply any required control power during the interim time period required for the EDGs to become operational. Once the EDGs are operational, the 125V DC distribution system canbe powered from the diesels through the battery chargers.Reference DocumentEIR 51-9133191, NSCA, Sections 4.0 and 5.3!NMP1, April 2013Page B-25 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref NEI 00-01 Guidance3.1.2.6.2 Cooling Systems Various cooling water systems may be required to support safe shutdown system operation, based on plant-specific considerations.Typical uses include:* RHR/SDC/DH Heat Exchanger cooling water* Safe shutdown pump cooling (seal coolers, oil coolers)* Diesel generator coolingApplicability CommentsApplicable NoneAlignment StatementAlignsAlignment BasisVarious cooling water systems are required to support safe shutdown system operation:-The Reactor Building Closed Loop Cooling and Emergency Service Water Systems provide cooling to the control room ventilation system and Shutdown Cooling Heat Exchangers-The EDGs are cooled by the EDG Raw Water pumps-The Containment Spray and Containment Spray Raw Water Systems cool the torus for the secondary cooldown method-The Chilled Water System is provided for control room ventilationReference DocumentEIR 51-9133191, NSCA, Section 5.3NMPI, April 2013 Page B-26 IINMP1, April 2013Page B-26 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref NEI 00-01 Guidance3.1.2.6.3 HVAC Systems HVAC Systems may be required to assure that safe shutdown equipment remains within its operating temperature range, as specified inmanufacturer's literature or demonstrated by suitable test methods, and to assure protection for plant operations staff from the effects offire (smoke, heat, toxic gases, and gaseous fire suppression agents).HVAC systems, however, are not required to support post-fire safe shutdown in all cases. The need for HVAC system operation is basedon plant specific configurations and plant specific heat loads. Typical potential uses include:* Main control room, cable spreading room, relay room* ECCS pump compartments* Diesel generator rooms* Switchgear roomsPlant specific evaluations are necessary to determine which HVAC systems could be required or useful in supporting post-fire safeshutdown. Transient temperature response analyses are often utilized to demonstrate that specific HVAC systems would or would not berequired. If HVAC systems are credited, the potential for adverse fire effects to the HVAC system must also be considered, including:* Dampers closing due to direct fire exposure or due to hot gases flowing through ventilation ducts from the fire area to an areanot directly affected by the fire. Where provided, smoke dampers should consider similar effects from smoke.* Recirculation or migration of toxic conditions (e.g., smoke from the fire, suppressants such as Carbon Dioxide).In certain situations, adequate time exists to open doors to provide adequate cooling to allow continued equipment operation. Therefore,the list of required safe shutdown components as it relates to HVAC Systems may be determined based on transient temperatureanalysis. Should this analysis demonstrate that adequate time exists to open doors to provide the necessary cooling, this is anacceptable approach to achieving HVAC Cooling. The temperature analysis must demonstrate the adequacy of the cooling effect fromopening the door within the specified time. Only those components whose operation is required to provide HVAC Cooling for requiredsafe shutdown components in a time frame that cannot be justified for operator manual actions are considered themselves to be requiredsafe shutdown components. This latter set of HVAC Cooling Components are required to meet the criteria for required safe shutdowncomponents with regard to the available mitigating tools.Applicability CommentsApplicable NoneAlignment StatementAlignsAlignment BasisHVAC Systems required to support post-fire shutdown are:-Main control room ventilation (except for control room evacuation)-EDG rooms (fans and roll-up doors)Paths developed for the control room ventilation system do not necessarily match the paths for process systems. Paths for the ventilation system are as follows:-Path A -Recirculation flow path-Path B -Smoke purge flow pathINMPI, April 2013Page B-27 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review-Path C -Train 11 positive pressure flow path.-Path D -Train 12 positive pressure flow pathGiven the failure modes of dampers and EDG power requirement, Paths C & D are generally credited for post-fire shutdown. Any ventilation path can be supported by Train 11 orTrain 12 Chilled Water.Reference DocumentEIR 51-9133191, NSCA, Section 5.3NMPI, April 2013 Page B-28 I!NMPI, April 2013Page B-28 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.1.3 Methodology For ShutdownSystem CollectionNEI 00-01 GuidanceRefer to Figure 3-2 for a flowchart illustrating the various steps involved in selecting safe shutdown systems and developing the shutdownpaths. The following methodology may be used to define the safe shutdown systems and paths for an Appendix R analysis:ApplicabilityApplicableCommentsNoneAlignment StatementNot RequiredAlignment BasisThis paragraph provides introductory information and contains no specific requirements. Discussion is provided in subsequent sub-sections.Reference DocumentNMPI, April 2013 Page B-29 IINMPI, April 2013Page B-29 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.1.3.1 Identify Safe ShutdownFunctionsNEI 00-01 GuidanceReview available documentation to obtain an understanding of the available plant systems and the functions required to achieve andmaintain safe shutdown. Documents such as the following may be reviewed:Operating Procedures (Normal, Emergency, Abnormal)System descriptionsFire Hazard AnalysisSingle-line electrical diagramsPiping and Instrumentation Diagrams (P&IDs)[BWR] GE Report GE-NE-T43-00002-00-01-R02 entitled "Original Shutdown Paths for the BWR"ApplicabilityApplicableAlignment StatementAlignsCommentsNoneAlignment BasisSafe shutdown systems and functions required to satisfy the safe shutdown performance goals were developed and identified from available plant documentation. This documentationincludes but is not limited to electrical one line diagrams, schematics, piping and instrumentation diagrams (P&lDs), operating procedures, UFSAR, Fire Hazards Analysis, and thesystems descriptions.Reference DocumentEIR 51-9133191, NSCA, Section 5.4NM P1, April 2013 Page B-30 IINMP1, April 2013Page B-30 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.1.3.2 Identify Combinations ofSystems That Satisfy Each SafeShutdown FunctionApplicabilityApplicableAlignment StatementAlignsNEI 00-01 GuidanceGiven the criteria/assumptions defined in Section 3.1.1, identify the available combinations of systems capable of achieving the safeshutdown functions of reactivity control, pressure control, inventory control, decay heat removal, process monitoring and support systemssuch as electrical and cooling systems (refer to Section 3.1.2). This selection process does not restrict the use of other systems. Inaddition to achieving the required safe shutdown functions, consider other equipment whose mal-operation or spurious operation couldimpact the required safe shutdown function. The components in this latter set are classified as either required for hot shutdown or asimportant to SSD as explained in Appendix H.CommentsAlignment BasisSafe shutdown systems, components, and cables were selected using the criteria and assumptions of NEI 00-01, Sections 3.1.1 and 3.1.2 to satisfy each safe shutdown functionperformance goal, including process monitoring and support systems. NEI 00-01 was used as the guidance in developing the Safe Shutdown Equipment List and the Safe Shutdownsuccess paths.The following combination of systems are capable of achieving safe shutdown functions.1 ) Reactivity Control -Control Rod Drive SystemThe safe shutdown performance and design requirements for the reactivity control function can be met without automatic scram capability. The post-fire safe shutdown analysis needonly provide the capability to manually scram the reactor.2) Pressure Control Systems -ERVsThe ERVs are opened to maintain hot shutdown conditions or to depressurize the vessel to allow injection using low pressure systems. These are operated manually. Automaticinitiation of the ADS is not a required function.3) Inventory ControlSystems selected for the inventory control function are capable of supplying sufficient reactor coolant to maintain the fuel in a safe and stable condition.a.hiev... and maintain hotshuewri.- Manual initiation of these systems is acceptable. Automatic initiation functions are not required.4) Decay Heat RemovalSystems selected for the decay heat removal function(s) are capable of:Removing sufficient decay heat from primary containment to prevent containment over-pressurization and failure.Satisfying the net positive suction head requirements of any safe shutdown systems taking suction from the tows.Removing sufficient decay heat from the reactor to maintain the fuel in a safe and stable condition.aa-.hia-' sh'tdo'wn.!NMP1, April 2013Page B-31 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review5) Process MonitoringThe process monitoring function is to be provided for all safe shutdown paths. NEI 00-01 refers to NRC IN 84-09, Attachment 1,Section IX, as providing guidance on theinstrumentation acceptable to and preferred by the NRC for meeting the process monitoring function. This instrumentation is that which monitors the process variables necessary toperform and control the functions specified in Appendix R Section III.L.I. Such instrumentation must be demonstrated to remain unaffected by the fire. The IN 84-09 list of processmonitoring functions is applied to alternative shutdown (III.G.3). In general, process monitoring instruments similar to those listed below are needed to successfully use existingoperating procedures related to post-fire shutdown (including Abnormal Operating Procedures).-Reactor coolant level and pressure-Torus level and temperature-Level indication for tanks needed for safe shutdown-Diagnostic instrumentation for safe shutdown systemsThe specific instruments required may be based on operator preference, safe shutdown procedural guidance strategy (symptomatic vs. prescriptive), and systems and paths selectedfor safe shutdown.Note -For NMP1, Emergency Condenser level is a required process monitoring function.6) Support SystemsA. Electrical SystemsAC Distribution SystemPower for safe shutdown equipment is supplied from either offsite power sources or the emergency diesel generator. No credit is taken for a fire causing a loss of offsite power.DC Distribution SystemThe 125V DC distribution system supplies control power to various 125V DC control panels including switchgear breaker controls. The 125V DC distribution panels also supply powerto the 120V AC distribution panels via static inverters. These distribution panels supply power for instrumentation necessary to complete process monitoring functions. For fire eventsthat result in an interruption of power to the AC electrical bus, the station batteries are necessary to supply any required control power during the interim time period required for the-diesel generators to become operational. Once the diesels are operational, the 125V DC distribution systems can be powered from the diesels through the battery chargers.B. Cooling SystemsVarious cooling water systems are required to support safe shutdown system operation, based on plant-specific considerations. Cooling system uses include:-SDC Heat Exchanger cooling water-Safe shutdown pump cooling (seal coolers, oil coolers)-Diesel generator cooling-HVAC system cooling waterC. HVAC SystemsHVAC Systems are required to assure that safe shutdown equipment remains within its operating temperature range, as specified in manufacturer's literature or demonstrated bysuitable test methods, and to assure protection for plant operations staff from the effects of fire (smoke, heat, toxic gases, and gaseous fire suppression agents).NMPI, April 2013 Page B-32 IINMP1, April 2013Page B-32 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology ReviewHVAC uses include:-Main control room, cable spreading room, relay room-ECCS pump compartments-Diesel generator rooms-Switchgear RoomsPaths developed for the control room ventilation system do not necessarily match the paths for process systems. Paths for the ventilation system are as follows:-Path A -Recirculation flow path-Path B -Smoke purge flow path-Path C -Train 11 positive pressure flow path.-Path D -Train 12 positive pressure flow pathGiven the failure modes of dampers and EDG power requirement, Paths C & D are generally credited for post-fire shutdown. Any ventilation path can be supported by Train 11 orTrain 12 Chilled Water.Reference DocumentEIR 51-9133191, NSCA, Sections 4.0, 5.0, and 5.4NMPI, April 2013 Page B-33 IINMP1, April 2013Page B-33 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.1.3.3 Define Combination ofSystems for Each Safe ShutdownPathNEI 00-01 GuidanceSelect combinations of systems with the capability of performing all of the required safe shutdown functions and designate this set ofsystems as a safe shutdown path. In many cases, paths may be defined on a divisional basis since the availability of electrical power andother support systems must be demonstrated for each path. During the equipment selection phase, identify any additional supportsystems and list them for the appropriate path.ApplicabilityApplicableCommentsNoneAlignment StatementAlignsAlignment BasisSafe shutdown systems, components, and cables were selected using the criteria and assumptions of NEI 00-01, Sections 3.1.1 and 3.1.2 to satisfy each safe shutdown functionperformance goal, including process monitoring and support systems for each path. NEI 00-01 was used as the guidance in developing the Safe Shutdown Equipment List and theSafe Shutdown success paths. The shutdown paths and equipment selection for the shutdown performance goals are identified in the NSCA.Reference DocumentEIR 51-9133191, NSCA, Sections 4.0 and 5.0NMPI, April 2013 Page 6-34 I!NMP1, April 2013Page B-34 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.1.3.4 Assign Shutdown Paths toEach Combination of SystemsNEI 00-01 GuidanceAssign a path designation to each combination of systems. The path will serve to document the combination of systems relied upon forsafe shutdown in each fire area. Refer to Attachment I to this document for an example of a table illustrating how to document thevarious combinations of systems for selected shutdown paths.ApplicabilityApplicableCommentsNoneAlignment StatementAlignsAlignment BasisSafe shutdown systems, components, and cables were selected using the criteria and assumptions of NEI 00-01, Sections 3.1.1 and 3.1.2 to satisfy each safe shutdown functionperformance goal, including process monitoring and support systems for each path. NEI 00-01 was used as the guidance in developing the Safe Shutdown Equipment List and theSafe Shutdown success paths. Safe Shutdown Paths were designed based on the combination of systems in the respective fire area.The major systems for safe shutdown success paths are as follows:SUCCESS PATH "A"Hot shutdown is achieved via the use of ECs 111 & 112 for Decay Heat Removal. The DFP is used as an inventory source for the EC system. Cold ANhutdwn is aseempliched Aia theuse of Tra-in 11 SDC System components supported by the T-Fain I1 RBCLC and ESW Systems. The Train 11 CRD pump or the DFP is used for inventory control/RPV makeup. PathA is generally supported by the Train 11 AC and DC power systems. Some components in Path A systems are powered by the Train 12 AC and/or DC power system while othercomponents may require local operator action due to a potential loss of instrument air.SUCCESS PATH "B"Hot shutdown is achieved via the use of ECs 121 & 122 for Decay Heat Removal. The DFP is used as an inventory source for the EC system ...ld. h..dWn.. is acmp.sh.d vA th,use of Tahin 12 SDC System campnotSM. Supported by the TrFain 12 RBCLC and ESW Systems. The Train 12 CRD pump or the DFP is used for inventory control/RPV makeup. PathB is generally supported by the Train 12 AC and DC power systems. Some components in Path B systems are powered by the Train 11 AC and/or DC power system while othercomponents may require local operator action due to a potential loss of instrument air.SUCCESS PATH "C"Rath hot and cold shu tdo', aroHot shutdown is achieved via use of the Train 11 ERVs (PSV-01-102A, PSV-01-102B, PSV-01-102E) to depressurize the reactor. Whenpressure drops to the appropriate level, the CS System is used to flood the vessel and place the RCS in a recirculation mode to the Torus. Torus cooling is provided by the CTSsystem supported by the CTSRW System. These systems can be are-maintained in service to directly achieve cold shutdown conditions. Path C is generally supported by the Train11 AC and DC power systems. Some components in Path C systems are powered by the Train 12 AC and/or DC power system.NMPI, April 2013 Page B-35 IINMP1, April 2013Page B-35 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology ReviewSUCCESS PATH "D"Hot shutdown is -ot- hot chu-tdo4-. and "old h-chutdo.a, are achieved via use of the Train 12 ERVs (PSV-01-102C, PSV-01-102D, PSV-01-102F) to depressurze the reactor. When Ipressure drops to the appropriate level, the CS System is used to flood the vessel and place the RCS in a recirculation mode to the Torus. Torus cooling is provided by the CTSSystem supported by the CTSRW System. These systems can be are-maintained in service to directly achieve cold shutdown conditions. Path D is generally supported by the Train I12 AC and DC power systems. Some components in Path D systems are powered by the Train 11 AC and/or DC power system.Path A and Path B are the preferred shutdown paths to maintain the fuel in a safe and stable condition for both hot sh-Atdo.- and "ald r-hutdoA.- as these paths present the least Ithermal hydraulic impact to the plant.It is possible that for some fire areas, one Path may be employed for hot shutdown and another for cold shutdown. This depends upon the impacts to power supplies and othercomponents in any particular fire area, the controls design of credited components and the use of damage repair procedures. Consequently, it is possible that Path B may be creditedfor hot shutdown and Path D for cold shutdown, or any other combination. However, if Path C or D is credited for hot shutdown, that path would also be credited for cold shutdown dueto the dynamics of the shutdown process.Reference DocumentEIR 51-9133191, NSCA, Section 5.3Closure of National Fire Protection Association 805 Transition Program Frequently Asked Question Number 08-0054 (ML110140183)NMPI, April 2013 Page B-36 IlNMP1, April 2013Page B-36 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.2 Safe Shutdown EquipmentSelectionNEI 00-01 GuidanceThe previous section described the methodology for selecting the systems and paths necessary to achieve and maintain safe shutdownfor an exposure fire event (see Section 5.0 DEFINITIONS for "Exposure Fire"). This section describes the criteria/assumptions andselection methodology for identifying the specific safe shutdown equipment necessary for the systems to perform their Appendix Rfunctions. The selected equipment should be related back to the safe shutdown systems that they support and be assigned to the samesafe shutdown path as that system. The list of safe shutdown equipment will then form the basis for identifying the cables necessary forthe operation or that can cause the mal-operation of the safe shutdown systems. For each path it will be important to understand whichcomponents are classified as required safe shutdown components and which are classified as important to safe shutdown components.When evaluating the fire-induced impact to each affected cable/component in each fire area, this classification dictates the tools availablefor mitigation the affects.ApplicabilityApplicableCommentsNoneAlignment StatementNot RequiredAlignment BasisThis paragraph provides introductory information and contains no specific requirements. Discussion is provided in subsequent sub-sections.Reference DocumentNMPI, April 2013 Page B-37 IINMP1, April 2013Page B-37 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref NEI 00-01 Guidance3.2.1 Criteria/Assumptions Consider the following criteria and assumptions when identifying equipment necessary to perform the required safe shutdown functions:Applicability CommentsApplicable NoneAlignment StatementNot RequiredAlignment BasisThis paragraph provides introductory information and contains no specific requirements. Discussion is provided in subsequent sub-sections.Reference DocumentNMPI, April 2013 Page B-38 I!NMP1, April 2013Page B-38 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref NEI 00-01 Guidance3.2.1.1 Safe Shutdown Equipment Safe shutdown equipment can be divided into two categories. Equipment may be categorized as (1) primary components or (2) secondaryCategories components. Typically, the following types of equipment are considered to be primary components:* Pumps, motor operated valves, solenoid valves, fans, gas bottles, dampers, unit coolers, etc.* All necessary process indicators and recorders (i.e., flow indicator, temperature indicator, turbine speed indicator, pressure indicator,level recorder).* Power supplies or other electrical components that support operation of primary components (i.e., diesel generators, switchgear,motor control centers, load centers, power supplies, distribution panels, etc.).Secondary components are typically items found within the circuitry for a primary component. These provide a supporting role to theoverall circuit function. Some secondary components may provide an isolation function or a signal to a primary component via either aninterlock or input signal processor. Examples of secondary components include flow switches, pressure switches, temperature switches,level switches, temperature elements, speed elements, transmitters, converters, controllers, transducers, signal conditioners, handswitches, relays, fuses and various instrumentation devices.Applicability CommentsApplicable NoneAlignment StatementAlignsAlignment BasisThe Safe Shutdown Equipment List (SSEL) is a list of analyzed components that are utilized in the post-fire safe shutdown analysis to ensure that one success path (structures,systems, and components) necessary to achieve safe shutdown is free of fire damage without crediting plant or system repair.The current NMP1 SSEL and Safe Shutdown Analysis was reviewed against the criteria outlined in NEI 00-01 (Sections 3.1 and 3.2) for safe shutdown systems and equipmentselection. This review addressed potential fire induced circuit failure issues, either within or beyond the plant's existing licensing basis. Additional equipment has been included toaddress any multiple spurious operation concerns.The SSEL is divided into primary and secondary components. Primary and secondary components were defined as being consistent with NEI 00-01 guidance. Equipment identified asprimary components is included in the SSEL. Equipment identified as secondary components is included in the SSEL Database with the primary component(s) that would be affectedby fire damage to the secondary component. By doing this, the SSEL is kept to a manageable size and the equipment included in the SSEL can be readily related to required post-firesafe shutdown systems and functions.Secondary components were generally combined with primary components except where groups of secondary components were defined as "pseudo-components."A "pseudo-component" is an artificial association of equipment and cables that perform a common function as a single entity for analysis purposes. The "pseudo-component" isassigned for analysis purposes only and is not an actual plant hardware designation. The concept of a "pseudo-component" was developed to account for those cables whichconstitute a circuit common to several components. The use of "pseudo-components" precludes the need to repeat cable selection and circuit analysis of these cables for eachprimary component. This generic name is interlocked with the affected primary components for analysis purposes and it inherits the attributes (path, system, train, etc.) of thecomponents that it may affect. The nomenclature of the "pseudo-component" is similar to other equipment as defined in the plant equipment database.NMPI, April 2013 Page B-39 IINMPI, April 2013Page B-39 I Constellation Energy Nuclear GroupReference DocumentEIR 51-9133191, NSCA, Section 5.4EIR 51-9177678-000, Definitions SectionAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology ReviewNMPI, April 2013 Page B-40 INMP1, April 2013Page B-40 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.2.1.2 Manual Valves and PipingNEI 00-01 GuidanceAssume that exposure fire damage to manual valves and piping does not adversely impact their ability to perform their pressure boundaryor safe shutdown function (heat sensitive piping materials, including tubing with brazed or soldered joints, are not included in thisassumption). Fire damage should be evaluated with respect to the ability to manually open or close the valve should this be necessary asa part of the post-fire safe shutdown scenario. For example, post-fire coefficients of friction for rising stem valves cannot be readilydetermined. Handwheel sizes and rim pulls are based on well lubricated stems. Any post-fire operation of a rising stem valve should bewell justified using an engineering evaluation.ApplicabilityApplicableCommentsInstrument tubing failure damage due to a fire is addressed in NSCA ( EIR 51-9133191, Section 8.4).Alignment StatementAligns With IntentAlignment BasisThe NMP1 fire area assessments assume that an exposure fire does not adversely affect the ability of manual valves and piping to perform their pressure boundary or safe shutdownfunction. Fire damage to valves has been evaluated with respect to the ability to manually open or close the valve should this be necessary as a part of the post-fire safe shutdownscenario. Post-fire operation of manual valves within the affected fire area has been evaluated in the Fire Area Analysis.Instrument sensing lines were reviewed for susceptibility to physical fire damage that may cause a loss of inventory. Sensing lines for SSEL components are constructed of eitherstainless steel or carbon steel. Consequently, they are not susceptible to physical damage as the result of a postulated fire.Reference DocumentEIR 51-9133191, NSCA, Sections 5.4 and 8.4HNP RAI 3-15, NRC Request for Additional Information dated August 6, 2009 (ML092170715)NMPI, April 2013 Page 6-41 I!NMP1, April 2013Page B-41 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref NEI 00-01 Guidance3.2.1.3 Valves in Normal Position Assume that all components, including manual valves, are in their normal position as shown on P&IDs or in the plant operatingprocedures, that there are no LCOs in effect, that the Unit is operating at 100% power and that no equipment has been taken out ofservice for maintenance.Applicability CommentsApplicable NoneAlignment StatementAlignsAlignment BasisManual valves are assumed to be in their normal operating position as shown on P&IDs or as identified in the plant operating procedures.CommentsNoneReference DocumentEIR 51-9133191, NSCA, Sections 5.4, 8.1, and 9.0NMPI, April 2013 Page B-42 IINMP1, April 2013Page B-42 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.2.1.4 Check Valves NEI 00-01 GuidanceAssume that a check valve closes in the direction of potential flow diversion and seats properly with sufficient leak tightness to preventflow diversion. Therefore, check valves do not adversely affect the flow rate capability of the safe shutdown systems being used forinventory control, decay heat removal, equipment cooling or other related safe shutdown functions.ApplicabilityApplicableCommentsNoneAlignment StatementAlignsAlignment BasisCheck valves are assumed to close in the direction of potential flow diversion and seat properly with sufficient leak tightness to prevent flow diversion or inventory loss. Check valvesdo not adversely affect flow rate capability of the safe shutdown systems.CommentsNoneReference DocumentEIR 51-9133191, NSCA, Section 5.4NMPI, April 2013 Page B-43 IINMP1, April 2013Page B-43 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.2.1.5 Instrument FailureNEI 00-01 GuidanceInstruments (e.g., resistance temperature detectors, thermocouples, pressure transmitters, and flow transmitters) are assumed to failupscale, midscale, or downscale as a result of fire damage, whichever is worse. An instrument performing a control function is assumedto provide an undesired signal to the control circuit.ApplicabilityApplicableCommentsNoneAlignment StatementAlignsAlignment BasisInstruments are assumed to fail in the most undesirable worst state, whether upscale, downscale, or mid-scale. An instrument providing a control function is assumed to provide anundesired signal to the control circuit.CommentsNoneReference DocumentEIR 51-9133191, NSCA, Sections 5.4 and 8.1NMPI, April 2013 Page B-~ IINMP1, April 2013Page B-44 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref NEI 00-01 Guidance3.2.2 Methodology For Equipment Refer to Figure 3-3 for a flowchart illustrating the various steps involved in selecting safe shutdown equipment. Use the followingSelection methodology to select the safe shutdown equipment for a post-fire safe shutdown analysis:Applicability CommentsApplicable NoneAlignment StatementNot RequiredAlignment BasisThis paragraph provides introductory information and contains no specific requirements. Discussion is provided in subsequent sub-sections.Reference DocumentNMPI, April 2013 Page B-45 IINMP1, April 2013Page B-45 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.2.2.1 Identify the System FlowPath for Each Shutdown PathNEI 00-01 GuidanceMark up and annotate a P&ID to highlight the specific flow paths for each system in support of each shutdown path. Refer to Attachment2 for an example of an annotated P&ID illustrating this concept When developing the SSEL, determine which equipment should beincluded on the Safe Shutdown Equipment List (SSEL). As an option, include secondary components with a primary component(s) thatwould be affected by fire damage to the secondary component. By doing this, the SSEL can be kept to a manageable size and theequipment included on the SSEL can be readily related to required post-fire safe shutdown systems and functions.ApplicabilityApplicableCommentsNoneAlignment StatementAlignsAlignment BasisThe Safe Shutdown Equipment List (SSEL) is a list of analyzed components that are utilized in the post-fire safe shutdown analysis to ensure that one success path (structures,systems, and components) necessary to maintain the fuel in a safe and stable condition.'eac-hi cAfe ch.ut-doa:.Afr- e oIs f fire damag. eith"ut Gediting plnt OFr repair. TheSSEL is divided into primary and secondary components. Primary and secondary components were defined as being consistent with NEI 00-01 section 3.2.1.1 guidance. Secondarycomponents were generally combined with primary components.Combinations of components and systems with the capability to satisfy the required NFPA 805 performance goals safo funtone were designated as safe shutdown flowpaths. P&IDs and Electrical drawings were marked up and annotated to highlight the selected primary safe shutdown equipment and flow paths for each system in support of eachshutdown path. The specific group of equipment supporting each system was populated into the safe shutdown database.Reference DocumentEIR 51-9133191, NSCA, Sections 4.0 and 6.0INMPI, April 2013 Page B-46 I!NMP1, April 2013Page B-46 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.2.2.2 Identify the Equipment inEach Safe Shutdown System FlowPathNEI 00-01 GuidanceReview the applicable documentation (e.g. P&IDs, electrical drawings, instrument loop diagrams) to assure that all equipment in eachsystem's flow path has been identified. Assure that any equipment that could spuriously operate and adversely affect the desired systemfunction(s) is also identified. Additionally, refer to Section 4 for the Resolution Methodology for determining the Plant Specific List ofMSOs requiring evaluation. Criteria for making the determination as to which of these components are to be classified as required for hotshutdown or as important to SSD is contained in Appendix H. If additional systems are identified which are necessary for the operation ofthe safe shutdown system under review, include these as required for hot shutdown systems. Designate these new systems with thesame safe shutdown path as the primary safe shutdown system under review (Refer to Figure 3-1).ApplicabilityApplicableCommentsNoneAlignment StatementAlignsAlignment BasisThe safe shutdown flow paths identify the primary components that are required to meet the safe shutdown performance goals. The safe shutdown components were compiled basedon each system's performance and safe shutdown function. These components establish the primary safe shutdown flowpath for system operation. Also included in the safe shutdownflow paths are those components whose spurious operation could impact safe shutdown system operability. Systems identified as necessary for the operation of the safe shutdownsystem under review are included in the safe shutdown equipment list and designated with the same shutdown path as the primary safe shutdown system. These components mayinvolve branch flow paths that must be isolated and remain isolated to assure that flow will not be diverted from the primary flow path. The list of primary components may also includeselected mechanical components required to support safe shutdown.The criteria used in evaluating spurious actuation of components are those identified in NEI 00-01, Section 4, Identification and Treatment of Multiple Spurious Operations (MSO). TheNuclear Safety Capability Fire Area Assessments includes the potential impact of multiple spurious component actuations per the guidance of NEI 00-01. MSO componentcombinations, as documented in the Technical Report on Identification & Classification of the NMP1 MSO Scenarios Using an Expert Panel -Review of New Generic Scenarios, wereaddressed in EIR 51-9133191 and included in the fire area assessments.Reference DocumentEIR 51-9133191, NSCA, Sections 4.0, 5.0, 6.0, 8.6.1, and 8.6.3Technical Report on Identification & Classification of the NMP-1 MSO Scenarios Using an Expert Panel -Review of New Generic ScenariosNMPI, April 2013 Page B-47 IINMPI, April 2013Page B-47 i Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.2.2.3 Assign Safe Shutdown FlowPathsNEI 00-01 GuidancePrepare a table listing the equipment identified for each system and the shutdown path that it supports. Identify any valves or otherequipment that could spuriously operate and impact the operation of that safe shutdown system. Criteria for making the determination asto which of these components are to be classified as required for hot shutdown or as important to SSD is contained in Appendix H.Assign the safe shutdown path for the affected system to this equipment. During the cable selection phase, identify additional equipmentrequired to support the safe shutdown function of the path (e.g., electrical distribution system equipment). Include this additionalequipment in the safe shutdown equipment list. Attachment 3 to this document provides an example of a (SSEL). The SSEL identifiesthe list of equipment within the plant considered for post-fire safe shutdown and it documents various equipment-related attributes used inthe analysis.Identify instrument tubing that may cause subsequent effects on instrument readings or signals as a result of fire. Determine and considerthe fire area location of the instrument tubing when evaluating the effects of fire damage to circuits and equipment in the fire area.ApplicabilityApplicableCommentsNoneAlignment StatementAlignsAlignment BasisThe Safe Shutdown Equipment List (SSEL) includes equipment for each system supporting the flow paths needed to achieve safe shutdown. This equipment, identified from thehighlighted P&IDs, includes both the normal and diversion flow paths required to meet the system performance goals and safe shutdown functions. These components also includevalves or equipment that could impact safe shutdown by spuriously operating or whose failure would threaten the capability to achieve safe shutdown. The components werepopulated into the SSEL database and assigned to safe shutdown system success paths.During the cable selection process additional support components such as electrical distribution equipment were added to the SSD paths and populated into the database. Thedatabase reports produce tables listing equipment and related information which is very similar to the table provided in Attachment 3, of NEI 00-01. This group of components and thevarious equipment related attributes makes up the SSEL.Instrument sensing lines for level, pressure, flow, etc. that are exposed to a fire are considered to have the potential of causing erratic or unreliable indication. The instrument tubinglines were traced and their routes correlated to fire areas. Cable identifications were given to the sensing lines and were subjected to the same compliance issues and analyticaltechniques as safe shutdown cables and similarly analyzed for separation in the fire area assessments.Reference DocumentEIR 51-9133191, NSCA, Sections 4.0, 5.3, 6.0, and 8.4NMPI, April 2013 Page B-48 IINMP1, April 2013Page B-48 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.2.2.4 Identify EquipmentInformation Required for the SafeShutdown AnalysisNEI 00-01 GuidanceCollect additional equipment-related information necessary for performing the post-fire safe shutdown analysis for the equipment. In orderto facilitate the analysis, tabulate this data for each piece of equipment on the SSEL. Refer to Attachment 3 to this document for anexample of a SSEL. Examples of related equipment data should include the equipment type, equipment description, safe shutdownsystem, safe shutdown path, drawing reference, fire area, fire zone, and room location of equipment. Other information such as thefollowing may be useful in performing the safe shutdown analysis: normal position, hot shutdown position, cold shutdown position, failedair position, failed electrical position, high/low pressure interface concem, and spurious operation concern. Criteria for making thedetermination as to which of these components are to be classified as required for hot shutdown or as important to SSD is contained inAppendix H.ApplicabilityApplicableCommentsNoneAlignment StatementAlignsAlignment BasisThe systems and components of the Safe Shutdown Equipment List (SSEL) were selected to meet the NFPA 805 performance goals to ensure the fuel remains in a safe and stablecondition.ashieve post fire chutd~m' a- spo.,fied I 1.CFR5O, Appendix R. Additional secondary components which are modeled as a result of the primary component selectionswere also populated into the database.This additional equipment related information necessary to perform the Fire Area Assessments was collected and included in the SSEL. The SSEL contains the following information;system, train, component, component description, path, Hi/Lo pressure interface determination, normal position, hot shutdown position, cold shutdown position, failed electricalposition, failed air position, Fire Area, and Fire Zone.The SSEL database contains equipment and related information similar to the information identified in Attachment 3 of NEI 00-01. The SSEL contains the primary components whichare required for hot shutdown. The secondary components are typically items found within the circuitry for a primary component and provide a supporting role. Components that areimportant to safe shutdown are all components not classified as required for hot shutdown.Reference DocumentEIR 51-9133191, NSCA, Sections 4.0 and 6.0NMPI, April 2013 Page B-49 IINMP1, April 2013Page B-49 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.1 Nuclear Safety Capability Systems and Equipment SelectionNEI 00-01 Ref3.2.2.5 Identify DependenciesBetween Equipment, SupportingEquipment, Safe Shutdown Systemsand Safe Shutdown Paths.NEI 00-01 GuidanceIn the process of defining equipment and cables for safe shutdown, identify additional supporting equipment such as electrical power andinterlocked equipment. As an aid in assessing identified impacts to safe shutdown, consider modeling the dependency betweenequipment within each safe shutdown path either in a relational database or in the form of a Safe Shutdown Logic Diagram (SSLD).Attachment 4 provides an example of a SSLD that may be developed to document these relationships.ApplicabilityApplicableCommentsNoneAlignment StatementAlignsAlignment BasisAs part of the process of preparing the Safe Shutdown Equipment List (SSEL) and defining equipment and cables for safe shutdown, additional equipment and cables that support theSSEL components were identified (such as interlocked components, normal and alternate electrical power supplies, cascading power supplies). The process included development ofa cascading interlock analysis.This information was populated into a relational type database necessary to analyze for post-fire safe shutdown.Reference DocumentEIR 51-9133191, NSCA, Sections 4.0, 5.0, and 6.0NMPI, April 2013 Page B-50 IlNMP1, April 2013Page B-50 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.2 Nuclear Safety Capability Circuit Analysis2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety functions shall be identified. This includes circuits that are required for operation, thatcould prevent the operation, or that result in the mal-operation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts(external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safetyperformance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated.2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shallbe evaluated for their impact on the ability to achieve nuclear safety performance criteria.(a) Common Power Supply Circuits. Those circuits whose fire induced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shallbe identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure couldcause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faultson inadequately protected cables or via inadequately sealed fire area boundaries.NEI 00-01 Ref3.3 Safe Shutdown Cable SelectionAnd LocationNEI 00-01 GuidanceThis section provides industry guidance on one acceptable approach for selecting safe shutdown cables and determining their potentialimpact on equipment required for achieving and maintaining safe shutdown of an operating nuclear power plant for the condition of anexposure fire. The Appendix R safe shutdown cable selection criteria are developed to ensure that all cables that could affect the properoperation or that could cause the mal-operation of safe shutdown equipment are identified and that these cables are properly related tothe safe shutdown equipment whose functionality they could affect. Through this cable-to-equipment relationship, cables become part ofthe safe shutdown path assigned to the equipment affected by the cable. The classification of a cable as either an important to SSDcircuit cable or a required safe shutdown cable is also derived from the classification applied to the component that it supports. Thisclassification can vary from one fire area to another depending on the approach used to accomplish post-fire safe shutdown in the area.Refer to Appendix H for the criteria to be used for classifying required and important to SSD components.ApplicabilityApplicableCommentsNoneAlignment StatementNot RequiredAlignment BasisThis paragraph provides introductory information and contains no specific requirements. Discussion is provided in subsequent sub-sections.Reference DocumentNMPI, April 2013 Page B-51INMP1, April 2013Page B-51 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.2 Nuclear Safety Capability Circuit AnalysisNEI 00-01 Ref3.3.1 Criteria/AssumptionsNEI 00-01 GuidanceTo identify an impact to safe shutdown equipment based on cable routing, the equipment must have cables that affect it identified.Carefully consider how cables are related to safe shutdown equipment so that impacts from these cables can be properly assessed interms of their ultimate impact on safe shutdown system equipment.Consider the following criteria when selecting cables that impact safe shutdown equipment:ApplicabilityApplicableCommentsNoneAlignment StatementNot RequiredAlignment BasisThis paragraph provides introductory information and contains no specific requirements. Discussion is provided in subsequent sub-sections.Reference DocumentNMPI, April 2013 Page B-52 IINMP1, April 2013Page B-52 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.2 Nuclear Safety Capability Circuit AnalysisNEI 00-01 Ref3.3.1.1.1 Cable FailuresNEI 00-01 GuidanceThe list of cables whose failure could impact the operation of a piece of safe shutdown equipment includes more than those cablesconnected to the equipment. The relationship between cable and affected equipment is based on a review of the electrical or elementarywiring diagrams. To assure that all cables that could affect the operation of the safe shutdown equipment are identified, investigate thepower, control, instrumentation, interlock, and equipment status indication cables related to the equipment. Review additional schematicdiagrams to identify additional cables for interlocked circuits that also need to be considered for their impact on the ability of theequipment to operate as required in support of post-fire safe shutdown. As an option, consider applying the screening criteria fromSection 3.5 as a part of this section.ApplicabilityApplicableAlignment StatementAlignsCommentsNoneAlignment BasisThe cables necessary to operate and/or maintain the status of each Safe Shutdown component were evaluated by a detailed review of the drawings. Cables that could impact safeshutdown equipment were identified using the component's control schematic, instrument loop, wiring diagram, if available, or the component's electrical elementary wiring diagrams,one-line drawings, or other available wiring diagrams. These drawings were used as a guide to perform a point to point review of the associated connection diagrams. Cablesassociated with power, control, instrumentation, indication, interlock and any other cable that could impact the component were considered.Fault analysis during cable identification led to the cable fault codes P, L, 0, C, and I as defined in EIR 51-9133191. This made the final compliance analysis bounding. Furtheranalysis determined the effects of a fire induced hot short, open circuit and short to ground during the fire area compliance assessment task. Additional schematic diagrams werereviewed for secondary or interlocked circuits, as necessary, which could impact the operation of components required for safe shutdown.Reference DocumentEIR 51-9133191, NSCA, Sections 2.6, 7.0, and 8.0CNG-FES-017, NFPA 805 Safe Shutdown Equipment Cable SelectionNMPI, April 2013 Page B-53 IINMP1, April 2013Page B-53 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.2 Nuclear Safety Capability Circuit AnalysisNEI 00-01 Ref3.3.1.1.2 Cable Failures AffectingMultiple Safe Shutdown EquipmentNEI 00-01 GuidanceIn cases where the failure (including spurious operations) of a single cable could impact more than one piece of safe shutdownequipment, associate the cable with each piece of safe shutdown equipment.ApplicabilityApplicableAlignment StatementCommentsNoneAligns With IntentAlignment BasisFor cases where single cables have the potential to impact multiple components, the cable would be listed against each component. For control logic circuits, where multiplecomponents receive signals from common control logic, the control logic was analyzed as a primary component and given a pseudo-component identification.A pseudo-component is an artificial association of equipment and cables that performs a common function that is combined into a single entity for analysis purposes only and is not anactual plant hardware designation. The pseudo-component was interlocked to the other associated primary components so that the effect of the control logic could be evaluated on anindividual component level.This methodology was used for similar circuit scenarios such as common power supplies. Whereas this approach does not assign the cable to each individual component, the effect oneach component due to fire damage is analyzed. This method serves to reduce the duplication of cable data when the same cables are assigned to multiple components.Pseudo-components and other primary components, whose associated cabling can affect another primary component based on interposing contacts, were identified on the CableSelection Worksheet of the affected component as an interlocked primary component.Reference DocumentEIR 51-9133191, NSCA, Section 6.0EIR 51-9177678-0OW, Definitions SectionNMP1, April 2013Page B-54 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.2 Nuclear Safety Capability Circuit AnalysisNEI 00-01 Ref3.3.1.1.2.1 Electrical DevicesNEI 00-01 GuidanceElectrical devices such as relays, switches and signal resistor units are considered to be acceptable isolation devices. In the case ofinstrument loops and electrical metering circuits, review the isolation capabilities of the devices in the loop to determine that an acceptableisolation device has been installed at each point where the loop must be isolated so that a fault would not impact the performance of thesafe shutdown instrument function. Refer to Section 3.5 for the types of faults that should be considered when evaluating theacceptability of the isolation device being credited.ApplicabilityApplicableAlignment StatementAlignsCommentsNoneAlignment BasisCables were identified and selected using the component's control schematic, electrical elementary diagrams, one-lines, and/or instrument loop diagrams. These drawings were usedas a guide while performing a point to point review of the associated connection diagrams.Electrical isolation devices prevent malfunctions in one section of a circuit from causing unacceptable effects in other portions of the circuit or other circuits (e.g., open contacts, fuses,switches, instrument isolation modules). Devices credited as providing electrical isolation were identified in the circuit analysis for the affected component.Fault analysis during cable identification led to the P, L, 0, C, and I fault codes. All circuits/cables that are electrically connected to the circuit under the analysis are identified up to acredited isolation device including the instrument loops.Reference DocumentEIR 51-9133191, NSCA, Sections 2.6 and 8.3NMPI, April 2013 Page B-55 IINMP1, April 2013Page B-55 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.2 Nuclear Safety Capability Circuit AnalysisNEI 00-01 Ref3.3.1.1.3 Screening Out Cables WithNo ImpactNEI 00-01 GuidanceScreen out cables for circuits that do not impact the safe shutdown function of a component (i.e., annunciator circuits, space heatercircuits and computer input circuits) unless some reliance on these circuits is necessary. To be properly screened out, however, thecircuits associated with these devices must be isolated from the component's control scheme in such a way that a cable fault would notimpact the performance of the circuit. Refer to Section 3.5 for the types of faults that should be considered when evaluating theacceptability of the isolation device being credited.ApplicabilityApplicableAlignment StatementAlignsCommentsNoneAlignment BasisCables necessary to operate and/or maintain the status of each safe shutdown component were identified and evaluated by a detailed review of the component's control schematic,elementary diagrams, one-lines, instrument loop diagrams or other available wiring diagrams. These drawings were used as a guide while performing a point to point review of theassociated connection diagrams. Cables associated with outputs from auxiliary contacts to computer points, annunciators or motor heaters were excluded from the cable selectionwhen it was concluded that the cable failure would not impact the primary component or performance of the circuit.Panel wires that are completely contained within a panel were not explicitly listed as SSD cables. These wires are inherently included in the analysis in the same manner assecondary components.Reference DocumentEIR 51-9133191. NSCA, Sections 7.0 and 8.3NMPI, April 2013 Page B-56 IINMP1, April 2013Page B-56 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.2 Nuclear Safety Capability Circuit AnalysisNEI 00-01 Ref3.3.1.1.4 Power Supply to SafeShutdown EquipmentNEI 00-01 GuidanceFor each circuit requiring power to perform its safe shutdown function, identify the cable supplying power to each safe shutdown and/orrequired interlock component. Initially, identify only the power cables from the immediate upstream power source for these interlockedcircuits and components (i.e., the closest power supply, load center or motor control center). Review further the electrical distributionsystem to capture the remaining equipment from the electrical power distribution system necessary to support delivery of power fromeither the offsite power source or the emergency diesel generators (i.e., onsite power source) to the safe shutdown equipment. Add thisequipment to the safe shutdown equipment list. The set of cables described above are classified as required safe shutdown cables.Evaluate the power cables for breaker coordination concerns. The non-safe shutdown cables off of the safe shutdown buses areclassified as required for hot shutdown or as important to SSD based on the criteria contained in Appendix H.ApplicabilityApplicableAlignment StatementAlignsCommentsNoneAlignment BasisThe power cables were selected using the component's control schematic or electrical elementary diagrams, one-lines, or instrument loop diagrams or other available wiring diagrams.During the cable selection process, the supporting power sources and interlocks for each primary component were identified. The cascading power supplies (pseudo-componentscreated for power supply interlocks) and the cascading interlocks all serve to identify required power supplies to ensure safe shutdown components are supplied with electrical power.The relationship between the power source and their load components was documented and their dependency was considered during the Fire Area Assessment phase by reviewingthe power source load list report from the database.Breaker coordination calculation 32-9151404-000, Nine Mile Point Unit 1 -NFPA 805 Coordination Study, reviewed the existing and any new electrical circuits that could impact safeshutdown. This calculation identified fire protection program breaker coordination issues concerning proper coordination between the supply breaker/fuse and the load breaker/fusesfor power sources required for hot shutdown. This document meets the nuclear safety capability requirements of NFPA-805 and guidance of NEI 00-01.Reference DocumentEIR 51-9133191, NSCA, Section 8.0EIR 51-9177678-000, Definitions SectionEIR 32-9151404-000, Nine Mile Point Unit I -NFPA 805 Coordination StudyNMPI, April 2013 Page B-57 INMP1, April 2013Page B-57 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.2 Nuclear Safety Capability Circuit AnalysisNEI 00-01 Ref3.3.1.1.4.1 Automatic Initiation LogicsNEI 00-01 GuidanceThe automatic initiation logics for the credited post-fire safe shutdown systems are generally not required to support safe shutdown.Typically, each system can be controlled manually by operator actuation in the main control room or emergency control station. Theemergency control station includes those plant locations where control devices, such as switches, are installed for the purpose ofoperating the equipment. If operator actions to manually manipulate equipment at locations outside the MCR or the emergency controlstation are necessary, those actions must conform to the regulatory requirements on operator manual actions (See Appendix E). If notprotected from the effects of fire, the fire-induced failure of automatic initiation logic circuits should be considered for their potential toadversely affect any post-fire safe shutdown system function.ApplicabilityApplicableAlignment StatementAlignsCommentsNoneAlignment BasisThe Safe Shutdown Analysis takes credit for automatic transfer to an alternate power source if the transfer circuit and power source is not affected by the fire. As an example, the ECsystem can be initiated either manually or automatically. The RPS instruments and logic that automatically initiate the EC system on high reactor pressure or low-low reactor levelhave been included in the analysis. Manual initiation of the EC system can be accomplished from either the Control Room or RSP depending on the fire location. AC power is notrequired to manually initiate DHR via the ECs.Adverse effects due to fire have been considered for the automatic initiation logic circuits. Fire area compliance assessments demonstrate that safe shutdown capability is notadversely affected by a fire in any plant area that disables automatic functions (including initiation logic).Reference DocumentEIR 51-9133191, NSCA, Section 4.0NMPI, April 2013 Page B-58 IINMP1, April 2013Page B-58 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.2 Nuclear Safety Capability Circuit AnalysisNEI 00-01 Ref3.3.1.1.5 Associated CircuitsNEI 00-01 GuidanceCabling for the electrical distribution system is a concern for those breakers that feed circuits and are not fully coordinated with upstreambreakers. With respect to electrical distribution cabling, two types of cable associations exist. For safe shutdown considerations, thedirect power feed to a primary safe shutdown component is associated with the primary component and classified as a required safeshutdown cable. For example, the power feed to a pump is necessary to support the pump. Similarly, the power feed from the loadcenter to an MCC supports the MCC. However, for cases where sufficient branch-circuit coordination is not provided, the same cablesdiscussed above would also support the power supply. For example, the power feed to the pump discussed above would support the busfrom which it is fed because, for the case of a common power source analysis, the concern is the loss of the upstream power source andnot the connected load. Similarly, the cable feeding the MCC from the load center would also be necessary to support the load center.Additionally, the non-safe shutdown circuits off of each of the required safe shutdown components in the electrical distribution system canimpact safe shutdown if not properly coordinated. These cables are classified as required for hot shutdown based on the criteriacontained in Appendix H.ApplicabilityApplicableCommentsNoneAlignment StatementAlignsAlignment BasisThe concern for cabling of the electrical distribution system for primary safe shutdown components involves those breakers that feed associated circuits and may not be fullycoordinated with upstream breakers. This involves circuits that are the direct power feed to primary safe shutdown components and/or the power feed to motor control centers thatsupport components or other motor control centers. The concern for these circuits is not the load itself but the upstream power source.For the NMP1 electrical distribution system, it was assumed that the safe shutdown components and their associated power load were coordinated with their upstream power supplieswhen identifying cables for all existing and any new safe shutdown components.For safe shutdown circuits, the cables included the direct power feed from the load center to the component and any cables associated with that component. In addition, coordinationwas also assumed for any branch circuits related to the safe shutdown component's power source.Associated circuits are those circuits which are not completely independent of the safe shutdown systems or components. Failure or spurious operation of these circuits couldpotentially defeat the safe shutdown capability of a safety system. A fire in a given fire area could potentially affect systems and components thought to be independent of thatparticular fire area.For the purpose of this analysis, an associated circuit must be associated with both a fire area and a safe shutdown system or component. The associated circuits include circuitswhich share a common power supply with safe shutdown component and circuits whose spurious operation would adversely affect the safe shutdown capability.Breaker coordination calculation 32-9151404-000, Nine Mile Point Unit 1 -NFPA 805 Coordination Study, demonstrates the existing coordination status (including any new electricalcircuits) for required common power supplies that could impact safe shutdown. This calculation identifies if there are any fire protection program breaker coordination issuesconcerning proper coordination between the supply breaker/fuse and the load breaker/fuses for power sources required for hot shutdown. This document meets the nuclear safetycapability requirements of NEI 00-01 and NFPA-805.NMPI, April 2013 Page B-59INMP1, April 2013Page B-59 I Constellation Energy Nuclear Group Attachment B -PReference DocumentEIR 51-9133191, NSCA, Section 8.5EIR 32-9151404-000, Nine Mile Point Unit 1 -NFPA 805 Coordination StudyJEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology ReviewINMPI, April 2013 Page B-60 IINMP1, April 2013Page B-60 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.2 Nuclear Safety Capability Circuit AnalysisNEI 00-01 Ref3.3.1.1.6 Exclusion AnalysisNEI 00-01 GuidanceExclusion analysis may be used to demonstrate a lack of potential for any impacts to post-fire safe shutdown from a component or groupof components regardless of the cable routing. For these cases, rigorous cable searching and cable to component associations may notbe required.ApplicabilityNot ApplicableCommentsNoneAlignment StatementNot RequiredAlignment BasisExclusion analysis was not used to demonstrate a lack of potential for any impacts to post-fire safe-shutdown.Reference DocumentNMPI, April 2013 Page B-61 I!NMP1, April 2013Page B-61 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.2 Nuclear Safety Capability Circuit AnalysisNEI 00-01 Ref NEI 00-01 Guidance3.3.2 Associated Circuit Of Concern Appendix R, through the guidance provided in NRC Generic Letter 81-12, requires that separation features be provided for associatedCables non-safety circuits that could prevent operation or cause mal-operation due to hot shorts, open circuits, or shorts to ground, of redundanttrains of systems necessary to achieve hot shutdown. The three types of associated circuits were identified in Reference 7.1.5 and furtherclarified in a NRC memorandum dated March 22, 1982 from R. Mattson to D. Eisenhut, Reference 7.1.6. They are as follows:* Spurious actuations* Common power source* Common enclosure.Each of these cables is classified as an associated circuit of concern cable.Cables Whose Failure May Cause Spurious OperationsSafe shutdown system spurious operation concerns can result from fire damage to a cable whose failure could cause the spuriousoperation/mal-operation of equipment whose operation could affect safe shutdown. These cables are identified in Section 3.3.3 togetherwith the remaining safe shutdown cables required to support control and operation of the equipment.Common Power Source CablesThe concern for the common power source associated circuits of concern is the loss of a safe shutdown power source due to inadequatebreaker/fuse coordination. In the case of a fire-induced cable failure on a non-safe shutdown load circuit supplied from the safe shutdownpower source, a lack of coordination between the upstream supply breaker/fuse feeding the safe shutdown power source and the loadbreaker/fuse supplying the non-safe shutdown faulted circuit can result in loss of the safe shutdown bus. This would result in the loss ofpower to the safe shutdown equipment supplied from that power source preventing the safe shutdown equipment from performing itsrequired safe shutdown function. Identify these cables together with the remaining safe shutdown cables required to support control andoperation of the equipment. Refer to Section 3.5.2.4 for an acceptable methodology for analyzing the impact of these cables on post-firesafe shutdown.Common Enclosure CablesThe concern with common enclosure associated circuits of concern is fire damage to a cable whose failure could propagate to other safeshutdown cables in the same enclosure either because the circuit is not properly protected by an isolation device (breaker/fuse) such thata fire-induced fault could result in ignition along its length, or by the fire propagating along the cable and into an adjacent fire area. Thisfire spread to an adjacent fire area could impact safe shutdown equipment in that fire area, thereby resulting in a condition that exceedsthe criteria and assumptions of this methodology (i.e., multiple fires). Refer to Section 3.5.2.5 for an acceptable methodology foranalyzing the impact of these cables on post-fire safe shutdown.Applicability CommentsApplicable NoneAlignment StatementNot RequiredAlignment BasisThis section is a generic discussion concerning the separation of Associated Circuit Cables and non-safety circuits of components required for safe shutdown. This information isdiscussed in more detail in subsequent sub-sections 3.3.3, 3.5.2.4 and 3.5.2.5.Reference DocumentINMPI, April 2013Page B-62 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.2 Nuclear Safety Capability Circuit AnalysisNEI 00-01 Ref3.3.3 Methodology for CableSelection and LocationNEI 00-01 GuidanceRefer to Figure 3-4 for a flowchart illustrating the various steps involved in selecting the cables necessary for performing a post-fire safeshutdown analysis.Use the following methodology to define the cables required for safe shutdown including cables that may be circuits of concerns for apost-fire safe shutdown analysis. Criteria for making the determination as to which circuits are to be classified as required for hotshutdown or as important to SSD is contained in Appendix H.ApplicabilityApplicableCommentsNoneAlignment StatementNot RequiredAlignment BasisThis paragraph provides introductory information and contains no specific requirements. Discussion is provided in subsequent sub-sections.Reference DocumentNMPI, April 2013 Page B-63 IINMPI, April 2013Page B-63 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.2 Nuclear Safety Capability Circuit AnalysisNEI 00-01 Ref NEI 00-01 Guidance3.3.3.1 Identify Circuits Necessary For each piece of safe shutdown equipment defined in section 3.2, review the appropriate electrical diagrams including the followingfor the Operation of the Safe documentation to identify the circuits (power, control, instrumentation) required for operation or whose failure may impact the operation ofShutdown Equipment each piece of equipment:* Single-line electrical diagrams* Elementary wiring diagram* Electrical connection diagrams* Instrument loop diagrams.For electrical power distribution equipment such as power supplies, identify any circuits whose failure may cause a coordination concernfor the bus under evaluation.If power is required for the equipment, include the closest upstream power distribution source on the safe shutdown equipment list.Through the iterative process described in Figures 3-2 and 3-3, include the additional upstream power sources up to either the offsite orthe emergency power source.Applicability CommentsApplicable NoneAlignment StatementAlignsAlignment BasisCable selection was performed to identify all conductors/wires that may be required for a component to perform its safe shutdown function, or whose failure could be adverse to thecomponent's safe shutdown function. The cables were selected by a point-to-point review of the applicable connection diagrams, single line electrical diagrams, elementary wiringdiagrams or instrument loop wiring diagrams.During the initial review, any cable that had a potential to impact the safe shutdown component was identified and associated to that component. Cables identified for each SafeShutdown component, including any additional reference drawings, were populated in the database during the cable selection process. Cables that are computer inputs or that haveadequate isolation are excluded.Figures 3-2 and 3-3 of NEI 00-01 were used to develop the Safe Shutdown Systems, the systems Paths and Safe Shutdown Equipment List. These lists included electrical distributionequipment identified for circuits whose failure may cause a coordination concern. The power related electrical equipment included upstream power sources up to either offsite power orthe emergency power source.Coordination of power supplies was assumed when assigning cables to the safe shutdown components; however, this may not encompass any new components and circuits beingadded to the program. Breaker coordination calculation 32-9151404-000, Nine Mile Point Unit 1 -NFPA 805 Coordination Study, demonstrates the existing coordination status for therequired common power sources. This calculation identifies if there are any fire protection program breaker coordination issues concerning proper coordination between the supplybreaker/fuse and the load breaker/fuses for power sources required for hot shutdown. This document meets the nuclear safety capability requirements of NEI 00-01, Section 3.5.2.4and Figure 3.5.2-6, and NFPA 805, Section 2.4.2.2.2.Reference DocumentEIR 51-9133191, NSCA, Sections 7.0 and 8.0EIR 32-9151404-000, Nine Mile Point Unit 1 -NFPA 805 Coordination StudyNMP1, April 2013Page B-64 Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.2 Nuclear Safety Capability Circuit AnalysisNEI 00-01 Ref NEI 00-01 Guidance3.3.3.2 Identify Interlocked Circuits In reviewing each control circuit, investigate interlocks that may lead to additional circuit schemes, cables and equipment. Assign to theand Cables that Could Affect Safe equipment any cables for interlocked circuits that can affect the equipment.ShutdownWhile investigating the interlocked circuits, additional equipment or power sources may be discovered. Include these interlockedequipment or power sources in the safe shutdown equipment list (refer to Figure 3-3) if they can impact the operation of the equipmentunder consideration in an undesirable manner that impacts post-fire safe shutdown.Applicability CommentsApplicable NoneAlignment StatementAligns With IntentAlignment BasisFor control logic circuits where multiple components receive signals from a common logic, the control logic was analyzed as a primary component and a pseudo-componentidentification was created for the control logic. A pseudo-component is an artificial association of equipment and cables that performs a common function into a single entity foranalysis purposes. The pseudo-component is not an actual plant hardware designation.The pseudo-component was interlocked to the other associated primary components so that the effect of the control logic could be evaluated on an individual component level. Thismethodology was used for similar circuit scenarios such as common power supplies. Whereas this approach does not assign the cable to each individual component, the effect oneach component due to fire damage is analyzed. This method serves to reduce the duplication of cable data when the same cables are assigned to multiple components.Pseudo-components and other primary components whose associated cabling can affect another primary component based on interposing contacts were identified on the CableSelection Worksheet of the affected component as an interlocked primary component. This meets the intent of the guidance.Reference DocumentEIR 51-9133191, NSCA, Sections 5.4 and 6.0EIR 51-9177678-&4=, Definitions SectionNMPI, April 2013 Page B-65 IINMP1, April 2013Page B-65 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.2 Nuclear Safety Capability Circuit AnalysisNEI 00-01 Ref3.3.3.3 Assign Cables to SafeShutdown EquipmentNEI 00-01 GuidanceGiven the criteria/assumptions defined in Section 3.3.1, identify the cables required to operate or that may result in mal-operation of eachpiece of safe shutdown equipment. Cables are classified as either required for hot shutdown or important to SSD based on theclassification of the component to which they are associated and the function of that component in supporting post-fire safe shutdown ineach particular fire area. Refer to Appendix H for additional guidance.Tabulate the list of cables potentially affecting each piece of equipment in a relational database including the respective drawing numbers,their revision and any interlocks that are investigated to determine their impact on the operation of the equipment. In certain cases, thesame cable may support multiple pieces of equipment. Relate the cables to each piece of equipment, but not necessarily to eachsupporting secondary component.If adequate coordination does not exist for a particular circuit, relate the power cable to the power source. This will ensure that the powersource is identified as affected equipment in the fire areas where the cable may be damaged. Criteria for making the determination as towhich cables are to be classified as required for hot shutdown or as important to SSD is contained in Appendix H.ApplicabilityApplicableAlignment StatementAlignsAlignment BasisCommentsNoneCable selection for Safe Shutdown components was performed by identifying all cables required for a component to perform its safe shutdown function. These cables were selected bya point-to-point review using the component's elementary diagram. The cables were selected in accordance with NEI 00-01, Section 3.3.1 and then entered into the database. Forcases where cables affected multiple components, either the cable was assigned to each component or a pseudo-component was used with the cables assigned to the pseudo-component instead of the primary component.In addition to the cables, any component interlocks were identified to investigate their impact on the operation of the safe shutdown component. The relationship between theseinterlocks and the primary component were documented and their dependency was considered during the Fire Area Compliance Assessment.Coordination of power supplies is addressed in Section 3.5.2.4 of this document.Reference DocumentEIR 51-9133191, NSCA, Section 8.0EIR 51-9177678-=0, Definitions SectionNMPI, April 2013 Page B-66 I!NMP1, April 2013Page B-66 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.2 Nuclear Safety Capability Circuit AnalysisNEI 00-01 Ref3.5 Circuit Analysis and EvaluationNEI 00-01 GuidanceThis section on circuit analysis provides information on the potential impact of fire on circuits used to monitor, control and power requiredfor hot shutdown and important to safe shutdown equipment. Applying the circuit analysis criteria will lead to an understanding of how firedamage to the cables may affect the ability to achieve and maintain post-fire safe shutdown in a particular fire area. This section shouldbe used in conjunction with Section 3.4, to evaluate the potential fire-induced impacts that require mitigation. Additionally, whenassessing fire-induced damage to circuits that could potentially result in MSOs, the circuit failure criteria in Appendix B should be used.Appendix R Section II.G.2 identifies the fire-induced circuit failure types that are to be evaluated for impact from exposure fires on safeshutdown equipment.Section III.G.2 of Appendix R requires consideration of hot shorts, shorts-to-ground and open circuits.ApplicabilityApplicableCommentsNoneAlignment StatementNot RequiredAlignment BasisThis paragraph provides introductory information and contains no specific requirements. Discussion is provided in subsequent sub-sections.Reference DocumentNMP1, April 2013Page B-67 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.2 Nuclear Safety Capability Circuit AnalysisNEI 00-01 Ref NEI 00-01 Guidance3.5.1 Criteria/Assumptions Apply the following criteria/assumptions when performing fire-induced circuit failure evaluations. Refer to the assessment of the NEI/EPRIand CAROLFIRE Cable Test Results in Appendix B to this document for the basis for these criteria and for further elaboration on theapplication of the criteria.Applicability CommentsApplicable NoneAlignment StatementNot RequiredAlignment BasisThis paragraph provides introductory information and contains no specific requirements. Discussion is provided in subsequent sub-sections.Reference DocumentNM P1, April 2013 Page B-68 IINMP1, April 2013Page B-68 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.2 Nuclear Safety Capability Circuit AnalysisNEI 00-01 Ref NEI 00-01 Guidance3.5.1.1 Circuit Failure Criteria Circuit Failure Criteria: The criteria provided below addresses the effects of multiple fire-induced circuit failures impacting circuits forcomponents classified as either "required for hot shutdown" or "important to safe shutdown". Consider the following circuit failure types oneach conductor of each unprotected cable. Criteria differences, however, do apply depending on whether the component is classified asrequired for hot shutdown or important to safe shutdown." A hot short may result from a fire-induced insulation breakdown between conductors of the same cable, a different cable or fromsome other external source resulting in a compatible but undesired impressed voltage or signal on a specific conductor. A hot shortmay cause a spurious operation of safe shutdown equipment.0 A hot short in the control circuitry for an MOV can bypass the MOV protective devices, i.e. torque and limit switches. Thisis the condition described in NRC Information Notice 92-18. In this condition, the potential exists to damage the MOVmotor and/or valve. Damage to the MOV could result in an inability to operate the MOV either remotely, using separatecontrols with separate control power, or manually using the MOV hand wheel. This condition could be a concern in twoinstances: (1) For fires requiring Control Room evacuation and remote operation from the Remote Shutdown Panel, theAuxiliary Control Panel or Auxiliary Shutdown Panel; (2) For fires where the selected means of addressing the effects offire induced damage is the use of an operator manual action. In each case, analysis must be performed to demonstratethat the MOV can be subsequently operated electrically or manually, as required by the safe shutdown analysis.* An open circuit may result from a fire-induced break in a conductor resulting in the loss of circuit continuity. An open circuit mayprevent the ability to control or power the affected equipment. An open circuit may also result in a change of state for normallyenergized equipment. (e.g. [for BWRs] loss of power to the Main Steam Isolation Valve (MSIV) solenoid valves due to an opencircuit will result in the closure of the MSIVs). [Note: Open circuits as a result of conductor melting have not occurred in any of therecent cable fire testing and they are not considered to be a viable form of cable failure.]" A short-to-ground may result from a fire-induced breakdown of a cable insulation system, resulting in the potential on the conductorbeing applied to ground potential. A short-to-ground may have all of the same effects as an open circuit and, in addition, a short-to-ground may also cause an impact to the control circuit or power train of which it is a part. A short-to-ground may also result in achange of state for normally energized equipment.Circuits for "required for hot shutdown" components: Because Appendix R Section III.G.1 requires that the hot shutdown capability remain"free of fire damage", there is no limit on the number of concurrent/simultaneous fire-induced circuit failures that must be considered forcircuits for components "required for hot shutdown: located within the same fire area. For components classified as "required for hotshutdown", there is no limit on the duration of the hot short. It must be assumed to exist until an action is taken to mitigate its effects.Circuits required for the operation of or that can cause the mal-operation of "required for hot shutdown" components that are impacted bya fire are considered to render the component unavailable for performing its hot shutdown function unless these circuits are properlyprotected as described in the next sentence. The required circuits for any "required for hot shutdown" component, if located within thesame fire area where they are credited for achieving hot shutdown, must be protected in accordance with one of the requirements ofAppendix R Section III.G.2 or plant specific license conditions.NMPI, April 2013 Page B-69 IINMP1, April 2013Page B-69 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology ReviewCircuits for "important to safe shutdown" components: Circuits for components classified as "important to safe shutdown" are notspecifically governed by the requirements of Appendix R Section III.G.1, III.G.2 or III.G.3. To address fire-induced impacts on thesecircuits, consider the three types of circuit failures identified above to occur individually on each conductor with the potential to impact any"important to safe shutdown" component with the potential to impact components "required for hot shutdown". In addition, consider thefollowing additional circuit failure criteria for circuits for "important to safe shutdown" components located within the same fire area with thepotential to impact components "required for hot shutdown":" As explained in Figure 3.5.2-3, multiple shorts-to-ground are to be evaluated for their impact on ungrounded circuits.* As explained in Figure 3.5.2-5, for ungrounded DC circuits, a single hot short from the same source is assumed to occur unless itcan be demonstrated that the occurrence of a same source short is not possible in the affected fire area. If this approach is used, ameans to configuration control this condition must be developed and maintained." For the double DC break solenoid circuit design discussed in the NRC Memo from Gary Holahan, Deputy Director Division ofSystems Technology, dated December 4, 1990 and filed under ML062300013, the effect of two hot shorts of the proper polarity inthe same multi-conductor cable should be analyzed for non-high low pressure interface components. [Reference Figure B.3.3 (f) ofNFPA 805-2001.]" Multiple spurious operations resulting from a fire-induced circuit failure affecting a single conductor must be included in the post-firesafe shutdown analysis.* Multiple fire-induced circuit failures affecting multiple conductors within the same multi-conductor cable with the potential to cause aspurious operation of an "important to safe shutdown" component must be assumed to exist concurrently." Multiple fire-induced circuit failures affecting separate conductors in separate cables with the potential to cause a spurious operationof an "important to safe shutdown" component must be assumed to exist concurrently when the effect of the fire-induced circuitfailure is sealed-in or latched.Conversely, multiple fire-induced circuit failures affecting separate conductors in separate cables with the potential to cause aspurious operation of an "important to safe shutdown" component need not be assumed to exist concurrently when the effect of thefire-induced circuit failure is not sealed-in or latched. This criterion applies to consideration of concurrent hot shorts in secondarycircuits and to their effect on a components primary control circuit. It is not to be applied to concurrent single hot shorts in primarycontrol circuit for separate components in an MSO combination.* For components classified as "important to safe shutdown", the duration of a hot short may be limited to 20 minutes. (If the effect ofthe spurious actuation involves a "sealing in" or "latching" mechanism, that is addressed separately from the duration of the spuriousactuation, as discussed above.)* For any impacted circuits for "important to safe shutdown" components that are located within the same fire area, protection inaccordance with the requirements of Appendix R Section Ill.G.2 or plant specific license conditions may be used. In addition,consideration may be given to the use of fire modeling or operator manual actions, as an alternative to the requirements of AppendixR Section III.G.2. (Other resolution options may also be acceptable, if accepted by the Authority Having Jurisdiction.)NMPI, April 2013 Page B-70 IlNMP1, April 2013Page B-70 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology ReviewApplicability CommentsApplicable NoneAlignment StatementAlignsAlignment BasisDuring the cable selection process, a circuit fault analysis for each component cable was initially performed to determine the effects of a fire-induced hot short, open circuit and short toground, as applicable. Per the NEI 00-01 guidance, all combinations of circuit failures (hot shorts, open circuit, and short-to ground) on each conductor for each unprotected safeshutdown cable were considered. Further analysis was performed, as required, for secondary or interlocked circuits.The circuit failures were evaluated to determine the potential impact of a fire on the safe shutdown equipment (including the path) that is associated with that cable/conductor. In somecases, the cables and components had adequate separation from their redundant circuits and components as required by the regulations and were not required to be analyzed.It was assumed that fire damage results in an unusable cable that cannot be considered functional with regard to ensuring proper circuit operation. The insulation and external jacketmaterial of electrical cables are susceptible to fire damage. Damage may assume several forms including deformation, loss of structure, cracking, and ignition. The relationshipbetween exposure of electrical cable insulation to fire conditions, the failure mode, and time to failure may vary with the configuration and cable type. To accommodate theseuncertainties in a consistent and conservative manner, the circuit analysis assumes that the functional integrity of electrical cables is lost when cables are exposed to a fire, exceptwhere protected by a fire rated barrier.The types of circuit failures considered for this analysis are those identified in NEI 00-01, Appendix B, Table B.1.0, "Types of Fire-Induced Circuit Failures Required to Be Considered."Consistent with NEI 00-01, hot shorts were considered to be either internal cable wire-to-wire shorts or external cable-to-cable shorts. No credit was taken for physical cable attributes(armored, thermo-set, etc.) preventing cable-to-cable hot shorts.Reference DocumentEIR 51-9133191, NSCA, Section 8.2HNP RAI 3-16, NRC Request for Addition Information (ML092170715)NMPI, April 2013 Page B-71 I!NMP1, April 2013Page B-71 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.2 Nuclear Safety Capability Circuit AnalysisNEI 00-01 Ref NEI 00-01 Guidance3.5.1.2 Spurious Operation Criteria Spurious Operation Criteria: The following criteria address the effect of multiple spurious operations of components classified as either"required for hot shutdown" or "important to safe shutdown" on post-fire safe shutdown. These criteria are to be applied to the populationof components whose spurious operation has been determined to be possible based on an application of the circuit failure criteriadescribed above when assessing impacts to post-fire safe shutdown capability in any fire area.* The set of concurrent combinations of spurious operations provided through the MSO Process outlined in Section 4 and the list ofMSO contained in Appendix G must be included in the analysis of MSOs.* MSOs do not need to be combined, except as explained in Section 4.4.3.4 of this document.* Section 4.4.3.4 states that the expert panel should review the plant specific list of MSOs to determine whether any of the individualMSOs should be combined due to the combined MSO resulting in a condition significantly worse than either MSO individually.* In this review, consideration of key aspects of the MSOs should be factored in, such as the overall number of spurious operations inthe combined MSOs, the circuit attributes in Appendix B, and other physical attributes of the scenarios.o Specifically, if the combined MSOs involve more than a total of four components or if the MSO scenario requires considerationof sequentially selected cable faults of a prescribed type, at a prescribed time, in a prescribed sequence in order for thepostulated MSO combination to occur, then this is considered to be beyond the required design basis for MSOs.Applicability CommentsApplicable NoneAlignment StatementAlignsAlignment BasisAn MSO Expert panel reviewed the generic list of scenarios listed in NEI-00-01 Appendix G, screening out any scenario that was a combination of five or more spurious operations.There were no screening criteria based on number, timing, or type of circuit failures. The result of the review was a list of MSO scenarios, both generic and site specific, that wereincluded in the Fire Area Analysis for further evaluation.Reference DocumentEIR 51-9133191, NSCA, Section 8.1ONS RAI 3-38, NRC Request for Additional Information dated July 30, 2010 (ML102110394)Technical Report on Identification & Classification of the NMP-1 MSO Scenarios Using an Expert Panel -Review of New Scenarios, Rev.1NMPI, April 2013 Page B-72 IINMPI, April 2013Page B-72 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.2 Nuclear Safety Capability Circuit AnalysisNEI 00-01 Ref3.5.1.3 Circuit Contact PositionNEI 00-01 GuidanceAssume that circuit contacts are initially positioned (i.e., open or closed) consistent with the normal mode/position of the "required for hotshutdown" or "important to safe shutdown" equipment as shown on the schematic drawings. The analyst must consider the position of the"required for hot shutdown" and "important to safe shutdown" equipment for each specific shutdown scenario when determining the impactthat fire damage to a particular circuit may have on the operation of the "required for hot shutdown" and "important to safe shutdownequipment".ApplicabilityApplicableCommentsNoneAlignment StatementAlignsAlignment BasisThe analysis assumes that the circuit contacts are positioned (i.e., open or closed) consistent in the normal mode/position of the safe shutdown equipment as shown on the schematicdrawings or defined by procedure. The fire damage impact on the position of the safe shutdown equipment was considered for each shutdown scenario.Reference DocumentEIR 51-9133191, NSCA, Section 8.6.1NMPI, April 2013 Page B-73INMP1, April 2013Page B-73 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.2 Nuclear Safety Capability Circuit AnalysisNEI 00-01 Ref3.5.2 Types Of Circuit FailuresNEI 00-01 GuidanceAppendix R requires that nuclear power plants must be designed to prevent exposure fires from defeating the ability to achieve andmaintain post-fire safe shutdown. Fire damage to circuits that provide control and power to equipment required for hot shutdown andimportant to safe shutdown in each fire area must be evaluated for the effects of a fire in that fire area. Only one fire at a time is assumedto occur. The extent of fire damage is assumed to be limited by the boundaries of the fire area. Given this set of conditions, it must beassured that one redundant train of equipment necessary to achieve and maintain hot shutdown is free of fire damage for fires in everyplant location. To provide this assurance, Appendix R requires that equipment and circuits required for hot shutdown be free of firedamage and that these circuits be designed for the fire-induced effects of a hot short, short-to-ground, or an open circuit. With respect tothe electrical distribution system, the issue of breaker coordination must also be addressed. Criteria for making the determination as towhich breakers are to be classified as required for hot shutdown is contained in Appendix H.This section will discuss specific examples of each of the following types of circuit failures:* Open circuit* Short-to-ground* Hot shortAlso, refer to Appendix B for the circuit failure criteria to be applied in assessing the impact of the Plant Specific List of MSOs on post-firesafe shutdown.ApplicabilityApplicableCommentsNoneAlignment StatementNot RequiredAlignment BasisThis paragraph provides introductory information and contains no specific requirements. Discussion is provided in subsequent sub-sections.Reference DocumentNMPI, April 2013 Page B-74 IINMPI, April 2013Page B-74 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.2 Nuclear Safety Capability Circuit AnalysisNEI 00-01 Ref NEI 00-01 Guidance3.5.2.1 Circuit Failures Due to an This section provides guidance for addressing the effects of an open circuit for required for hot shutdown and important to safe shutdownOpen Circuit equipment. An open circuit is a fire-induced break in a conductor resulting in the loss of circuit continuity. An open circuit will typicallyprevent the ability to control or power the affected equipment. An open circuit can also result in a change of state for normally energizedequipment. For example, a loss of power to the main steam isolation valve (MSIV) solenoid valves [for BWRs] due to an open circuit willresult in the closure of the MSIV.* Loss of electrical continuity may occur within a conductor resulting in de-energizing the circuit and causing a loss of power to, orcontrol of, the required for hot shutdown and important to safe shutdown equipment.* In selected cases, a loss of electrical continuity may result in loss of power to an interlocked relay or other device. This loss ofpower may change the state of the equipment. Evaluate this to determine if equipment fails safe.* Open circuit on a high voltage (e.g., 4.16 kV) ammeter current transformer (CT) circuit may result in secondary damage,possibly resulting in the occurrence of an additional fire in the location of the CT itself.Applicability CommentsApplicable NoneAlignment StatementAlignsAlignment BasisOpen circuits are analyzed as referenced in the NEI 00-01 guidance, Section 3.5.2.1 and Figure 3.5.2-1. An open circuit is a condition experienced when an individual conductor withina cable loses electrical continuity due to a fire induced break. This could cause the loss of power from de-energizing the circuit or the ability to control affected components, or onenergized equipment could cause a change of position of the component. In addition, as stated in the guidance, an open circuit on a high voltage ammeter CT circuit may result insecondary damage to that circuit.The Nuclear Safety Capability Assessment (NSCA) assumed that fire damage results in an unusable cable that cannot be considered functional with regard to ensuring proper circuitoperation. The insulation and external jacket material of electrical cables are susceptible to fire damage. Damage may assume several forms including deformation, loss of structure,cracking, and ignition. The relationship between exposure of electrical cable insulation to fire conditions, the failure mode, and time to failure may vary with the configuration and cabletype. To accommodate these uncertainties in a consistent and conservative manner, the circuit analysis assumes that the functional integrity of electrical cables was lost when cablesare exposed to a fire, except where protected by a fire rated barrier.Other associated circuit concerns are related to open secondary circuits and 4 KV bus CT that may result in high currents producing an additional fire at the transformer location. Thisassociated circuit concern is related to CT where an open secondary circuit may develop high voltages within a transformer potentially resulting in a secondary fire at the transformerlocation. This issue is evaluated in Fire Protection Engineering Evaluation FPEE-1-04-002, Rev. 0, Fire Effects on CTs and Instrument Sensing Lines and The Plant Safe ShutdownCapability. The evaluation concludes that, for all CTs in use at NMP1, an open transformer secondary will not develop voltages that are high enough to threaten the Safe Shutdowncapability.Reference DocumentEIR 51-9133191, NSCA, Sections 8.2 and 8.5Fire Protection Engineering Evaluation FPEE-1 -04-002, Rev. 0, Fire Effects on CTs ONS RAI 3-48, NRC Request for Additional Information dated July 30, 2010 (ML102110394)HNP RAI 3-17, NRC Request for Additional Information dated August 6, 2009 (ML092170715)INMP1, April 2013Page B-75 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology ReviewConstellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.2 Nuclear Safety Capability Circuit AnalysisNEI 00-01 Ref3.5.2.2 Circuit Failures Due to aShort-to-GroundNEI 00-01 GuidanceThis section provides guidance for addressing the effects of a short-to-ground on circuits for required for hot shutdown and important tosafe shutdown equipment. A short-to-ground is a fire-induced breakdown of a cable insulation system resulting in the potential on theconductor being applied to ground potential. A short-to-ground can cause a loss of power to or control of required safe shutdownequipment. In addition, a short-to-ground may affect other equipment in the electrical power distribution system in the cases where propercoordination does not exist.There is no limit to the number of shorts-to-ground that could be caused by the fire.Consider the following consequences in the post-fire safe shutdown analysis when determining the effects of circuit failures related toshorts-to-ground:* A short to ground in a power or a control circuit may result in tripping one or more isolation devices (i.e. breaker/fuse) andcausing a loss of power to or control of required safe shutdown equipment.* In the case of certain energized equipment such as HVAC dampers, a loss of control power may result in loss of power to aninterlocked relay or other device that may cause one or more spurious operations.CommentsNoneApplicabilityApplicableAlignment StatementAlignsAlignment BasisThe methodology assumes multiple fire induced failures including short-to-ground. The short-to-ground issue incorporates circuit failures for both ungrounded and grounded circuits.Postulated cable and component failures were identified utilizing the techniques referenced in NEI 00-01 Figure 3.5.2-2 and Figure 3.5.2-3. The safe shutdown analysis may excludecertain cables if their postulated fire induced faults have no adverse effect on the component.A short-to-ground fault for grounded circuits could cause the tripping of a circuit thereby causing a loss of power to the control circuit. For certain cases of energized components, aloss of control power may result in a loss of power to relays and other devices interlocked with the device.Unless otherwise justified by circuit analysis, short- to-ground for ungrounded circuits are treated the same as short- to-ground for grounded circuits, and are postulated to result in aloss of motive power or control power. This is consistent with NFPA 805 Appendix B, Section B.3.2.g which states: "For ease of analysis when analyzing an ungrounded DC circuit forthe effects of a short-to-ground, it should be assumed that an existing ground fault from the same power source is present."Reference DocumentEIR 51-9133191, NSCA, Sections 8.1 and 8.2NFPA 805, Appendix B, Section B.3.2.gNMP1, April 2013Page B-76 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.2 Nuclear Safety Capability Circuit AnalysisNEI 00-01 Ref NEI 00-01 Guidance3.5.2.3 Circuit Failures Due to a Hot This section provides guidance for analyzing the effects of a hot short on circuits for required for required for hot shutdown and importantShort to safe shutdown equipment. A hot short is defined as a fire-induced insulation breakdown between conductors of the same cable, adifferent cable or some other external source resulting in an undesired impressed voltage on a specific conductor. The potential effect ofthe undesired impressed voltage would be to cause equipment to operate or fail to operate in an undesired manner.Consider the following specific circuit failures related to hot shorts as part of the post-fire safe shutdown analysis:" A hot short between an energized conductor and a de-energized conductor within the same cable may cause a spuriousoperation of equipment. The spuriously operated device (e.g., relay) may be interlocked with another circuit that causes thespurious operation of other equipment. This type of hot short is called an intra-cable hot short (also known as conductor-to-conductor hot short or an internal hot short)." A hot short between any external energized source such as an energized conductor from another cable and a de-energizedconductor may also cause a spurious operation of equipment. This is called an inter-cable hot short (also known as cable-to-cable hot short/external hot short).* A hot short in the control circuitry for an MOV can bypass the MOV protective devices, i.e. torque and limit switches. This is thecondition described in NRC Information Notice 92-18. In this condition, MOV motor damage can occur. Damage to the MOVmotor could result in an inability to operate the MOV either remotely, using separate controls with separate control power, ormanually using the MOV hand wheel. This condition could be a concern in two instances: (1) For fires requiring Control Roomevacuation and remote operation from the Remote Shutdown Panel; (2) For fires where the selected means of addressing theeffects of fire induced damage is the use of an operator manual action. In this latter case, analysis must be performed todemonstrate that the MOV thrust at motor failure does not exceed the capacity of the MOV hand wheel. For either case,analysis must demonstrate the MOV thrust at motor failure does not damage the MOV pressure boundary.Applicability CommentsApplicable NoneAlignment StatementAlignsAlignment BasisA hot short is a condition experienced when an energized individual conductor of the same or different cable comes into contact with another conductor of the same or different cableresulting in electrical continuity between the conductors. The potential effect is that the energized conductor becomes an undesired source of power for the circuit being analyzed. Hotshorts were considered to be either internal conductors of the same cable, identified as internal shorts, or shorts between conductors of different cables, identified as external shorts.The potential of circuit failures due to hot shorts can cause components to operate or cause them to fail to operate in an undesired manner.The Nuclear Safety Capability Assessment (NSCA) assumed that fire damage results in an unusable cable that cannot be considered functional with regard to ensuring proper circuitoperation. The insulation and external jacket material of electrical cables are susceptible to fire damage. Damage may assume several forms including deformation, loss of structure,cracking, and ignition. The relationship between exposure of electrical cable insulation to fire conditions, the failure mode, and time to failure may vary with the configuration and cabletype. To accommodate these uncertainties in a consistent and conservative manner, the circuit analysis assumes that the functional integrity of electrical cables is lost when cablesare exposed to a fire, except where protected by a fire rated barrier.NMPI, April 2013 Page B-77 IINMP1, April 2013Page B-77 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology ReviewConsistent with NEI 00-01, hot shorts were considered to be either internal cable wire-to-wire shorts or cable-to-cable (external) shorts. No credit was taken for physical cableattributes (armored, thermo-set, etc.) preventing cable-to-cable hot shorts.For cable failures due to hot shorts on grounded or ungrounded circuits, the methodology initially assumes the hot short would have sufficient potential to cause a spurious operation ofthe component. Two types of cable hot short conditions are considered to be of sufficiently low likelihood that they are not assumed credible, except for analysis involving high/lowpressure interface components. These hot shorts are 3-phase AC power circuit cable-to-cable proper phase sequence faults and 2-wire ungrounded DC circuit cable-to-cable properpolarity faults.Instrument circuits that operate at low signal levels (4-20 mA, 0-1 V, 1-5 V, etc.) and are enclosed in a grounded metal shield are not considered to be susceptible to hot shorts fromother adjacent instrument circuits external to the shield. External circuits are assumed to short to ground via the shield and do not have the potential of creating a signal of properpolarity and amplitude to simulate a valid instrument signal.Reference DocumentEIR 51-9133191, NSCA, Section 8.2ONS RAI 3-41, NRC Request for Additional Information dated July 30, 2010 (ML102110394)NMPI, April 2013 Page B-78INMP1, April 2013Page B-78 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.3 Nuclear Safety Equipment and Cable LocationPhysical location of equipment and cables shall be identified.NEI 00-01 Ref3.3.3.4 Identify Routing of CablesNEI 00-01 GuidanceIdentify the routing for each cable including all raceway and cable endpoints. Typically, this information is obtained from joining the list ofsafe shutdown cables with an existing cable and raceway database.ApplicabilityApplicableCommentsNoneAlignment StatementAlignsAlignment BasisThis task involved identifying the routing and location for all raceways and endpoints for cables associated with safe shutdown equipment. The cable routes and their endpoint locationwere populated into the database. The database is a relational database that contains all the required information for safe shutdown cable routing and endpoint information. Theoriginal cable routing and cable endpoint data was provided from the NMP1 cable raceway database (TRAK2000).CommentsNoneReference DocumentEIR 51-9133191, NSCA, Sections 2.1 and 8.4TRAK2000, Revision 6.01NMPI, April 2013 Page B-79 IINMPI, April 2013Page B-79 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.3 Nuclear Safety Equipment and Cable LocationNEI 00-01 Ref3.3.3.5 Identify Raceway and Cablesby Fire AreaNEI 00-01 GuidanceIdentify the fire area location of each raceway and cable endpoint identified in the previous step and join this information with the cablerouting data. For raceway and cable endpoints in multiple fire areas, each fire area where the raceway or cable endpoint exists must beincluded. In addition, identify the location of field-routed cable by fire area. This produces a database containing all of the cablesrequiring fire area analysis, their locations by fire area, and their raceway.ApplicabilityApplicableCommentsNoneAlignment StatementAlignsAlignment BasisFire area locations were identified for each cable raceway and cable endpoint by obtaining the location coordinates from applicable cable tray, conduit or equipment layoutarrangement drawings or by field walkdown, if necessary. The fire area locations were identified by comparing the cable tray/conduit arrangement drawings, equipment arrangementdrawings, or field walkdown data with Fire Area Floor Plans drawings. This correlation between cable raceway locations and Fire Areas and Rooms was populated into the databaseto produce computer generated reports. The reports contained the cable related raceway information required to prepare the Fire Area Analysis.Reference DocumentEIR 51-9133191, NSCA, Sections 5.4 and 8.4Fire Area Floor Plans Drawings B40141C through B40148CNMPI, April 2013 Page B-80 I!NMP1, April 2013Page B-80 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.3 Nuclear Safety Equipment and Cable LocationNEI 00-01 Ref NEI 00-01 Guidance3.5.2.4 Circuit Failures Due to The evaluation of circuits of a common power source consists of verifying proper coordination between the supply breaker/fuse and theInadequate Circuit Coordination load breakers/fuses for power sources that are required for hot shutdown. The concern is that, for fire damage to a single power cable,lack of coordination between the supply breaker/fuse and the load breakers/fuses can result in the loss of power to a safe shutdownpower source that is required to provide power to safe shutdown equipment.A coordination study should demonstrate the coordination status for each required common power source. For coordination to exist, thetime-current curves for the breakers, fuses and/or protective relaying must demonstrate that a fault on the load circuits is isolated beforetripping the upstream breaker that supplies the bus. Furthermore, the available short circuit current on the load circuit must be consideredto ensure that coordination is demonstrated at the maximum fault level.The methodology for identifying potential circuits of a common power source and evaluating circuit coordination cases on a single circuitfault basis is as follows:* Identify the power sources required to supply power to safe shutdown equipment.* For each power source, identify the breaker/fuse ratings, types, trip settings and coordination characteristics for the incomingsource breaker supplying the bus and the breakers/fuses feeding the loads supplied by the bus.* For each power source, demonstrate proper circuit coordination using acceptable industry methods. For example, for breakersthat have internal breaker tripping devices and do not require control power to trip the breaker, assure that the time-currentcharacteristic curve for any affected load breaker is to the left of the time-current characteristic curve for the bus feeder breakerand that the available short circuit current for each affected breaker is to the right of the time-current characteristic curve for thebus feeder breaker or that the bus feeder breaker has a longer time delay in the breaker instantaneous range than the loadbreaker. For breakers requiring control power for the breaker to trip, the availability of the required control power must bedemonstrated in addition to the proper alignment of the time-current characteristic curves described above. The requirement forthe availability of control power would apply to load breakers fed from each safe shutdown bus where a fire-induced circuitfailure brings into questions the availability of coordination for a required for hot shutdown component.* For power sources not properly coordinated, tabulate by fire area the routing of cables whose breaker/fuse is not properlycoordinated with the supply breaker/fuse. Evaluate the potential for disabling power to the bus in each of the fire areas in whichthe circuit of concern are routed and the power source is required for hot shutdown. Prepare a list of the following informationfor each fire area:* Cables of concern.* Affected common power source and its path.* Raceway in which the cable is enclosed.* Sequence of the raceway in the cable route.* Fire zone/area in which the raceway is located.For fire zones/areas in which the power source is disabled, the effects are mitigated by appropriate methods.* Develop analyzed safe shutdown circuit dispositions for the circuit of concern cables routed in an area of the same path asrequired by the power source. Evaluate adequate separation and other mitigation measures based upon the criteria inAppendix R, NRC staff guidance, and plant licensing bases.NMPI, April 2013 Page B-81INMP1, April 2013Page B-81 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology ReviewApplicability CommentsApplicable NoneAlignment StatementAlignsAlignment BasisThe NMP1 Fire Area Assessments are performed to support transition to a performance based fire protection licensing basis. While performing the primary component circuit analysisfor safe shutdown components, it was assumed that electrical coordination exists for all power supplies for each level of electrical power.Breaker coordination calculation 32-9151404-000, Nine Mile Point Unit 1 -NFPA 805 Coordination Study, demonstrates the existing coordination status for the required commonpower sources. This calculation identifies any fire protection program breaker coordination issues concerning proper coordination between the supply breaker/fuse and the loadbreaker/fuses for power sources required for hot shutdown. This document meets the nuclear safety capability requirements of NEI 00-01, Section 3.5.2.4, Figure 3.5.2-6 and NFPA805, Section 2.4.2.2.2. Any identified issues have been addressed in the NSCA.Other potential coordination concerns involve associated non-safe shutdown circuits that are not independent of safe shutdown circuits that could also potentially defeat the functionsof safe shutdown circuits if not properly protected. These circuits must be associated with both a fire area and a safe shutdown system or component to warrant consideration. Theseassociated circuits are divided into three categories.* Circuits that share a common power supply with safe shutdown circuits.* Circuits that share a common enclosure with safe shutdown circuits.* Circuits for components the spurious operation of which would adversely affect the shutdown process.The associated circuits are defined in the NMP1 Coordination study which reviews the 4.16 kV, 600 VAC, 480 VAC, 208/120 VAC and 125 VDC power supplies credited for post-fireshutdown.Proper circuit coordination for power supplies was reviewed, analyzed and addressed in EIR 51-9133191, NSCA.Reference DocumentEIR 51-9133191, NSCA, Section 8.5HNP RAI 3-18 and RAI 3-19, NRC Request for Additional Information dated August 6, 2009 (ML092170715)EIR 32-9151404-000, Nine Mile Point Unit 1 -NFPA 805 Coordination StudyNMPI, April 2013 Page B-82 IINMP1, April 2013Page B-82 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.3 Nuclear Safety Equipment and Cable LocationNEI 00-01 Ref3.5.2.5 Circuit Failures Due toCommon Enclosure ConcernsNEI 00-01 GuidanceThe common enclosure concern deals with the possibility of causing secondary failures due to fire damage to a circuit either whoseisolation device fails to isolate the cable fault or protect the faulted cable from reaching its ignition temperature, or the fire somehowpropagates along the cable into adjoining fire areas.The electrical circuit design for most plants provides proper circuit protection in the form of circuit breakers, fuses and other devices thatare designed to isolate cable faults before ignition temperature is reached. Adequate electrical circuit protection and cable sizing areincluded as part of the original plant electrical design maintained as part of the design change process. Proper protection can be verifiedby review of as-built drawings and change documentation. Review the fire rated barrier and penetration designs that preclude thepropagation of fire from one fire area to the next to demonstrate that adequate measures are in place to alleviate fire propagationconcerns.ApplicabilityApplicableCommentsNoneAlignment StatementAlignsAlignment BasisCircuit failures due to common enclosure concerns are addressed by breaker coordination calculation 32-9151404-000, Nine Mile Point Unit 1 -NFPA 805 Coordination Study. Thiscalculation demonstrates the existing coordination status for electrical circuits that could impact safe shutdown concerning proper coordination between the supply breaker/fuse andthe load breaker/fuses for power sources required for hot shutdown. This document meets the nuclear safety capability requirements of NEI 00-01, Section 3.5.2.4, Figure 3.5.2-6 andNFPA 805, Section 2.4.2.2.2.Reference DocumentEIR 51-9133191, NSCA, Section 8.5HNP RAI 3-18 and RAI 3-19, NRC Request for Additional Information dated August 6, 2009 (ML092170715)EIR 32-9151404-000, Nine Mile Point Unit 1 -NFPA 805 Coordination StudyNMPI, April 2013 Page B-83 IlNMP1, April 2013Page B-83 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.4 Fire Area AssessmentFire Area Assessment. An engineering analysis shall be performed in accordance with the requirements of Section 2.3 for each fire area to determine the effects of fire or firesuppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic).NEI 00-01 Ref3.4 Fire Area Assessment andCompliance StrategiesNEI 00-01 GuidanceBy determining the location of each component and cable by fire area and using the cable to equipment relationships described above,the affected safe shutdown equipment in each fire area can be determined. Using the list of affected equipment in each fire area, theimpacts to safe shutdown systems, paths and functions can be determined. Based on an assessment of the number and types of theseimpacts, the required safe shutdown path for each fire area can be determined. The specific impacts to the selected safe shutdown pathcan be evaluated using the circuit analysis and evaluation criteria contained in Section 3.5 of this document. Knowing which componentsand systems are performing which safe shutdown functions, the required and important to SSD components can be classified. Oncethese component classifications have been made the tools available for mitigating the effects of fire induced damage can be selected.Refer to Appendix H for additional guidance on classifying components as either required for hot shutdown or important to safe shutdown.For MSOs the Resolution Methodology outlined in Section 4, Section 5, Appendix B and Appendix G should be applied. Components ineach MSO are classified as either required for hot shutdown or important to safe shutdown components using the criteria from AppendixH. Similarly, this classification determines the available tools for mitigating the effects of fire-induced damage to the circuits for thesecomponents.Having identified all impacts to the required safe shutdown path in a particular fire area, this section provides guidance on the techniquesavailable for individually mitigating the effects of each of the potential impacts.ApplicabilityApplicableCommentsNoneAlignment StatementNot RequiredAlignment BasisThis paragraph provides introductory information and contains no specific requirements. Discussion is provided in subsequent sub-sections.Reference DocumentNMPI, April 2013 Page B-84 IINMP1, April 2013Page B-84 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.4 Fire Area AssessmentNEI 00-01 Ref3.4.1 Criteria/AssumptionsNEI 00-01 GuidanceThe following criteria and assumptions apply when performing "deterministic" fire area compliance assessment to mitigate theconsequences of the circuit failures identified in the previous sections for the required safe shutdown path in each fire area.ApplicabilityApplicableCommentsNoneAlignment StatementNot RequiredAlignment BasisThis paragraph provides introductory information and contains no specific requirements. Discussion is provided in subsequent sub-sections.Reference DocumentNMPI, April 2013 Page B-85 IINMP1, April 2013Page B-85 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.4 Fire Area AssessmentNEI 00-01 Ref3.4.1.1 Assume a Single FireNEI 00-01 GuidanceAssume only one fire in any single fire area at a time.Applicability CommentsApplicable NoneAlignment StatementAlignsAlignment BasisOnly one fire is assumed to occur in any single fire area at a time.Reference DocumentEIR 51-9133191, NSCA, Section 9.0NMPI, April 2013 Page B-86 IINMP1, April 2013Page B-86 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.4 Fire Area AssessmentNEI 00-01 Ref3.4.1.2 Fire Affects All UnprotectedCables and EquipmentNEI 00-01 GuidanceAssume that the fire may affect all unprotected cables and equipment within the fire area. This assumes that neither the fire size nor thefire intensity is known. This is conservative and bounds the exposure fire that is postulated in the regulation.ApplicabilityApplicableCommentsNoneAlignment StatementAlignsAlignment BasisFor a conservative approach which bounds the exposure fire required by the regulations, the analysis assumes a fully involved fire and that all equipment and unprotected cablingwithin a given fire area are damaged by the fire.Reference DocumentEIR 51-9133191, NSCA, Section 9.0NMPI, April 2013 Page B-87 I!NMP1, April 2013Page B-87 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.4 Fire Area AssessmentNEI 00-01 Ref3.4.1.3 Address all Cable andEquipment Impacts Affecting theRequired Safe Shutdown PathNEI 00-01 GuidanceAddress all cable and equipment impacts affecting the required safe shutdown path in the fire area. All potential impacts within the firearea must be addressed. The focus of this section is to determine and assess the potential impacts to the required safe shutdown pathselected for achieving post-fire safe shutdown and to assure that the required safe shutdown path for a given fire area is properlyprotected.ApplicabilityApplicableCommentsNoneAlignment StatementAlignsAlignment BasisFire Area Assessments were performed on a Fire Area basis in order to ensure compliance in accordance with the safe shutdown requirements of NFPA 805. The Safe ShutdownSystem and Component drawings were analyzed for each Fire Area to ensure that a success path is available based upon the postulated equipment and/or cable losses in the area.The potentially affected equipment and cables in each Fire Area were reviewed and impacts on safe shutdown success paths analyzed.The route and location of all safe shutdown cables were loaded into the safe shutdown database. This data was used to generate Fire Area Component Impact Reports, whichidentified affected systems and components on a Fire Area basis. The Fire Area Component Impact Reports were used as a means to determine the least impacted safe shutdownpath for each fire area.The Fire Area Analysis methodology assumed multiple fire-induced failures and multiple spurious actuations, based on the cables and components present in the fire area of concern.All postulated cable and component failures were identified and a resolution provided at the component level.The least impacted safe shutdown success path was analyzed so that mitigating strategies could be developed and documented in the Fire Area Assessment. A success pathdetermination for all safe shutdown functions was performed. Generally the path with the least amount of failures was recovered to demonstrate a success path for safe shutdown.Support systems were reviewed in order to assess the impact on the systems being supported. The credited safe shutdown success path was documented in the fire area complianceassessment.Reference DocumentEIR 51-9133191, NSCA, Section 9.0NMPI, April 2013 Page B-88 IINMP1, April 2013Page B-88 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.4 Fire Area AssessmentNEI 00-01 Ref3.4.1.4 Classify EachCable/ComponentNEI 00-01 GuidanceUse the criteria from Appendix H to classify each impacted cable/component as either a required or important to SSD cable/component.ApplicabilityApplicableCommentsNoneAlignment StatementNot RequiredAlignment BasisUsing the criteria from Appendix H to classify each impacted cable/component as either a required or important to SSD cable/component was not required to support the transition toNFPA 805. Therefore, cables and components were not classified as "required for safe shutdown" or "important for safe shutdown." The safe shutdown flow paths identify the primarycomponents that are required to meet the safe shutdown performance goals. The safe shutdown cables/components were compiled based on each system's performance and safeshutdown function. These components establish the primary safe shutdown flowpath for system operation. Also included in the safe shutdown flow paths are thosecables/components whose spurious operation could impact safe shutdown system operability. Systems, components, and cables identified as necessary for the operation of the safeshutdown system under review are included in the safe shutdown equipment and cables lists and are designated with the same shutdown path as the primary safe shutdown system.The components may involve branch flow paths that must be isolated and remain isolated to assure that flow will not be diverted from the primary flow path. The list of primarycomponents may also include selected mechanical components required to support safe shutdown.Reference DocumentEIR 51-9133191, NSCA, Section 9.0NMPI, April 2013 Page B-89 I!NMP1, April 2013Page B-89 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.4 Fire Area AssessmentNEI 00-01 Ref3.4.1.5 Manual ActionsNEI 00-01 GuidanceUse operator manual actions where appropriate, for cable/component impacts classified as important to SSD cable/components, toachieve and maintain post-fire safe shutdown conditions in accordance with NRC requirements (refer to Appendix E). For additionalcriteria to be used when determining whether an operator manual action may be used for a flow diversion off of the primary flow path,refer to Appendix H.ApplicabilityApplicableAlignment StatementAlignsCommentsNoneAlignment BasisManual actions performed as prescribed in procedures or otherwise are documented in each Fire Area Assessment (FAA). Included in each FAA was the identification of any requiredoperator actions outside the Main Control Room. The operator actions are those directed by operating procedures, repair procedures, or otherwise identified as necessary during thecourse of the individual FAA. Actions performed at locations other than primary control stations are identified as recovery actions requiring further review as part of the fire riskevaluations. These actions are identified in the Report of Manual Actions and Report of Procedure Directed Manual Actions included in each FAA.Reference DocumentEIR 51-9133191, NSCA, Section 9.0NMPI, April 2013 Page B-90 IINMP1, April 2013Page B-90 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.4 Fire Area AssessmentNEI 00-01 Ref3.4.1.6 RepairsNEI 00-01 GuidanceWhere appropriate to achieve and maintain cold shutdown within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , use repairs to equipment required in support of post-fireshutdown.ApplicabilityApplicableAlignment StatementCommentsNoneAligns with IntentAlignment BasisRepairs whito are relied upon t achieve and maintain cold shutdown will be performed as Fequired. Repairs are directed by plant procedures. However, NFPA 805 requires only thatthe plant be maintained in a safe and stable condition. Nor does NFPA 805 require that the plant achieve cold shutdown in 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Refer to Section 4.2.1.2 for a description of safeand stable as applied to NMP1. NMPI demonstrates the ability to maintain for each fire area the fuel in a safe and stable condition with one of four designated shutdown paths.Reference DocumentEIR 51-9133191, NSCA, Sections 2.1, 8.5, and 9.0Closure of National Fire Protection Association 805 Transition Program Frequently Asked Question Number 08-0054 (MLI 10140183)NMPI, April 2013 Page B-91 I!NMP1, April 2013Page B-91 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.4 Fire Area AssessmentNEI 00-01 Ref NEI 00-01 Guidance3.4.1.7 Appendix R Compliance For the components on the required safe shutdown path classified as required hot shutdown components as defined in Appendix H,Criteria Appendix R compliance requires that one train of systems necessary to achieve and maintain hot shutdown conditions from either thecontrol room or emergency control station(s) is free of fire damage (III.G.1 .a). When cables or equipment are within the same fire areaoutside primary containment and separation does not already exist, provide one of the following means of separation for the required safeshutdown components impacted circuit(s):" Separation of cables and equipment and associated non-safety circuits of redundant trains within the same fire area by a firebarrier having a 3-hour rating (lll.G.2.a)" Separation of cables and equipment and associated non-safety circuits of redundant trains within the same fire area by ahorizontal distance of more than 20 feet with no intervening combustibles or fire hazards. In addition, fire detectors and anautomatic fire suppression system shall be installed in the fire area (lll.G.2.b)." Enclosure of cable and equipment and associated non-safety circuits of one redundant train within a fire area in a fire barrierhaving a one-hour rating. In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area(lll.G.2.c).For fire areas inside non-inerted containments, the following additional options are also available:* Separation of cables and equipment and associated non-safety circuits of redundant trains by a horizontal distance of morethan 20 feet with no intervening combustibles or fire hazards (llI.G.2.d);* Installation of fire detectors and an automatic fire suppression system in the fire area (llI.G.2.e);* Separation of cables and equipment and associated non-safety circuits of redundant trains by a noncombustible radiant energyshield (llI.G.2.f).Use exemptions, deviations, LARs and licensing change processes to satisfy the requirements mentioned above and to demonstrateequivalency depending upon the plant's license requirements.Applicability CommentsApplicable NoneAlignment StatementAligns With IntentAlignment BasisEach NMP1 fire area containing safe shutdown equipment or cables was reviewed in a deterministic fashion for the ability to achieve post-fire safe shutdown. The affected shutdownrelated cables and components in each area were identified and the resultant information used to determine the preferred shutdown path to achieve safe shutdown.The credited safe shutdown success paths were analyzed and mitigating strategies (procedu"al ...i.n., .epair actios.. Or ..were developed and documented in the SafeShutdown Analysis, fire area compliance assessments. The results of the assessments confirm that in the event of a postulated exposure fire, the safe shutdown capability of NMP1will be maintained such that the fuel remains in a safe and stable condition..NMPI, April 2013 Page B-92 IINMP1, April 2013Page B-92 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology ReviewThe non-safe shutdown circuits which are not completely independent of safe shutdown circuits are associated circuits. These associated circuits are those circuits that couldadversely affect the safe shutdown capability or components. These circuits would be associated with a safe shutdown system or component and analyzed the same as other safeshutdown circuits affected within that fire area.Reference DocumentEIR 51-9133191, NSCA, Sections 5.1, 5.2, and 5.3NMPI, April 2013 Page B-93 IINMPI, April 2013Page B-93 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.4 Fire Area AssessmentNEI 00-01 Ref NEI 00-01 Guidance3.4.1.8 Alternate/Backup Equipment Consider selecting other equipment that can perform the same safe shutdown function as the impacted equipment. In addressing thisSelection situation, each equipment impact, including spurious operation, is to be addressed in accordance with regulatory requirements and theNPP's current licensing basis. With respect to MSOs, the criteria in Section 4, Appendix B, Appendix G and Appendix H should be used.Applicability CommentsApplicable Consideration of Multiple Spurious Operations is addressed in the NSCA (EIR 51-9133191, Section 8.6).Alignment StatementAlignsAlignment BasisComponent selection was performed for all fire areas in order to populate the database with equipment information required to be analyzed against the requirements of NFPA805.10CFR50, Appendix R. The components selected are documented in the Safe Shutdown Equipment List (SSEL). The objective of the SSEL is to provide a list of analyzedcomponents that are utilized in the NFPA 805 NSCA to demonstrate the fuel can be maintained in safe and stable condition, post firo safo shutdwn analysis to .ncuro that (1) onesuccesrs path (turutu Fr.s, system s, and rnmpo n ents) noceessarFy to aih ioo w-sA-afe s hut-do- ie' 46 frog o1Gf firoe dam ag a without red itinRg plant Or systemA repair F apab 0iities and (2 ) onesucces6 path (sVUtructuo, systems, and components) necessary to adhieve Gcold- s-hutd-GHow %it-hin 72 haoura is free of fira damage, Or Awaiqlabl ;Athin 72 hour-s after Grediting plant Orsystem ropair -apabilities-.The current SSEL was reviewed against the criteria outlined in NEI 00-01 and considered where additional equipment may need to be included to addressmultiple spurious operation concems or other separation concerns.Consideration of component spurious actuation is not limited to the licensing basis criteria documented in UFSAR Appendix 10B, Section 5.9.4. As part of the transition to a NFPA 805licensing basis, the criteria used in evaluating spurious actuation of components are those identified in NEI 00-01, Section 4, Identification and Treatment of Multiple SpuriousOperations, which envelopes the plant licensing basis as discussed in the UFSAR. MSO component combinations were included in the assessments.The number of potential spuriously operating valves in a line was not limited by number. The NSCA incorporates equipment identified during the review and the cable selection phaseby providing an updated SSEL Report of the safe shutdown primary components and the Safe Shutdown Success Paths. In addition, the NSCA supports incorporation of secondarycomponents into the database that were modeled as a result of the primary component selections.A success path determination for all safe shutdown functions was performed. Generally the path with the least amount of failures was utilized to demonstrate a success path for safeshutdown. Support systems were reviewed in order to assess the impact on the systems being supported.Primary and secondary shutdown methodologies have been developed. The primary method employs the use of Emergency Condensers. The secondary method employs the use ofERVs and ECCS equipment. This approach provides for versatility by employing diverse equipment resulting in four potential shutdown success paths.Reference DocumentEIR 51-9133191-000, NSCA, Section 8.6HNP RAI 3-14, NRC Request for Additional Information dated August 6, 2009 (ML092170715)Technical Report on Identification & Classification of the NMP-1 MSO Scenarios Using an Expert Panel -Review of New Scenarios, Rev.1Closure of National Fire Protection Association 805 Transition Program Frequently Asked Question Number 08-0054 (ML110140183)UFSAR Appendix lOB, Section 5.9.4NMPI, April 2013 Page B-94 IINMPI, April 2013Page B-94 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.4 Fire Area AssessmentNEI 00-01 Ref NEI 00-01 Guidance3.4.1.9 Fluid Density Effects Consider the effects of the fire on the density of the fluid in instrument tubing and any subsequent effects on instrument readings orsignals associated with the protected safe shutdown path in evaluating postfire safe shutdown capability. This can be done systematicallyor via procedures such as Emergency Operating Procedures.Applicability CommentsApplicable Instrument tubing failure due to a fire is addressed in the NSCA (EIR 51-9133191, Section 1.14).Alignment StatementAlignsAlignment BasisInstrument sensing lines for level, pressure, flow, etc. that are exposed to a fire are considered to have the potential of causing erratic or unreliable signals or indication, unless a firehazards analysis demonstrates that this failure is not credible. Fire damage to instrument sensing lines can be as detrimental to the instruments as fire damage is to safe shutdowncables and components. Even though the integrity of the tubing is expected to withstand the fire, the accuracy of the instrument may not be reflected correctly due to the heating of thefluid.The instrument sensing lines route locations are developed and inputted as design input to the analysis. The input consisted of a list of instrument sensing lines (located outsideContainment) including the fire areas and associated routing locations through the plant. This information was entered into the database as cables using fictitious cable numbers,including the route and endpoint identifications.The analysis treated the tubing like cables and associated it with the instrument. The sensing lines are subject to the same compliance issues and similar analytical techniques as safeshutdown cables. Sensing lines of instruments required for safe shutdown are included within the scope of a fire area assessment. In this manner, the sensing lines are included forconsideration along with cables when performing the fire area assessments. If instruments were impacted by the fire, then alternate instruments, not impacted by the fire, would berelied upon for safe shutdown.Instrument sensing lines were reviewed for susceptibility to physical fire damage that may cause a loss of inventory. Sensing lines for SSEL components are constructed of eitherstainless steel or carbon steel. Consequently, they are not susceptible to physical damage as the result of a postulated fire.Reference DocumentEIR 51-9133191, NSCA, Sections 5.4 and 8.4HNP RAI 3-15, NRC Request for Additional Information dated August 6, 2009 (ML092170715)NMPI, April 2013 Page B-95 IINMP1, April 2013Page B-95 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.4 Fire Area AssessmentNEI 00-01 Ref3.4.2 Methodology for Fire AreaAssessmentNEI 00-01 GuidanceRefer to NEI 00-01 Figure 3-5 for a flowchart illustrating the various steps involved in performing a fire area assessment. Use thefollowing methodology to assess the impact to safe shutdown and demonstrate Appendix R complianceApplicability CommentsApplicable NoneAlignment StatementNot RequiredAlignment BasisThis paragraph provides introductory information and contains no specific requirements. Discussion is provided in subsequent sub-sections.Reference DocumentNMPI, April 2013 Page B-96 I!NMP1, April 2013Page B-96 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.4 Fire Area AssessmentNEI 00-01 Ref3.4.2.1 Identify the AffectedEquipment by Fire AreaNEI 00-01 GuidanceIdentify the safe shutdown cables, equipment and systems located in each fire area that may be potentially damaged by the fire. Providethis information in a report format. The report may be sorted by fire area and by system in order to understand the impact to each safeshutdown path within each fire area.ApplicabilityApplicableCommentsNoneAlignment StatementAlignsAlignment BasisThe information needed to support the Post Fire Safe Shutdown Analysis is maintained in a safe shutdown database which contains the safe shutdown systems, components, cables,and their associated fire area location. This information is available in report format and can be sorted by fire area, system, train, component, cable, safe shutdown path, or variouscombinations of each. These reports are used to assess potential damage due to fire in each area of the plant. The database reports provide the same information identified onAttachment 5 of NEI 00-01.Reference DocumentEIR 51-9133191, NSCA, Sections 5.4 and 9.0NMPI, April 2013 Page B-97 IINMPI, April 2013Page B-97 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.4 Fire Area AssessmentNEI 00-01 Ref3.4.2.2 Determine the Least ImpactedShutdown PathNEI 00-01 GuidanceBased on a review of the systems, equipment and cables within each fire area, determine which shutdown paths are either unaffected orleast impacted by a postulated fire within the fire area. Typically, the safe shutdown path with the least number of cables and equipmentin the fire area would be selected as the required safe shutdown path. Consider the circuit failure criteria and the possible mitigatingstrategies, however, in selecting the required safe shutdown path in a particular fire area. Review support systems as a part of thisassessment since their availability will be important to the ability to achieve and maintain safe shutdown. For example, impacts to theelectric power distribution system for a particular safe shutdown path could present a major impediment to using a particular path for safeshutdown. By identifying this early in the assessment process, an unnecessary amount of time is not spent assessing impacts to thefrontiine systems that will require this power to support their operation. Determine which components are required hot shutdowncomponents and which components are important to SSD components using the guidance in Appendix H.Based on an assessment as described above, designate the required safe shutdown path(s) for the fire area. Classify the components onthe required safe shutdown path necessary to perform the required safe shutdown functions as required safe shutdown components.Identify all equipment not in the safe shutdown path whose spurious operation or mal-operation could affect the shutdown function.Criteria for classifying these components as required for hot shutdown or as important to SSD is contained in Appendix H. Include theaffected cables in the shutdown function list. For each of the safe shutdown cables (located in the fire area) that are part of the requiredsafe shutdown path in the fire area, perform an evaluation to determine the impact of a fire-induced cable failure on the correspondingsafe shutdown equipment and, ultimately, on the required safe shutdown path.When evaluating the safe shutdown mode for a particular piece of equipment, it is important to consider the equipment's position for thespecific safe shutdown scenario for the full duration of the shutdown scenario. It is possible for a piece of equipment to be in two differentstates depending on the shutdown scenario or the stage of shutdown within a particular shutdown scenario. Document informationrelated to the normal and shutdown positions of equipment on the safe shutdown equipment list.ApplicabilityApplicableCommentsNoneAlignment StatementAlignsAlignment BasisThe Fire Area Compliance Assessments demonstrate the ability to achieve safe shutdown by ensuring at least one safe shutdown success path is available to accomplish theperformance goals identified in NEI 00-01. NEI 00-01, Section 3 was used as guidance in performing the assessments. The elements of the assessments performed for NMP1 reflectthe NEI 00-01 guidance as discussed in the following.Fire Area Assessments were performed on a Fire Area basis in order to ensure compliance in accordance with the requirements of NFPA 805.10 CFR 50, R. The safeshutdown database reports provide the potentially affected equipment and cables in each Fire Area, which were analyzed for impacts on safe shutdown success paths.The Safe Shutdown Equipment List (SSEL) contains equipment data such as the equipment type, description, safe shutdown path, drawing reference, fire area, fire zone, and roomlocation. Other equipment information would include normal position, hot shutdown position, cold shutdown position, failed air position, failed electrical position, high/low pressureinterface concern, and spurious operation concern.INMPI, April 2013 Page B-98 IINMP1, April 2013Page B-98 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology ReviewThe route and location of all Appendix R cables (located by Fire Area) was used to generate the Fire Area Component Impact Reports, which identified affected systems andcomponents on a Fire Area basis. The Fire Area Component Impact Reports were used as a means to determine the least impacted safe shutdown path for each fire area.The cable selection task involved identifying all cables associated with the control and operation of a safe shutdown component. These cables were analyzed to determine the impactof fire induced cable failure on the selected equipment. A circuit analysis was performed as part of the scope for selected cables/components, as required, in order to demonstrate thatthe cable is or is not required so that the analyzed component can be credited to perform its required function for the safe shutdown path.The Fire Area Analysis methodology identified fire-induced component and cable failures and spurious actuations, based on the cables and components present in the fire area ofconcern. All postulated cable and component failures were assessed and a resolution provided at the component level.Once the above was complete, the least impacted safe shutdown success path was identified so that mitigating strategies could be developed and documented in the Fire AreaAssessment.Reference DocumentEIR 51-9133191, NSCA, Sections 5.1, 5.2, 5.3, and 5.4NMPI, April 2013 Page B-99 IINMPI, April 2013Page B-99 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.4 Fire Area AssessmentNEI 00-01 Ref3.4.2.3 Determine Safe ShutdownEquipment ImpactsNEI 00-01 GuidanceUsing the circuit analysis and evaluation criteria contained in Section 3.5 of this document, determine the equipment that can impact safeshutdown and that can potentially be impacted by a fire in the fire area, and what those possible impacts are.ApplicabilityApplicableAlignment StatementAlignsCommentsNoneAlignment BasisThe Safe Shutdown Equipment List was developed based on system requirements and other plant impacts. During the identification of the Safe Shutdown Equipment List forcomponent cables, a circuit fault analysis for each component's cables was performed to determine the effects of a fire-induced hot short, open circuit and short-to-ground. The circuitsassociated with the components operation and whose failure could affect the components operation was considered as required. The fire area analysis assumed multiple fire-inducedfailures and multiple spurious actuations (MSOs), based on the cables and components present in the fire area of concern. The cable and component failures were evaluated and aresolution and disposition was provided for component and cable impacted in that fire area.In addition, spurious operating equipment concerns are addressed in the MSO Expert Panel Report," which consists of the multiple spurious operation review. The purpose of thisreview is to document the potential Multiple Spurious Operation combinations.The results of this activity identifies equipment, whose fire-induced spurious operation could result in consequences that may be adverse to both the Fire PRA risk models and meetingthe nuclear safety performance criteria of NFPA 805. The equipment identified in this task that could affect the Fire PRA will be integrated into the Fire PRA Equipment list.Reference DocumentEIR 51-9133191, NSCA, Sections 5.0, 8.0, and 9.0ONS RAI 3-41 and RAI 3-43, NRC Request for Additional Information dated July 30, 2000 (ML102110394)Technical Report on Identification & Classification of the NMP-1 MSO Scenarios Using an Expert Panel -Review of New Scenarios, Rev.1NMPI, April 2013 Page 6-100 I!NMP1, April 2013Page B-100 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.4 Fire Area AssessmentNEI 00-01 Ref NEI 00-01 Guidance3.4.2.4 Develop a Compliance The available deterministic methods for mitigating the effects of circuit failures are summarized as follows (see Figure 1-1):Strategy or Disposition Required for Hot Shutdown Components:" Re-design the circuit or component to eliminate the concern. This option will require a revision to the post-fire safe shutdownanalysis." Re-route the cable of concern. This option will require a revision to the post-fire safe shutdown analysis." Protect the cable in accordance with III.G.2." Provide a qualified 3-fire rated barrier." Provide a 1-hour fire rated barrier with automatic suppression and detection." Provide separation of 20 feet or greater with automatic suppression and detection and demonstrate that there are no interveningcombustibles within the 20 foot separation distance." Perform a cold shutdown repair in accordance with regulatory requirements." Identify other equipment not affected by the fire capable of performing the same safe shutdown function." Develop exemptions, deviations, LARs, Generic Letter 86-10 evaluation or fire protection design change evaluations with a licensingchange process.Important to Safe Shutdown Components:" Any of the options provided for required for hot shutdown components." Perform and operator manual action in accordance with Appendix E." Address using fire modeling or a focused-scope fire PRA using the methods of Section 5 for MSO impacts. [Note: The use of firemodeling will require a review by the Expert Panel and the use of a focused-scope fire PRA will require a LAR.]Additional options are available for non-inerted containments as described in 10 CFR 50 Appendix R section IlI.G.2.d, e and f.Applicability CommentsApplicable NoneAlignment StatementAlignsAlignment BasisThe safe shutdown analysis provides a compliance strategy and various deterministic methods used for mitigating the effects of circuit failures. Potential impacts to safe shutdownwere addressed by using the path least impacted by the fire to assure at least one success path for safe shutdown. This was accomplished by using a combination of the strategieslisted in the guidance and taking credit for any existing features whenever possible.Circuit failures having the potential to adversely impact the shutdown process were identified as Open Items to be transitioned to the Fire Risk Evaluations.Reference DocumentEIR 51-9133191, NSCA, Section 9.0!NMP1, April 2013Page B-101 I Constellation Energy Nuclear GroupAttachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review2.4.2.4 Fire Area AssessmentNEI 00-01 Ref3.4.2.5 Document the ComplianceStrategy or DispositionNEI 00-01 GuidanceAssign compliance strategy statements or codes to components or cables to identify the justification or mitigating actions proposed forachieving safe shutdown. The justification should address the cumulative effect of the actions relied upon by the licensee to mitigate a firein the area. Provide each piece of safe shutdown equipment, equipment not in the path whose spurious operation or mal-operation couldaffect safe shutdown, and/or cable for the required safe shutdown path with a specific compliance strategy or disposition. Refer toAttachment 6 for an example of a Fire Area Assessment Report documenting each cable disposition.ApplicabilityApplicableAlignment StatementAlignsCommentsAlignment BasisFor the Safe Shutdown Analysis, success paths were developed and analyzed for each component impacted by a fire in the subject area using disposition codes that representconsistent standardized compliance statements. The compliance statement reflects the credited fire protection features, analysis of credible cable failures, and serves as the basis forachieving safe shutdown conditions for the analyzed fire areas. The disposition codes (i.e., resolution of component hits) and associated statements were entered into the safeshutdown database. The following is a sample of generic disposition codes used for the Fire Area Compliance Assessment:* Cable protected by rated fire barrier.* Failure of cable may result in loss of power/control of component.* Failure of cable may result in loss of indication or erroneous indication.* Component fails in desired SSD position/mode.* Series isolation valve(s) available and can be closed.* Capability to close valve is available from the MCR.* Component remains in desired SSD position.* Valve can spuriously open. Series isolation valve remains closedReference DocumentEIR 51-9133191, NSCA, Section 9.0NMPI, April 2013 Page B-102 IINMP1, April 2013Page B-102 I REVISIONS TO TRANSITION REPORT ATTACHMENT FFIRE-INDUCED MULTIPLE SPURIOUS OPERATIONS RESOLUTIONPages F-6 and F-7 with changes highlighted.

Constellation Energy Nuclear GroupAttachment F -Fire-induced MS09 ResolutionResults of Step 3:The results of the expert panel were included in Task 7.3.1 (NUREG/CR-6850 Task 2)and Task 7.4 (NUREG/CR-6850 Task 3) within the scope of the NMPI Fire PRA, and inTask 4.2.2, Table B-3 and Fire Area Analysis within the scope of the NMPI NSCA.Task 7.3.1 addressed spurious operations, including multiple spurious operationsidentified in the post-fire safe shutdown analysis, and those that resulted from the expertpanel review.The results of the Fire PRA model update are included in NMPI Fire PRA Notebook,"Equipment Selection," which includes the following MSO related information:" Identification and disposition of equipment from the review of MSOs (Table D-1of the "Equipment Selection" notebook); and" Fire PRA equipment list, which includes MSO identified components and theirassociated basic events (Table G-1 of the "Equipment Selection" notebook).The MSO combination components were also evaluated for inclusion into the NMPINSCA. As necessary, components were added to the NSCA Equipment List andLogics, and the appropriate circuit analysis and cable routing were performed.Step 4 -Evaluate for NFPA 805 ComplianceThe MSO combinations included in the NSCA should be evaluated with respect tocompliance with the deterministic requirements of NFPA 805, as discussed in Section4.2.3 of NFPA 805. For those situations in which the MSO combination does not meetthe deterministic requirements of NFPA 805 (VFDR), the issue with the componentsand associated cables should be mitigated by other means (e.g., performance-basedapproach per Section 4.2.4 of NFPA 805, plant modification, etc.).The performance-based approach may include the use of feasible and reliable recoveryactions. The use of recovery actions to demonstrate the availability of a success pathfor the nuclear safety performance criteria requires that the additional risk presented bythe use of these recovery actions be evaluated (NFPA 805 Section 4.2.4).Results of Step 4:The MSO combination components of concern were evaluated as part of the NMPINSCA and NPO analyses. For NSCA cases where the pre-transition MSO combinationcomponents did not meet the deterministic compliance, the MSO combinationcomponents were added to the scope of the fire risk evaluations. The process andresults for Fire Risk Evaluations are summarized in Section 4.5 of the Transition Report.MSO scenarios impacting NPO POSs are associated with Key Safety Function (KSF)success paths in the NPO analysis. A fire induced loss of these KSF success paths isaddressed through the FAQ 07-0040 resolution process, wherein recommendations areprovided to best manage fire risk in pinch point plant areas where KSFs may beimpacted by a fire (EIR 51-9171174).Step 5 -Document ResultsThe results of the process should be documented. The results should provide adetailed description of the MSO identification, analysis, disposition, and evaluation!NMP1, April 2013Page F-6 I Constellation Energy Nuclear GroupAttachment F -Fire-induced MSOs Resolutionresults (e.g., references used to identify MSOs; the composition of the expert panel, theexpert panel process, and the results of the expert panel process; disposition andevaluation results for each MSO, etc.). High level methodology utilized as part of thetransition process should be included in the 10 CFR 50.48(c) License AmendmentRequest/Transition Report.Results of Step 5:The NMP1 Results are documented in:" "Resolution of Issues Related to Fire-Induced Circuit Failures, Technical Reporton Identification & Classification of the NMP1 MSO Scenarios using an ExpertPanel"" NMP1 Fire PRA Notebook, N1 -ES-FOO1, "Equipment Selection (ES)"" NMP1 Fire PRA Notebook, N1-CS-F001, "Cable Selection, Detailed CircuitAnalysis and Route Location (CS)"" NMP1 Fire PRA Notebook, N1-PRM-F001, "Plant Response Model"" NMP1 Fire Area Transition -See Attachment C (NEI 04-02 Table B-3) of theTransition Report" EIR 51-9133191, NMP1 Nuclear Safety Capability Assessment (NSCA) Report" EIR 51-9137629, NMP1 Non-Power Operations KSF Equipment List" EIR 51-9171174, NMP1 NFPA 805 Transition- Non-Power OperationsComponent Pinch Point AnalysisNMPI, April 2013Page FT I REVISIONS TO TRANSITION REPORT ATTACHMENT GRECOVERY ACTIONS TRANSITIONPages G-1 through G-41 with changes highlighted.

Constellation Enerav Nuclear GrouoAttachment G -Recoverv Action TransitionC.n.tellation...e... Nuclear..... Gru Attahmen G eoeyAtinTastoG. Recovery Actions Transition36-31 Pages AttachedNMPI, April 2013 Page G-1 IINMPI, April 2013Page G-1 I Constellation Energy Nuclear GroupAttachment G -Recovery Action TransitionIn accordance with the guidance provided in NEI 04-02, FAQ 07-0030, Revision 5, andRG 1.205, the following methodology was used to determine recovery actions requiredfor compliance (i.e., determining the population of post-transition recovery actions). Themethodology consisted of the following steps:" Step 1: Define the primary control station(s) and determine which pre-transitionOMAs are taken at primary control station(s) (Activities that occur in the MainControl Room are not considered pre-transition OMAs). Activities that take placeat primary control station(s) or in the Main Control Room are not recoveryactions, by definition." Step 2: Determine the population of recovery actions that are required to resolveVFDRs (to meet the risk acceptance criteria or maintain a sufficient level ofdefense-in-depth)." Step 3: Evaluate the additional risk presented by the use of recovery actionsrequired to demonstrate the availability of a success path." Step 4: Evaluate the feasibility of the recovery actions." Step 5: Evaluate the reliability of the recovery actions.An overview of these steps and the results of their implementation are provided below.Step 1 -Clearly define the primary control station(s) and determine which pre-transition OMAs are taken at primary control station(s)The first task in the process of determining the post-transition population of recoveryactions was to apply the NFPA 805 definition of recovery action and the RG 1.205definition of primary control station to determine those activities that are taken atprimary control station(s).Results of Step 1:Based on the definition provided in RG 1.205, and the additional guidance provided inFAQ 07-0030, the following locations are considered as taking place at the primarycontrol station(s):1. Remote Shutdown Panel 11 is located in Fire Area 7, Fire Zone T2B, TurbineBuilding Elevation 250'-0".2. Remote Shutdown Panel 12 is located in Fire Area 5, Fire Zone T4A, TurbineBuilding Elevation 277'-0".The remote shutdown panels were approved by the NRC in SER entitled "SubjectModifications and Alternate Safe Shutdown Capabilities to Comply with theRequirements of Appendix R", dated March 3, 1983.Table G-1 -"Recovery Actions and Activities Occurring at the Primary ControlStation(s)" identifies the activities that occur at the primary control station(s). Activitiesnecessary to enable the primary control station(s) are also identified in Table G-1 asprimary control station(s) activities. These activities do not require the treatment ofadditional risk and are compliant with NFPA 805, Section 4.2.3.1.NMPI, April 2013 Page G-2 IINMPI, April 2013Page G-2 I Constellation Energy Nuclear GroupAttachment G -Recovery Action TransitionCoseltonEeg ulear Gru tahetG-Rcvr cinTastoNote that the Remote Shutdown Panels (RSPs) are primary control station(s) only for afire in Fire Area 11 for which a MCR evacuation is credited.Step 2 -Determine the population of recovery actions that are required to resolveVFDRs (to meet the risk or defense-in-depth criteria)On a fire area basis, all VFDRs were identified in the NEI 04-02 Table B-3 (seeAttachment C). Each VFDR not brought into compliance with the deterministicapproach was evaluated using the performance-based approach of NFPA 805 Section4.2.4. The performance-based evaluations resulted in the need for recovery actions tomeet the risk acceptance criteria or maintain a sufficient level of defense-in-depth.Results of Step 2:The FRE report provides the determination of recovery actions required to resolveVFDRs. These recovery actions are listed in Table G-1, "Recovery Actions andActivities Occurring at the Primary Control Station(s)."The actions contained in Table G-1 are identified on a fire area basis. Many of thesame actions are repeated in different fire areas. To assist in understanding the varioustypes of recovery actions contained in Table G-1, Table 4-1 has been created. Table 4-1 is a list of unique recovery actions only and does not include primary control stationactions. It is important to note that not every component listed in the Componentscolumn of Table 4-1 is associated with every fire area listed in the Fire Areas ofConcern column. The Fire Areas of Concern reflects the aggregate list of fire areaswhere the type of unique recovery action is credited to support shutdown. It is alsoimportant to note that item 4---12 in Table 4-1 is a proposed modification identified inAttachment S, specifically Table S-1. The final set of recovery actions is provided inTable G-1 -Recovery Actions and Activities Occurring at the Primary Control Station(s).NMPI, April 2013 Page G-3 I!NMPI, April 2013Page G-3 I Constellation Enerav Nuclear GroUDAttachment G -Recoverv Action TransitionTable 4-1: Unique Operator Recovery ActionsNo. Action Description Components Fire Areas ofConcern1 Operator determines vital LI-36-26, LI-36-28, LI-60-22C, LI-60-23C, 1,2, 5,10parameters PI-201.2-94, PI-201.2-5, PI-36-25, PI-36-at RSP 27, TI-201-50B, TI-201-51B TI-201.2-521B,TI-201.2-522B, TI-32-02B, TI-32-03B, TI-32-04B, TI-32-05B2 Operato lcally toproide W470-f3 4, 5, 6, 7, 0, 10, 11, 1 2,cooling MIto to 6hutdon4 43, 41 5,4I1A, 4gB,Gegl!1U.pS 17A, 2 OpVate locally to maintain RV-38 01, FVLV860-1 3, 5, 6, 7, 0, 10, 11, 12,conol domn rate 3811, iv38 01, 11 as 02, 1 11, 15,16A, 16B,E8n2 417A, 17B, 18,2920,24 Wirng rpaired QAd IV 39 01, IV 38 13, PMP 38 140, PMP 3 2 22,273,2044conhgent opereatd 45224 Valve locally throttled to VLV-60-11, VLV-60-12 4, 5,6,7,9,10,11, 12,control make-up to the 13, 14, 15,16A, 1613,Emergency Condensers 17A, 17B, 18,19,20,_________________21,22,23,2436 Connect portable charger BAT-B1I, BAT-B312 5,6, 7,9, 10, 11to charge Batteries__________47- Manually isolate to prevent FCV-39-15, FCV-39-16, VLV-05-311, VLV- 5, 6, 9,10, 11,18, 22,inventory loss 05-32 23, 2458 Vent air to close valve to IV-01-03, IV-01-04 5,6,7,9,10,11,18,prevent inventory loss 22, 23, 2460 Vent air to open to IV-39-05, IV-39-06 5, 10establish EmergencyCondensers on failure toopen740 Verify tripped for load PB-BB131, PB-BB12, UPS-UPS162A, 5, 6, 7, 9,10, 11shedding UPS-UPS162B, ,-UPS-UPS172A, UPS-UPS172B448 Shut down locally for load PMP-79.1-01, PMP-79.1-07, PMP-79.1-20, 5,6,7, 9,10,11shedding PMP-79.1-26942 Locally operate to IV-39-07R, IV-39-08R, IV-39-09R, IV-39- 10establish decay heat IORremoval throughEmergency Condensers1042 Emergency Condenser LCV-60-17, LCV-60-18 10Level Control Transferswitch to local4411 Operate manually to VLV-93-13, VLV-93-16 5, 11,24isolate Containment sprayon spurious start1245 Open NEW Disconnect to Recover Emergency Diesel Generator 5,7,9, 10, 11load Emergency DieselGenerator to Dead bus(Currently a DamageRepair Procedure action)13461 Scram control rods by BV-113-3091, VLV-113-230 5, 7,10, 11venting the scram airheader14 Locally operate the Fire BV-100-68, BV-100-69, PMP-100-02 4, 5, 6, 7, 9, 10, 11, 12,Water System using the 13,14,15, 16A, 16B,DFP to provide long term 17A, 17B, 18,19,20,Emergency Condenser 21, 22, 23, 24makeup tank supplyIIINMPI, April 2013Page G-4 I Constellation Enerav Nuclear GroupAttachment G -Recovery Action TransitionCoselto Eeg ula Gru AtahetG- eoeyAtinTastoStep 3: Evaluate the Additional Risk of the Use of Recovery ActionsNFPA 805 Section 4.2.3.1 does not allow recovery actions when using the deterministicapproach to meet the nuclear safety performance criteria. However, the use of recoveryactions is allowed by NFPA 805 using a risk informed, performance-based, approach,provided that the additional risk presented by the recovery actions is evaluated inaccordance with NFPA 805 Section 4.2.4.Results of Step 3:The set of recovery actions that are necessary to demonstrate the availability of asuccess path for the nuclear safety performance criteria (see Table G-1) were evaluatedfor additional risk using the process described in NEI 04-02, FAQ 07-0030, Revision 5,and RG 1.205 and compared against the guidelines of RG 1.174 and RG 1.205. Noneof the recovery actions were found to have an adverse impact on the Fire PRA. Theadditional risk of recovery actions is provided in Attachment W.Step 4: Evaluate the Feasibility of Recovery ActionsRecovery actions were evaluated against the feasibility criteria provided in NEI 04-02,FAQ 07-0030, Revision 5, and RG 1.205. Note that since actions taken at the primarycontrol station are not recovery actions their feasibility is evaluated in accordance withprocedures for validation of off normal procedures.Results of Step 4:The HRA evaluated the feasibility of recovery actions modeled in the Fire PRA andused to resolve VFDRs identified in the B-3 Table. This includes recovery actionsrelated to AC power, Emergency Diesel Generators, and long-term decay heat removalamong others. Feasibility of these recovery actions were evaluated in the HRA againstthe criteria outlined in NEI 04-02, FAQ 07-0030 Revision 5, and RG 1.205, makingextensive use of HEP quantifications.Recovery actions that are required by the FRE but not addressed in the HRA wereevaluated for feasibility using the NEI 04-02, FAQ 07-0030 Revision 5, and RG 1.205criteria and documented in EIR 51-9156521 entitled, "Recovery Action Review for NineMile Point Nuclear Power Station Unit 1 Transition to NFPA 805."Since actions taken at primary control stations are not recovery actions, no independentfeasibility evaluation is required.Results of the feasibility assessments in the HRA and in the EIR demonstrate that allcredited NFPA 805 recovery actions are feasible.Implementation items resulting from the feasibility evaluation include:Modify, as needed, the following procedures for recovery actions being evaluated:" NI-SOP-21.1" N1-SOP-21.2Operators will be trained and qualified on the revised procedures.INMPI, April 2013Page G-5 I Constellation Energy Nuclear GroupAttachment G -Recovery Action TransitionThese items are included as implementation items in Attachment S.Step 5: Evaluate the Reliability of Recovery ActionsThe evaluation of the reliability of recovery actions depends upon its characterization." The reliability of recovery actions that were modeled specifically in the Fire PRAwere addressed using Fire PRA methods (i.e., HRA)." The reliability of recovery actions not modeled specifically in the Fire PRA arebounded by the treatment of additional risk associated with the applicable VFDR.In calculating the additional risk of the VFDR, the compliant case recovers thefire-induced failure(s) as if the variant condition no longer exists. The resultingdelta risk between the variant and compliant condition bounds any additional riskfor the recovery action even if that recovery action were modeled.Results of Step 5:The reliability of recovery actions that were modeled specifically in the Fire PRA wereaddressed using Fire PRA methods. The HRA addresses the reliability of theserecovery actions, with consideration taken for various performance shaping factors,including cues and instrumentation, timing, procedures and training, complexity,workload pressure and stress, human-machine interface, environment, specialequipment, specific fitness needs, as well as crew communications, staffing, anddynamics. Accordingly, the HRA also evaluates recovery actions depending on whetherthey correspond or not to main control room abandonment situations.Recovery actions that are required by the FRE but not addressed in the HRA areevaluated for reliability and documented in EIR 51-9156521 entitled, "Recovery ActionReview for Nine Mile Point Nuclear Power Station Unit 1 Transition to NFPA 805."Since actions taken at primary control stations are not recovery actions, no independentreliability evaluation is required. It should however be noted that a reliability evaluationdocumented in the HRA was made for those actions taken at PCSs that are creditedand modeled in the Fire PRA.Results of the reliability assessments in the HRA and in EIR 51-9156521 demonstratethat all credited NFPA 805 recovery actions are reliable.NMPI, April 2013 Page G-6 IINMPI, April 2013Page G-6 I Constellation Enerav Nuclear GrouDAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)Fire Component Component Description Actions VFDR RA/PCSArea1 LI-36-28 REACTOR VESSEL LEVEL Operator determines vital parameters VFDR-01-009 RATI-201-50B DRYWELL TEMPERATURE from PNL-RSP 11. VFDR-01-0111 LI-36-26 REACTOR VESSEL LEVEL Operator determines vital parameters VFDR-01-009 RATI-201-51B DRYWELL TEMPERATURE from PNL-RSP12. VFDR-01-0112 LI-36-28 REACTOR VESSEL LEVEL Operator determines vital parameters VFDR-02-007 RATI-201-50B DRYWELL TEMPERATURE from PNL-RSP11. VFDR-02-008TI-201.2-521 B TORUS TEMPERATURE VFDR-02-0092 LI-36-26 REACTOR VESSEL LEVEL Operator determines vital parameters VFDR-02-007 RATI-201-51B DRYWELL TEMPERATURE from PNL-RSP12. VFDR-02-008TI-201.2-522B TORUS TEMPERATURE VFDR-02-0094 BV4O 63 44" AIR OPERATED BLOCKING 8V 70 63 is operatod locally to providc VFDR 04 006 RAVALVE BLC INLIE-TVALV 98oling Wator to the 69G PUMPS and ahWant a*nk far flip Q=C HXc4 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-04-009 RABV-100-69 & 122 MANUAL VALVES opened to periodically refill theEmergency Condenser Makeup tanks.4 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-04-009 RATURBINE FIRE PUMP run as needed to supply fire water toEmergency Condenser Makeup tanks.4 :RQV2-, Q 8" O.4 ^'PERATED FCV 38 09 is opo.atod locally to VFDR-04 8 RAFLO " CO T O.= .L VAV E cAn tQoe f&-- t-rough S -D C HX 3 8 1 362 to2SIT-0UTDOW C=OLINGZ HEA rgulato RCS cool downi Fato.EXCH.ANGEcR 11 FLOWCON-TRO'-VAIU4 AVK EPr DIAPRAG OPERATED FC3 TO 2O 10 is laoallytottlcl to VFDR-04-907 RAFLOWGCNTROL VALO E GOntE ol to Ewtr 1u 8G 12D HX 3p 122 IegHJTDWNl COIN HEAT C W egulato RsS corl decay FateoEXCANER12FLODW4 FGV-as i4 A" DIAPHR AGM OPERATED9 FCY 239 11 is opoatod locally to 1VFDR Q4 998 RAFLO COTRO VAVE onrol'10A flow throug SDC HX 38 120 toSHU TDOWNgh COO INGC HEA rogwlate RCS seel down rato.ECAGER 12 FLOW4 VLV-60-1 1 MAKEUP VALVE TO LOOP 12 VLV-60-11I is locally throttled to control VFDR-04-004 RAEMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide aRCS heat sink for decay heat removal.INMP1, April 2013Page G-7 I Constellation Eneray Nuclear GroupAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)FireArea Component Component Description Actions VFDR RA/PCS4 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-04-003 RAEMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide aRCS heat sink for decay heat removal.5 BAT-B1 1 125 VOLT DC STATION Portable charger with a generator, VFDR-05-035 RABATTERY 11 connected directly to Battery Board PB-BBI1 to charge Battery 811.5 BAT-B12 125 VOLT DC STATION Portable Charger with Generator VFDR-05-035 RABATTERY NUMBER 12 connected directly to Battery Board PB-BB12 to charge Battery B12.5 BKR-(103/1-1) DIESEL GEN 103 OUTPUT Operate disconnect switch locally. VFDR-05-040 RAR10321581 BREAKER 103/1-1 (R1032/581) to VFDR-05-043PB-1035 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-05-047 RABV-100-69 & 122 MANUAL VALVES opened to periodically refill theEmergency Condenser Makeup tanks.5 BV-1 13-3091 BLOCKING VALVE FOR AIR Operator unlocks and doses BV-1 13- VFDR-05-046 RASUPPLY TO SCRAM AIR 3091. Operator removes the vent pipeSYSTEM cap, unlocks, and opens VLV-1 13-VLV-1 13-230 BALL VALVE -SCRAM VALVE 230 to vent the SCRAM air header toPILOT HEADER VENT insert the control rods.6 6-V38-04 SH WTDOWN COOLNG PUMP 12 BV 38 04 ic oporat"d !Wcally to alin VFDR-0-6044 RASUCTION LOKIN 6LLV DC pump PMP 389 162 to #;e RCS toinfitatok Ct e wd eo heatFameval.6 RV70 !I," .AJR OpEp.RATED BLOC.KING BV 70 53 I6 ope-ated lca,3y to psodd.d VFDR 020 RAV.ALVE. RBCLCG INLET VALVE cooling ateF to the 6DC PUMPS and aTO SHUTDOWN~h COOL- IING heat cink for the SDC Hgs.-Y-S-T-rem5 EG-EDG103 EMERGENCY GENERATOR -Operate disconnect switch locally. VFDR-05-037 RAEMERGENCY DIESEL VFDR-05-038GENERATOR UNIT 1036 FGV 8- A2 DI.A-PHR-AG OPERA.TED FCV 38 10 ic operated locally to VFQR 06-021 RAFLOW CONTROL- I VAIVE contol fow throeugh GD HX 38 122 toSHU'-T-DOWAN COOLING HEA rogulate RCS cool downi Fate.EXCHANGE-R 12 RLONCN OLVALVEGNMPI, April 2013 Page G-8 IlNMP1, April 2013Page G-8 I Constellation Enerav Nuclear GrouDAttachment G -Recoverv Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)FireArea Component Component Description Actions VFDR RA/PCS5 FCV-39-15 EMERGENCY CONDENSER FCV-39-15 operated locally via the VFDR-05-003 RASTEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory lossCONTROL VALVE -FLOW flow path from the ECs 111 & 112 RCSCONTROL VALVE 11 STEAM return path drain line.LINE DRAIN5 FCV-39-16 EMERGENCY CONDENSER FCV-39-16 operated locally via the VFDR-05-002 RASTEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory lossCONTROL VALVE -FLOW flow path from the ECs 121 & 122 RCSCONTROL VALVE 12 STEAM return path drain line.LINE DRAIN5 IV-01-03 24" AIR OPERATED ( ANGLE) Air vented manually at MSIV IV-01-03 VFDR-05-009 RAISOLATION VALVE WITH to close valve to prevent inventory loss.SOLENOID VALVES -MAINSTEAM ISOLATION VALVE 35 IV-01-04 24" AIR OPERATED ( ANGLE) Air vented manually at MSIV IV-01-04 VFDR-05-009 RAISOLATION VALVE WITH to close valve to prevent inventory loss.SOLENOID VALVES -MAINSTEAM OUT ISOLATION VALVE46 V 38- MOA.0TOIR 0OPE2RA.TE-D At PB 167, B KR (4 67!D3)62, the VFQR 06 925 RSHUTDOWN COOLING OUTLET valve ....Rn is ropairod and the "alv'INSIDE ISOLA.TION VALVE .p..atd to faclitate otabliehing aSDC suction flow path frcm the RCS to___________accosmplish docay heat Femev.al.IV-38 02 MOTOR OPERA.TED IV 38 02 i6 operatod lcally "ia the VFDR 95 4O4 RASHUTDOWN COOLI'NG OUTLET hand ";,,cl to octablich a SDC QucstionOUTSIDE ISOA.TION VALVE fle=ath fram the RCS te ar.....plch________________________docay heat romoval.5 /2 2 REACTOR SHUTDOWN At 121 167, BKR (167!IG03)52, tho VFQR 95 025 RCOOL'ING RETURN ISOLA.TION 'alve Ic repairod and the "al'eVA6VE 1 oporatod to facilitato establishing a80C diechargo flow path to the RCS to_accemplich deray heat remoral.5 IV-39-05 EMERGENCY COOLING LOOP IV-39-05 is manually opened by VFDR-05-012 RA11 CONDENSATE RETURN AIR venting air from the valve to establish a VFDR-05-013OPERATED ISOLATION VALVE decay heat removal path using EC's VFDR-05-014(GLOBE) 111 & 112.NMPI, April 2013 Page G-9 IINMP1, April 2013Page G-9 I Constellation Energy Nuclear GroupAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)Fire Component Component Description Actions VFDR RA/PCSArea5 IV-39-06 EMERGENCY COOLING LOOP IV-39-06 is manually opened by VFDR-05-012 RA12 CONDENSATE RETURN AIR venting air from the valve to establish a VFDR-05-013OPERATED ISOLATION VALVE decay heat removal path using EC's VFDR-05-014(GLOBE) 121 & 122.5 PB-BB1 I 125VDC BATTERY BOARD 11 MG 167 motor BKR-MG167(BBI1/E03) VFDR-05-028 RAverified tripped for battery loadshedding to extend battery capability.5 PB-BB12 125VDC BATTERY BOARD 12 MG 167 motor BKR-MG167(BB12/FO3) VFDR-05-028 RAverified tripped for battery loadshedding to extend battery capability.5 PMP-79.1-01 EMERGENCY DIESEL PMP-79.1-01 shutdown locally for VFDR-05-028 RAGENERATOR 102 TURBO LUBE battery load shedding to extend batteryOIL PUMP capability.5 PMP-79.1-07 EMERGENCY DIESEL PMP-79.1-07 shutdown locally for VFDR-05-028 RAGENERATOR 102 CIRCULATING battery load shedding to extend batteryLUBE OIL PUMP ( 6 GALLONS capability.PER MINUTE)5 PMP-79.1-20 EMERGENCY DIESEL PMP-79.1-20 shutdown locally for VFDR-05-028 RAGENERATOR 103 TURBO LUBE battery load shedding to extend batteryOIL PUMP capability.5 PMP-79.1-26 EDG 103 CIRCULATING LUBE PMP-79.1-26 shutdown locally for VFDR-05-028 RAOIL PUMP (6 GPM) battery load shedding to extend batterycapability.5 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-05-047 RATURBINE FIRE PUMP run as needed to supply fire water toEmergency Condenser Makeup tanks.5 LI-36-26 REACTOR VESSEL LEVEL Operator determines vital parameters VFDR-05-016 RAPI-36-27 REACTOR VESSEL PRESSURE from PNL-RSP12. VFDR-05-0175 UPS-UPS162A POWER SUPPLY -UPS 162A switches HDS-UPSI62A- VFDR-05-028 RAUNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery loadSUPPLY shedding to extend battery capability.NMPI, April 2013 Page G-10 I!NMP1, April 2013Page G-10 I Constellation Energy Nuclear GroupAttachment G -Recovery Actions TransitionCoselto nrg ula ru AtahetG RcvryAtosTastoTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)FireArea Component Component Description Actions VFDR RA/PCS5 UPS-UPS162B POWER SUPPLY -UPS 162B switches HDS-UPS162B- VFDR-05-028 RAUNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery loadSUPPLY shedding to extend battery capability.5 UPS-UPS172A POWER SUPPLY -UPS 172A switches HDS-UPS172A- VFDR-05-028 RAUNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery loadSUPPLY shedding to extend battery capability.5 UPS-UPS172B POWER SUPPLY -UPS 172B switches HDS-UPS172B- VFDR-05-028 RAUNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery loadSUPPLY shedding to extend battery capability.5 VLV-05-31 LOOP 11 VENT LINE MANUAL VLV-05-31 is operated locally via the VFDR-05-004 RAISOLATION VALVE DOWN hand wheel to isolate an inventory lossSTREAM OF 05-01R AND 05-11 flow path from the ECs 111 & 112 RCSsteam supply vent line.5 VLV-05-32 LOOP 12 VENT LINE MANUAL VLV-05-32 is operated locally via the VFDR-05-004 RAISOLATION VALVE hand wheel to isolate an inventory lossDOWNSTREAM OF 05-04R AND flow path from the EC's 121 & 12205-12 RCS steam supply vent line.5 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-11 is locally throttled to control VFDR-05-018 RAEMERGENCY CONDENSERS makeup to EC's 121 & 122 to provide aRCS heat sink for decay heat removal.5 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-05-018 RAEMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide aRCS heat sink for decay heat removal.5 VLV-93-16 12" GATE VALVE -112 VLV-93-16 closed locally via the hand VFDR-05-022 RACONTAINMENT SPRAY RAW wheel to isolate CTSRW flow toWATER PUMP DISCHARGE Containment Spray Header #12.VALVE6 BAT-B11 125 VOLT DC STATION Portable charger with a generator VFDR-06-017 RABATTERY 11 connected directly to Battery Board PB-BB11 to charge Battery B131.6 BAT-B12 125 VOLT DC STATION Portable Charger with Generator VFDR-06-017 RABATTERY 12 connected directly to Battery Board PB-BB12 to charge Battery B12.INMP1, April 2013Page G-1 I I Constellation Energy Nuclear GroupAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)FireArea Component Component Description Actions VFDR RA/PCS6 V40--63 14" AIR OPERATED ALOCGKING BV 70 53 is oporatod locally to prow.de VFDR 06 0 RAwAt. .. ... ..... .ALVE 99.. In. Wat. .to the 60G pumps an. d aTO-'RW SHTDOWMN COOLING hcatl aink for t;A 2SDC HXA.6 FGV 38 99 8 DIHPRA.GM.4. OPER-AXTED F.V 3110 is opeted locally to VFDRQ O 46- RALOC ýI fWP threugh AGG HX 38 1 t6loCOOL ING HWEAT reg8t RCS cool don ratoCONTRO' AI ý6BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-1 00-68 and BV-1 00-69 are locally VFDR-06-019 RABV-100-69 & 122 MANUAL VALVES opened to periodically refill theEmergency Condenser Makeup tanks.6 FCV-39-15 EMERGENCY CONDENSER FCV-39-15 operated locally via the VFDR-06-001 RASTEAM LINE DRAIN PRESSURE hand whieel to isolate an inventory lossCONTROL VALVE -FLOW flow path from the ECs 111 & 112 RCSCONTROL VALVE 11 STEAM return path drain line.LINE DRAIN6 FCV-39-16 EMERGENCY CONDENSER FCV-39-16 operated locally via the VFDR-06-002 RASTEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory lossCONTROL VALVE -FLOW flow path from the ECs 121 & 122 RCSCONTROL VALVE 12 STEAM return path drain line.LINE DRAIN6 IV-01-03 24- AIR OPERATED (ANGLE) Air vented manually at MSIV IV-01-03 VFDR-06-007 RAISOLATION VALVE WITH P S hnweel to latprevent inventory loss.SOLENOID VALVES -MAINSTEAM ISOLATION VALVE 36 IV-01-04 24" AIR OPERATED (ANGLE) Air vented manually at MSIV IV-01-04 VFDR-06-008 RAISOLATION VALVE WITH to close valve to prevent inventory loss.SOLENOID VALVES -MAINSTEAM OUT ISOLATION VALVE46 IV 38 92 MQ r-O RAT-i W 38 02 ie oporated lecally '.ia the VFDR -0604 RASH'JT-DOW.N COOLINGC OUTLET hQAd ":Oh..' to a SDC cuo':fonOUTSIDE ISOL.~TION VAOLVe fil'ath from t.he RCS ton ac.mplishINMP1, April 2013Page G-12 I Constellation Enerav Nuclear GroUDAttachment G -Recoverv Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)Fire Component Component Description Actions VFDR RAIPCSAreadocay heat remoa4-.6 PB-BB11 125VDC BATTERY BOARD 11 MG 167 motor BKR-MG167(BB11/E03) VFDR-06-017 RAverified tripped for battery loadshedding to extend battery capability.6 PB-BB12 125VDC BATTERY BOARD 12 MG 167 motor BKR-MG167(BB12/F03) VFDR-06-017 RAverified tripped for battery loadshedding to extend battery capability.6 PMP-79.1-01 EMERGENCY DIESEL PMP-79.1-01 shutdown locally for VFDR-06-017 RAGENERATOR 102 TURBO LUBE battery load shedding to extend batteryOIL PUMP capability.6 PMP-79.1-07 EMERGENCY DIESEL PMP-79.1-07 shutdown locally for VFDR-06-017 RAGENERATOR 102 CIRCULATING battery load shedding to extend batteryLUBE OIL PUMP (6 GALLONS capability.PER MINUTE)6 PMP-79.1-20 EMERGENCY DIESEL PMP-79.1-20 shutdown locally for VFDR-06-017 RAGENERATOR 103 TURBO LUBE battery load shedding to extend batteryOIL PUMP capability.6 PMP-79.1-26 EDG 103 CIRCULATING LUBE PMP-79.1-26 shutdown locally for VFDR-06-017 RAOIL PUMP ( 6 GPM) battery load shedding to extend batterycapability.6 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-06-019 RATURBINE FIRE PUMP run as needed to supply fire water toEmergency Condenser Makeup tanks.6 UPS-UPS162A POWER SUPPLY -UPS 162A switches HDS-UPS162A- VFDR-06-017 RAUNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery loadSUPPLY shedding to extend battery capability.6 UPS-UPS162B POWER SUPPLY -UPS 162B switches HDS-UPS162B- VFDR-06-017 RAUNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery loadSUPPLY shedding to extend battery capability.NMPI, AprIl 2013 Page G-13 IINMPI, April 2013Page G-13 I Constellation Enerav Nuclear GrouwAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)FireArea Component Component Description Actions VFDR R.APCS6 UPS-UPS172A POWER SUPPLY -UPS 172A switches HDS-UPS172A- VFDR-06-017 RAUNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery loadSUPPLY shedding to extend battery capability.6 UPS-UPS172B POWER SUPPLY -UPS 172B switches HDS-UPS172B- VFDR-06-017 RAUNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery loadSUPPLY shedding to extend battery capability.6 VLV-05-31 LOOP 11 VENT LINE MANUAL VLV-05-31 is operated locally via the VFDR-06-004 RAISOLATION VALVE DOWN hand wheel to isolate an inventory loss VFDR-06-006STREAM OF 05-01R AND 05-11 flow path from the ECs 111 & 112 RCSsteam supply vent line.6 VLV-05-32 LOOP 12 VENT LINE MANUAL VLV-05-32 is operated locally via the VFDR-06-003 RAISOLATION VALVE hand wheel to isolate an inventory loss VFDR-06-005DOWNSTREAM OF 05-04R AND flow path from the ECs 121 & 122 RCS05-12 steam supply vent line.6 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-1 1 is locally throttled to control VFDR-06-011 RAEMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide aRCS heat sink for decay heat removal.6 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-06-011 RAEMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide aRCS heat sink for decay heat removal.7 BAT-B1 1 125 VOLT DC STATION Portable charger with a generator VFDR-07-012 RABATTERY 11 connected directly to Battery Board PB-BB11 to charge Battery B131.7 BAT-B12 125 VOLT DC STATION Portable Charger with Generator VFDR-07-012 RABATTERY 12 connected directly to Battery Board PB-BB12 to charge Battery B12.7 BV-113-3091 BLOCKING VALVE FOR AIR Operator unlocks and closes BV-113- VFDR-07-013 RASUPPLY TO SCRAM AIR 3091. Operator removes the vent pipeSYSTEM cap, unlocks, and opens VLV-113-VLV-1 13-230 BALL VALVE -SCRAM VALVE 230 to vent the SCRAM air header toPILOT HEADER VENT insert the control rods.NMPI, April 2013 Page G-14 IINMP1, April 2013Page G-14 I Constellation Enerav Nuclear GrouDAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)Fire Component Component Description Actions VFDR RA/PCSArea7 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-07-014 RABV-100-69 & 122 MANUAL VALVES opened to periodically refill theEmergency Condenser Makeup tanks.7 -V.70-63 44" AIR OERATED BLOCKIPNG I-V 70 53 is epwted loCally to VFDR-07 9-- RA~Is^,I~ ReCLC INLET. V'ALVE^= \ en .t... the ^ DC pU,,mpc and a"TO O STD Al COOL h101at11 -nk for tho CDC FzV- 3. 4. OF" OPPRAXTED FCV 38 410 *c op.-atod lc"ally to )ADR-07-04 RPA414O6 COTRL .'VE control Aoof throug SD 6C HX 38 1322 toSHUTDOWN COOLIN HIN AT rg-uto RCS G oQol doa"-. rto.EXHAGR 1 2 FLON7 IV-01-03 24" AIR OPERATED (ANGLE) Air vented manually at MSIV IV-01-03 VFDR-07-001 RAISOLATION VALVE WITH to close valve to prevent inventory loss.SOLENOID VALVES -MAINSTEAM ISOLATION VALVE 37 IV-01-04 24" AIR OPERATED (ANGLE) Air vented manually at MSIV IV-01-04 VFDR-07-002 RAISOLATION VALVE WITH to close valve to prevent inventory loss.SOLENOID VALVES -MAINSTEAM OUT ISOLATION VALVE47 PB-BB11 125VDC BATTERY BOARD 11 MG 167 motor BKR-MG167(BB11/E03) VFDR-07-012 RAverified tripped for battery loadshedding to extend battery capability.7 PB-BB12 125VDC BATTERY BOARD 12 MG 167 motor BKR-MG167(BB12/F03) VFDR-07-012 RAverified tripped for battery loadshedding to extend battery capability.PMP-38-71. SH1JTDOW.AN COO'ING PUMP 42 SDC pump PMP 33 1.2 io op.ratod at VRADR-07-0O RAAll12A PB6 176, BKR (47 .'00eA)52 toR~A&AM& a 20C fle~ath to....amplih d..ay heat ..me.al.7 PMP-79.1-01 EMERGENCY DIESEL PMP-79.1-01 shutdown locally for VFDR-07-012 RAGENERATOR 102 TURBO LUBE battery load shedding to extend batteryOIL PUMP capability.NMPI, AprIl 2013 Page G-1 5 IINMP1, April 2013Page G-1 5 1 Constellation Enerav Nuclear GrouDAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)Fire Component Component Description Actions VFDR RA/PCSArea7 PMP-79.1-07 EMERGENCY DIESEL PMP-79.1-07 shutdown locally for VFDR-07-012 RAGENERATOR 102 CIRCULATING battery load shedding to extend batteryLUBE OIL PUMP (6 GALLONS capability.PER MINUTE)7 PMP-79.1-20 EMERGENCY DIESEL PMP-79.1-20 shutdown locally for VFDR-07-012 RAGENERATOR 103 TURBO LUBE battery load shedding to extend batteryOIL PUMP capability.7 PMP-79.1-26 EDG 103 CIRCULATING LUBE PMP-79.1-26 shutdown locally for VFDR-07-012 RAOIL PUMP ( 6 GPM) battery load shedding to extend batterycapability.7 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-07-014 RATURBINE FIRE PUMP run as needed to supply fire water toEmergency Condenser Makeup tanks.7 UPS-UPS162A POWER SUPPLY -UPS 162A switches HDS-UPS162A- VFDR-07-012 RAUNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery loadSUPPLY shedding to extend battery capability.7 UPS-UPS162B POWER SUPPLY -UPS 162B switches HDS-UPS162B- VFDR-07-012 RAUNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery loadSUPPLY shedding to extend battery capability.7 UPS-UPS172A POWER SUPPLY -UPS 172A switches HDS-UPS172A- VFDR-07-012 RAUNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery loadSUPPLY shedding to extend battery capability.7 UPS-UPS172B POWER SUPPLY -UPS 172B switches HDS-UPS172B- VFDR-07-012 RAUNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery loadSUPPLY shedding to extend battery capability.7 VLV-60-1 I MAKEUP VALVE TO LOOP 12 VLV-60-11 is locally throttled to control VFDR-07-007 RAEMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide aRCS heat sink for decay heat removal.7 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-07-007 RAEMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide aRCS heat sink for decay heat removal.NMPI, April2013 Page G-16 IINMP1, April 2013Page G-16 I Constellation Energy Nuclear GroupAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)FireArea Component Component Description Actions VFDR RA/PCS9 BAT-B11 125 VOLT DC STATION Portable charger with a generator VFDR-09-018 RABATTERY 11 connected directly to Battery Board PB-BB11 to charge Battery B11.9 BAT-B12 125 VOLT DC STATION Portable Charger with Generator VFDR-09-018 RABATTERY 12 connected directly to Battery Board PB-BB12 to charge Battery B12.9 BKR-(102/2-1) DIESEL GEN 102 OUTPUT Operate disconnect switch locally to VFDR-09-019 RAR1022/571 BREAKER 2-1(R1022/571) to allow for recovery of diesel generator.PB1029 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-09-022 RABV-100-69 & 122 MANUAL VALVES opened to periodically refill theEmergency Condenser Makeup tanks.9 6V r-63 11" AIR 0-OPERATED BLOCKING 6V 7. 65 Is oprated casally to preode VFDR-O9 013 RAVALVEG RB-QCLC INILET VALVE seeling water to the 60C pumnps and aTSHTOWN COOL ING, hcAt Sink for the SDC HXc.S FCV4 29 an A" D'-A.PR.A.GM OPERATED FCV 38 00 is oporat-d Wocally to VIZ-309 015 RAFLW OTROL VALVE soto e floa'w -AA t-hre-ug h DC GHX 3 IS2 16% toSHU1TDOWN COOL--1INGC HET reoulato RCS cool- doa-.i rato.a FGV44 8" OPER-ATE-D FC-V 38 11 Is oporated locally to VFDR 09 04 4 RAFLOW C TROL VI V9ILV clWnl- Ilow through 6DG HX 3l 420 toSHUTODOWN COOL31ING HET rOgulato RCS coo! diNANi ratoEX.CHA=G.Ar-ER 13 FWLMO^-r ------- VAI Vs9 FCV-39-15 EMERGENCY CONDENSER FCV-39-15 operated locally via the VFDR-09-002 RASTEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory lossCONTROL VALVE -FLOW flow path from the ECs 111 & 112 RCSCONTROL VALVE 11 STEAM return path drain line.LINE DRAIN9 FCV-39-16 EMERGENCY CONDENSER FCV-39-16 operated locally via the VFDR-09-003 RASTEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory lossCONTROL VALVE -FLOW flow path from the ECs 121 & 122 RCSCONTROL VALVE 12 STEAM return path drain line.LINE DRAINlNMPI, April 2013Page G-17 I Constellation Energy Nuclear GroupAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)Fire Component Component Description Actions VFDR RA/PCSArea9 IV-01-04 24" AIR OPERATED (ANGLE) Air vented manually at MSIV IV-01-04 VFDR-09-008 RAISOLATION VALVE WITH to close valve to prevent inventory loss.SOLENOID VALVES -MAINSTEAM OUT ISOLATION VALVE4P IV-38B1 12OOR OPERATED IV 39 02 us moporBKatMd loGally G Ba1E VFDR-09-021 RASHTD I COO 1-ýING_ O-UTL E-T hand- boohe to cc-t-ablugch a 2SDC_ AcuctionOU TS IDE. ISLTI CALVE fiedpath roem tho RS teo alompalhshdeday that erxn bmatyal.9 PB-BB1 1 125VDC BATTERY BOARD 11 MG 167 motor BKR-MG167(BB 12/E03) VFDR-09-018 RAverified tripped for battery loadshedding to extend battery capability.9 PB-BB12 125VDC BATTERY BOARD 12 MG 167 motor BKR-MG167(BB129FO3) VFDR-09-018 RAverified tripped for battery loadshedding to extend battery capability.9 PMP-79.1-01 EMERGENCY DIESEL PMP-79.1-01 shutdown locally for VFDR-09-018 RAGENERATOR 102 TURBO LUBE battery load shedding to extend batteryOIL PUMP capability.9 PMP-79.1-07 EMERGENCY DIESEL PMP-79.1-07 shutdown locally for VFDR-09-018 RAGENERATOR 102 CIRCULATING battery load shedding to extend batteryLUBE OIL PUMP (6 GALLONS capability.PER MINUTE) ______________ _____9 PMP-79.1-20 EMERGENCY DIESEL PMP-79.1-20 shutdown locally for VFDR-09-018 RAGENERATOR 103 TURBO LUBE battery load shedding to extend batteryOIL PUMP capability.9 PMP-79.1-26 EDG 103 CIRCULATING LUBE PMP-79.1-26 shutdown locally for VFDR-09-018 RAOIL PUMP ( 6 GPM) battery load shedding to extend batterycapability.9 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-09-022 RATURBINE FIRE PUMP run as needed to supply fire water toEmergency Condenser Makeup tanks.NMPI, April 2013 Page G-18 IiNMPI, April 2013Page G-18 I Constellation Enemy Nuclear GroupAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)Fire Component Component Description Actions VFDR RAIPCSArea9 UPS-UPS162A POWER SUPPLY -UPS 162A switches HDS-UPS162A- VFDR-09-018 RAUNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery loadSUPPLY shedding to extend battery capability.9 UPS-UPS162B POWER SUPPLY -UPS 162B switches HDS-UPS162B- VFDR-09-018 RAUNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery loadSUPPLY shedding to extend battery capability.9 UPS-UPS172A POWER SUPPLY -UPS 172A switches HDS-UPS172A- VFDR-09-018 RAUNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery loadSUPPLY shedding to extend battery capability.9 UPS-UPS172B POWER SUPPLY -UPS 172B switches HDS-UPS172B- VFDR-09-018 RAUNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery loadSUPPLY shedding to extend battery capability.9 VLV-05-31 LOOP 11 VENT LINE MANUAL VLV-05-31 is operated locally via the VFDR-09-005 RAISOLATION VALVE DOWN hand wheel to isolate an inventory loss VFDR-09-007STREAM OF 05-01R AND 05-11 flow path from the ECs 111 & 112 RCSsteam supply vent line.9 VLV-05-32 LOOP 12 VENT LINE MANUAL VLV-05-32 is operated locally via the VFDR-09-004 RAISOLATION VALVE hand wheel to isolate an inventory loss VFDR-09-006DOWNSTREAM OF 05-04R AND flow path from the ECs 121 & 122 RCS05-12 steam supply vent line.9 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-11 is locally throttled to control VFDR-09-011 RAEMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide aRCS heat sink for decay heat removal.9 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-09-011 RAEMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide aRCS heat sink for decay heat removal.10 BAT-B11 125 VOLT DC STATION Portable charger with a generator VFDR-10-014 RABATTERY 11 connected directly to Battery Board PB-BB11 to charge Battery B131.10 BAT-B12 125 VOLT DC STATION Portable Charger with Generator VFDR-10-014 RABATTERY 12 connected directly to Battery Board PB-BB12 to charge Battery B12.INMP1, April 2013Page G-19 I Constellation Energy Nuclear GroupAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)FireArea Component Component Description Actions VFDR RA/PCS10 BKR-(102/2-1) DIESEL GEN 102 OUTPUT Operate disconnect switch locally to VFDR-10-021 RAR1022/571 BREAKER 2-1(R1 022/571) TO allow for recovery of diesel generator.PB10210 BV-1 13-3091 BLOCKING VALVE FOR AIR Operator unlocks and closes BV-1 13- VFDR-10-028 RASUPPLY TO SCRAM AIR 3091. Operator removes the vent pipeSYSTEM cap, unlocks, and opens VLV-1 13-VLV-1 13-230 BALL VALVE -SCRAM VALVE 230 to vent the SCRAM air header toPILOT HEADER VENT insert the control rods.10 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-10-029 RABV-100-69 & 122 MANUAL VALVES opened to periodically refill theEmergency Condenser Makeup tanks.40 BV40-53 14" AI R OPERATED BLOCKING BE 70 53 i6 0p-t GK. to proyide VFDR-40-949 RAVAi\/E RQQCLCINLETVALVE cooling Wa!F 11o tho SDC pumps and aTOSHTDON COOLING heat WAI; for thA 20C HXc.SYSTEM40 FGV 38 09 A" DIAPHWRAG" O1PERATED FCV 38 09 is oporated locsally to VFQR 10 020 RFLOWCONTOL VLVEcanal flow throu gh SDC HX 38 135 toRWI I-MAN QQ' IG Wrcgulate RCS cool devm Fate.EXCH A-NGER 11 FrLOWCONTIC3 VA '10 FCV-39-15 EMERGENCY CONDENSER FCV-39-15 operated locally via the VFDR-10-007 RASTEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory lossCONTROL VALVE -FLOW flow path from the ECs 111 & 112 RCSCONTROL VALVE 11 STEAM return path drain line.LINE DRAIN10 FCV-39-16 EMERGENCY CONDENSER FCV-39-16 operated locally via the VFDR-10-008 RASTEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory lossCONTROL VALVE -FLOW flow path from the ECs 121 & 122 RCSCONTROL VALVE 12 STEAM return path drain line.LINE DRAIN10 IV-01-03 24" AIR OPERATED (ANGLE) Air vented manually at MSIV IV-01-03 VFDR-10-009 RAISOLATION VALVE WITH to close valve to prevent inventory loss.SOLENOID VALVES -MAINSTEAM ISOLATION VALVE 310 IV-01-04 24" AIR OPERATED (ANGLE) Air vented manually at MSIV IV-01-04 VFDR-10-010 RAISOLATION VALVE WITH to close valve to prevent inventory loss.SOLENOID VALVES -MAINSTEAM OUT ISOLATION VALVE4INMPI, April 2013Page G-20 I Constellation Enerav Nuclear GrouDAttachment G -Recoverv Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)FireArea Component Component Description Actions VFDR RA/PCS40 IV-38 01 MOT-OR OPERATE[) AtP 67, 8KR (167!D03)52, the VFQR 10 0-11 RASHUTDOWN CONOL C NG1 IS I OUTLET .9 ...ale w.ring ic .opairad and the .a...INSIDE ISOLATION VALVE- oporatod to farsilitato establiching aSDc cucfion flow path from the RCS toacoosmplich docay heat romov.al.40 IV-39-02 0MOTOR OPERATED IV 38 02 i patd lnally o ia thb VFDR-10-012 RASH1-DOWNE COO- eINGtiT hand fAoml to eatablish a aOC cuotionOUTSE ISOLATION VALVE f(ecath fram the r RCoSa to atsinEChdewBy-heat rco&1 al.10 IV-39-05 EMERGENCY COOLING LOOP IV-39-05 is manually opened by VFDR-10-012 RA11 CONDENSATE RETURN AIR venting air from the valve to establish aOPERATED ISOLATION VALVE ( decay heat removal path using ECsGLOBE) 11 & 112.10 IV-39-06 EMERGENCY COOLING LOOP IV-39-06 is manually opened by VFDR-10-012 RA12 CONDENSATE RETURN AIR venting air from the valve to establish aOPERATED ISOLATION VALVE ( decay heat removal path using ECsGLOBE) 121 & 122.10 IV-39-07R MOTOR OPERATED LOOP 11 IV-39-07R is operated locally via the VFDR-10-012 RASTEAM OUTLET OUTSIDE hand wheel to establish a decay heatISOLATION VALVE 112 removal flowpath using ECs 111 & 112.10 IV-39-08R MOTOR OPERATED LOOP 12 IV-39-08R is operated locally via the VFDR-10-012 RASTEAM OUTLET OUTSIDE hand wheel to establish a decay heatISOLATION VALVE 122 removal flowpath using ECs 121 & 122.10 IV-39-09R MOTOR OPERATED LOOP 11 IV-39-09R is operated locally via the VFDR-10-012 RASTEAM OUTLET INSIDE hand wheel to establish a decay heatISOLATION VALVE 111 removal flowpath using ECs 111 & 112.10 IV-39-IOR MOTOR OPERATED LOOP 12 IV-39-I1OR is operated locally via the VFDR-10-012 RASTEAM OUTLET INSIDE hand wheel to establish a decay heatISOLATION VALVE 121 removal flowpath using ECs 121 & 122.10 LCV-60-17 EMERGENCY CONDENSER 111 Place EC 111/112 Level Control VFDR-10-012 RA-112 LEVEL CONTROL VALVE ( Transfer switch to Local and verify AutoLOOP 11 ) -AIR ACTUATED control by observing "A" on statusFAIL OPEN panel at PNL-RSP11 to override falseEC high level signal to support decayheat removal via EC's 111 & 112.INMPI, April 2013Page G-21 I Constellation Energy Nuclear GroupAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)Fire Component Component Description Actions VFDR RA/PCSArea10 LCV-60-18 EMERGENCY CONDENSER 121 Place EC 121/122 Level Control VFDR-10-012 RA-122 LEVEL CONTROL VALVE ( Transfer switch to Local and verify AutoLOOP 12 ) -AIR ACTUATED control by observing "A" on statusFAIL OPEN panel at PNL-RSP12 to override falseEC high level signal to support decayheat removal via EC's 121 & 122.10 LI-36-28 REACTOR VESSEL LEVEL Operator determines vital parameters VFDR-10-013 RAPI-36-25 REACTOR VESSEL PRESSURE from PNL-RSP11.10 PB-BB11 125VDC BATTERY BOARD 11 MG 167 motor BKR-MG167(BB1I/E03) VFDR-10-014 RAverified tripped for battery loadshedding to extend battery capability.10 PB-BB12 125VDC BATTERY BOARD 12 MG 167 motor BKR-MG167(BB12/F03) VFDR-10-014 RAverified tripped for battery loadshedding to extend battery capability.40 PMP 38 4,9 SHU....TD ......N C LING PUMP SDC pIPA8_- 149 110'rng ropai:rd and VF.R...026RNUO2A oporatod locally at PB3 lOB BKR(I16B/0OOA)52 to cae c a 2SDCMfie~'ath to accomplich decay hoat10 PMP-79.1-01 EMERGENCY DIESEL PMP-79.1-01 shutdown locally for VFDR-10-014 RAGENERATOR 102 TURBO LUBE battery load shedding to extend batteryOIL PUMP capability.10 PMP-79.1-07 EMERGENCY DIESEL PMP-79.1-07 shutdown locally for VFDR-10-014 RAGENERATOR 102 CIRCULATING battery load shedding to extend batteryLUBE OIL PUMP (6 GALLONS capability.PER MINUTE)10 PMP-79.1-20 EMERGENCY DIESEL PMP-79.1-20 shutdown locally for VFDR-10-014 RAGENERATOR 103 TURBO LUBE battery load shedding to extend batteryOIL PUMP capability.10 PMP-79.1-26 EDG 103 CIRCULATING LUBE PMP-79.1-26 shutdown locally for VFDR-10-014 RAOIL PUMP ( 6 GPM) battery load shedding to extend batterycapability.NMPI, April 2013 Page G-22 IINMP1, April 2013Page G-22 I Constellation Energy Nuclear GroupAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)FireArea Component Component Description Actions VFDR RA/PCS10 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-10-029 RATURBINE FIRE PUMP run as needed to supply fire water toEmergency Condenser Makeup tanks.10 UPS-UPS162A POWER SUPPLY -UPS 162A switches HDS-UPS162A- VFDR-10-014 RAUNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery loadSUPPLY shedding to extend battery capability.10 UPS-UPS162B POWER SUPPLY -UPS 162B switches HDS-UPS162B- VFDR-10-014 RAUNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery loadSUPPLY shedding to extend battery capability.10 UPS-UPS172A POWER SUPPLY -UPS 172A switches HDS-UPS172A- VFDR-10-014 RAUNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery loadSUPPLY shedding to extend battery capability.10 UPS-UPS172B POWER SUPPLY -UPS 172B switches HDS-UPS172B- VFDR-10-014 RAUNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery loadSUPPLY shedding to extend battery capability.10 VLV-05-31 LOOP 11 VENT LINE MANUAL VLV-05-31 is operated locally via the VFDR-10-005 RAISOLATION VALVE DOWN hand wheel to isolate an inventory loss VFDR-10-006STREAM OF 05-01R AND 05-11 flow path from the ECs 111 & 112 RCSsteam supply vent line.10 VLV-05-32 LOOP 12 VENT LINE MANUAL VLV-05-32 is operated locally via the VFDR-10-003 RAISOLATION VALVE hand wheel to isolate an inventory loss VFDR-10-004DOWNSTREAM OF 05-04R AND flow path from the ECs 121 & 122 RCS05-12 steam supply vent line.10 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-1 1 is locally throttled to control VFDR-10-017 RAEMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide aRCS heat sink for decay heat removal.10 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-10-017 RAEMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide aRCS heat sink for decay heat removal.11 BAT-B11 125 VOLT DC STATION Portable charger with a generator VFDR-1 1-020 RABATTERY NUMBER 11 connected directly to Battery Board PB- VFDR-1 1-030BB11 to charge Battery Bl1.NMPI, AprIl 2013 Page G-23 IINMPI, April 2013Page G-23 I Constellation Enerav Nuclear GrouoAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)FireArea Component Component Description Actions VFDR RAIPCS11 BAT-B 12 125 VOLT DC STATION Portable Charger with Generator VFDR-1 1-020 RABATTERY NUMBER 12 connected directly to Battery Board PB- VFDR-1 1-030BB12 to charge Battery B12.11 BV-1 13-3091 BLOCKING VALVE FOR AIR Operator unlocks and doses BV-1 13- VFDR-1 1-036 RASUPPLY TO SCRAM AIR 3091. Operator removes the vent pipeSYSTEM cap, unlocks, and opens VLV-113-VLV-113-230 BALL VALVE -SCRAM VALVE 230 to vent the SCRAM air header toPILOT HEADER VENT insert the control rods.11 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-11-037 RABV-100-69 & 122 MANUAL VALVES opened to periodically refill theEmergency Condenser Makeup tanks.14 BV-38-04 SuTDO, COOLING PUMP 12 2 8V 38 04 *c oporat.d locally to align VFDR- 4 --03 RAS'JCIONBLOKIN VAVE DC pump PMP 38 15 42 to the RCS to!.:ti SDC to pro-"ido dc::,"ay hcat44- A-70 A- ^4-IR OPE^RATED LO=CKING BV 70 63 is epcmt-d locally to p.o.ido VFrDR 4- 04 RAVAL^ VE RoEQCL- INLET V'ALVE, , OlG .O.. +^. to the SOD pumps and aTOQ SHUlTDOWN COOLING hcat aink for tha SADC HXc.11 EMERGENCY EMERGENCY COOLING Place Emergency Cooling Isolation N/A PCSCOOLING ISOLATION BYPASS SWITCH Bypass Switch in bypass to enableISOLATION operation of IV-39-05, IV-39-07, andBYPASS IV-39-09 from PNL-RSP1 1.SWITCH11 EMERGENCY EMERGENCY COOLING Place Emergency Cooling Isolation N/A PCSCOOLING ISOLATION BYPASS SWITCH Bypass Switch in bypass to enableISOLATION operation of IV-39-06, IV-39-08, andBYPASS IV-39-10 from PNL-RSP12.SWITCH4-1- FCA29 IQ DA-".W-PHRA .':.-.OPET -ED FOV 38 410 ic opoated locally to VFDR-14-036 RAFLWCNROL LA.L44 A acontra! flow 41th ra- ugh 6 DC HX 3 8 13- 2 toSHUl TDOWNI COOL ING HET rogulato RC QcGol damp rat.EXHAGR 12 FLOWCOTO UAI ý11 FCV-39-15 EMERGENCY CONDENSER FCV-39-15 operated locally via the VFDR-1 1-009 RASTEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory lossCONTROL VALVE -FLOW flow path from the ECs 111 & 112 RCSCONTROL VALVE 11 STEAM return path drain line.LINE DRAININMPI, April 2013Page G-24 I Constellation Enerav Nuclear GrouDAttachment G -Recoverv Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)FireArea Component Component Description Actions VFDR RA/PCS11 FCV-39-16 EMERGENCY CONDENSER FCV-39-16 operated locally via the VFDR-11-010 RASTEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory lossCONTROL VALVE -FLOW flow path from the ECs 121 & 122 RCSCONTROL VALVE 12 STEAM return path drain line.LINE DRAIN11 IV-01-03 24" AIR OPERATED ( ANGLE) Air vented manually at MSIV IV-01-03 VFDR-1 1-003 RAISOLATION VALVE WITH to close valve to prevent inventory loss.SOLENOID VALVES -MAINSTEAM ISOLATION VALVE 311 IV-01-04 24" AIR OPERATED (ANGLE) Air vented manually at MSlV IV-01-04 VFDR-11-004 RAISOLATION VALVE WITH to close valve to prevent inventory loss.SOLENOID VALVES -MAINSTEAM OUT ISOLATION VALVE414- IV,3849 MOTOR OPERATED At PB 167, BKR (1674D03)52, the QR RASHUDON OOLING OUTL ET valve YAFRig is ropairod and the valveINID SOATION VAL VE oporatcd to facilitate establishing aSDC 13cQuc-,..tien fo' path frome the RCS toaccompslih doray heat rFemea-.14- ,V-38 02 MOTOR OPERATED IV 38 02 ic epcrated lacally "ia the VFDR 11 035 RASHUTDOWN COOLING OUTLET. hand -Ae' l to " stabr hlic-h a SD Cri suc t 6ionOUIT-SIDE ISA-TION VAL VE f .wpath from the RCS to ah C Splith1CNENAERTUNAR dechay heat reMoaapa14 IV 38 i3 REACT-OR SHUTDOWN At PB 167,13141 (167!G03)52, the VFDR-4-4-O29 RAOONRETURN ISIOAATIO N Nalve ALVEg i trpaiud and the valveVAI Vr= operated to facilitate ectablishing a8DC diochargo flow path to the RCS toGacOBEamplish dersay heat rcmaval.11 IV-39-05 EMERGENCY COOLING LOOP IV-39-05 operated from PNL-RSP1 1 to N/A PCs11 CONDENSATE RETURN AIR establish a decay heat removal pathOPERATED ISOLATION VALVE ( through ECs 111 & 112.GLOBE )11 IV-39-06 EMERGENCY COOLING LOOP IV-39-06 operated from PNL-RSP12 to N/A PCS12 CONDENSATE RETURN AIR establish a decay heat removal pathOPERATED ISOLATION VALVE ( through ECs 121 & 122.___________GLOBE)11 IV-39-07R MOTOR OPERATED LOOP 11 IV-39-07R eperated from PNL-RSP1 1 N/A PCsSTEAM OUTLET OUTSIDE to establish a decay heat removal pathISOLATION VALVE 112 through ECs 111 & 112.INMPI, April 2013Page G-25 I Constellation Energy Nuclear GroupAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)FireArea Component Component Description Actions VFDR RA/PCS11 IV-39-08R MOTOR OPERATED LOOP 12 IV-39-08R operated from PNL-RSP12 N/A PCSSTEAM OUTLET OUTSIDE to establish a decay heat removal pathISOLATION VALVE 122 through ECs 121 & 122.11 IV-39-09R MOTOR OPERATED LOOP 11 IV-39-09R operated from PNL-RSP1 1 N/A PCSSTEAM OUTLET INSIDE to establish a decay heat removal pathISOLATION VALVE 111 through ECs 111 & 112.11 IV-39-1OR MOTOR OPERATED LOOP 12 IV-39-1OR operated from PNL-RSP12 N/A PCSSTEAM OUTLET INSIDE to establish a decay heat removal pathISOLATION VALVE 121 through ECs 121 & 122.11 LCV-60-17 EMERGENCY CONDENSER 111 Place EC 111/112 Level Control N/A PCS-112 LEVEL CONTROL VALVE ( Transfer switch to Local and verify AutoLOOP 11 ) -AIR ACTUATED control by observing "A" on statusFAIL OPEN panel at PNL-RSP11.11 LCV-60-18 EMERGENCY CONDENSER 121 Place EC 121/122 Level Control N/A PCS-122 LEVEL CONTROL VALVE ( Transfer switch to Local and verify AutoLOOP 12 ) -AIR ACTUATED control by observing "A" on statusFAIL OPEN panel at PNL-RSP12.11 LI-36-26 REACTOR VESSEL LEVEL Operator determines vital parameters N/A PCSLI-60-23C EMERGENCY CONDENSER 121 from PNL-RSP12.& 122PI-201.2-94 DRYWELL PRESSUREPI-36-27 REACTOR VESSEL PRESSURETI-201-51B DRYWELL TEMPERATURETI-201.2-522B TORUS TEMPERATURETI-32-04B REACTOR COOLANTTEMPERATURETI-32-05B REACTOR COOLANTTEMPERATURE11 MG-MG131 MG SET 131 Place MG Set #131 switch in the TRIP N/A PCSposition and confirm CONTROL RODSIN white light lit on PNL-RSP11.11 MG-MG141 MG SET 141 Place MG Set #141 switch in the TRIP N/A PCSposition and confirm CONTROL RODSIN white light lit on PNL-RSP12.NMPI, April 2013 Page G-26 I!NMPI, April 2013Page G-26 I Constellation Eneray Nuclear GroupAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)FireArea Component Component Description Actions VFDR RAIPCS11 PB-BB11 125VDC BATTERY BOARD 11 MG 167 motor BKR-MG167(BB11/E03) VFDR-11-020 RAverified tripped for battery load VFDR-1 1-030shedding to extend battery capability.11 PB-BB12 125VDC BATTERY BOARD 12 MG 167 motor BKR-MG167(BB12/F03) VFDR-11-020 RAverified tripped for battery load VFDR-1 1-030shedding to extend battery capability.44 PMP-38-165 SHUJTDOWNA!.h COOL3NG PUMP 12 SDC pump PMP 38 162 Is opo.ated at V-FDR4 042 RAISU 122 P29 476, BKfl (172E'00eA)92 toastabl-ah 2 29C fleovath to_______________________acwmpliish docay heat Faemoval.11 PMP-79.1-01 EMERGENCY DIESEL PMP-79.1-01 shutdown locally for VFDR-1 1-020 RAGENERATOR 102 TURBO LUBE battery load shedding to extend battery VFDR-1 1-030OIL PUMP capability.11 PMP-79.1-07 EMERGENCY DIESEL PMP-79.1-07 shutdown locally for VFDR-1 1-020 RAGENERATOR 102 CIRCULATING battery load shedding to extend battery VFDR-1 1-030LUBE OIL PUMP (6 GALLONS capability.PER MINUTE)11 PMP-79.1-20 EMERGENCY DIESEL PMP-79.1-20 shutdown locally for VFDR-11-020 RAGENERATOR 103 TURBO LUBE battery load shedding to extend battery VFDR-1 1-030OIL PUMP capability.11 PMP-79.1-26 EDG 103 CIRCULATING LUBE PMP-79.1-26 shutdown locally for VFDR-11-020 RAOIL PUMP (6 GPM) battery load shedding to extend battery VFDR-1 1-030capability.11 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-1 1-037 RATURBINE FIRE PUMP run as needed to supply fire water toEmergency Condenser Makeup tanks.11 UPS-UPS162A POWER SUPPLY- UPS 162A switches HDS-UPS162A- VFDR-11-020 RAUNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load VFDR-1 1-030SUPPLY shedding to extend battery capability.11 UPS-UPS162B POWER SUPPLY -UPS 162B switches HDS-UPS162B- VFDR-11-020 RAUNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load VFDR-1 1-030SUPPLY shedding to extend battery capability.INMPI, April 2013Page G-27 I Constellation Enerav Nuclear GroueAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)Fire Component Component Description Actions VFDR RAIPCSArea11 UPS-UPS172A POWER SUPPLY- UPS 172A switches HDS-UPS172A- VFDR-11-020 RAUNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load VFDR-1 1-030SUPPLY shedding to extend battery capability.11 UPS-UPS172B POWER SUPPLY -UPS 172B switches HDS-UPS172B- VFDR-11-020 RAUNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load VFDR-1 1-030SUPPLY shedding to extend battery capability.11 VLV-05-31 LOOP 11 VENT LINE MANUAL VLV-05-31 is operated locally via the VFDR-1 1-005 RAISOLATION VALVE DOWN hand wheel to isolate an inventory loss VFDR-1 1-007STREAM OF 05-01R AND 05-11 flow path from the ECs 111 & 112 RCSsteam supply vent line.11 VLV-05-32 LOOP 12 VENT LINE MANUAL VLV-05-32 is operated locally via the VFDR-1 1-006 RAISOLATION VALVE hand wheel to isolate an inventory loss VFDR-1 1-008DOWNSTREAM OF 05-04R AND flow path from the ECs 121 & 122 RCS05-12 steam supply vent line.11 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-11 is locally throttled to control VFDR-11-015 RAEMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide aRCS heat sink for decay heat removal.11 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-11-015 RAEMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide aRCS heat sink for decay heat removal.11 VLV-93-13 12" GATE VALVE -121 VLV-93-13 dosed locally via the hand VFDR-11-018 RACONTAINMENT SPRAY RAW wheel to isolate CTSRW flow toWATER PUMP DISCHARGE Containment Spray Header #11.VALVE11 VLV-93-16 12" GATE VALVE -112 VLV-93-16 closed locally via the hand VFDR-11-017 RACONTAINMENT SPRAY RAW wheel to isolate CTSRW flow toWATER PUMP DISCHARGE Containment Spray Header #12.VALVE12 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-12-009 RABV-100-69 & 122 MANUAL VALVES opened to periodically refill theEmergency Condenser Makeup tanks.412 BV-A) 63 4A11141I OPE-6-R-ATED- -RLOCGK!NG BY70 53 is oporatod loall opo.d VFDR.42 006 RAV.ALV4 R ,CL INCLET VAIVE 0 .at.. to the SDC pumps and aTo SHUTD O he at sonk for tho SDG WgsSYST-rmINMPI, April 2013Page G-28 I Constellation Enerav Nuclear GrouDAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)FireArea Component Component Description Actions VFDR RAIPCS42 FGV A" DIAPHRAGM OPERATED FCV 38 00 is op"rat.d locally to Vr-R-4DI042 RAFLOW ~ ~ ~ -eA&- COTOfiLEc ntrolfowthrugh 6DC HX 38 125 toHuTD-Oil COOLING HET r^gulate RCS cool do.- Fate.EXCHANG R 11 FL ON42 PMPA-102A DIELAPHRA OPERATED FDV 38 410 is mporanud loally to a nFDR-42-O09 RAFLORBONITROL VALU E cntal flow thosuppgh 6C rX 38 132 to12SHV60-KUTWPNAL COO' ING 2 LV Eeguk-t1 RiS soel dly Frote.EXHAGR 12 FLOWRCShet in fr ecy ea rmoal42 GVLV-s 80- " DMAKPHRVALV PERATED FCV 38 11 is ope1ated lcally to VFDR-42-3O- RAFLOWGCNTR VALOVE9S aoentro foEsw thrugh 69G HX 38 p2d toH1TDOWRN COOLING HEAT rOgulato RCS cool do" FateoEXCHANIGER 41 FLOWa6U4 _-= 4AI ýra12 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-12-009 RATURBINE FIRE PUMP run as needed to supply fire water to_________Emergency Condenser Makeup tanks.12 VLV-60-1 1 MAKEUP VALVE TO LOOP 12 VLV-60-1 1 is locally throttled to control VFDR-12-004 RAEMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide aRCS heat sink for decay heat removal.12 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-12-003 RAEMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide aRCS heat sink for decay heat removal.44 6V zo 63 :11" AR OPERATED SLO-KINGQ BV 70 53 we oporated locally to pmoýAde VFDR 13 008 PAVALE RCLCINL-ET 9ALE colig at8F to tho SDC pumps and aTO SHUTDW COLNG hat snok for tho 213C HXc.43 F= -38-00 A* DLAPHIlA.GM OPERATED FCV 38 00 Is oporatod Woally to VFDR 42-0 PAFLOW COT OL VALE QWontrol flow through SDC HX 38 1356 toHU1TDOWNN COO ING HEAT reogulto RC 2caol dmop Fate.44 FGV as io a" DL "'PA.M O1PERATD FC 39 10is oporatod locally to FR308PFLO CONTROL VAL VE contAfrol f lowA 10hreough rsDC HX 3 8 13- 2 to-SHU1_TDOWN COOLING. HEAT rOgulato RCS Gaol doein rate-EXCHWANGQER 12 FLOWMVC~II'ALVINMP1, April 2013Page G-29 I Constellation Energy Nuclear GroupAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)FireArea Component Component Description Actions VFDR RAJPCS43 FGVas -44 8" DIA-P-A.GM PA F "V 38 14 is operated W ,ally to VFWDR--42-007 RAFLW CONTROL VALVE ont-erol flvow tlhroaugh SDC HX 38 120 toC =U IAIT Q 1 ING EAI13 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-13-011 RABV-100-69 & 122 MANUAL VALVES opened to periodically refill theEmergency Condenser Makeup tanks.13 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-13-011 RATURBINE FIRE PUMP run as needed to supply fire water toEmergency Condenser Makeup tanks.13 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-1 1 is locally throttled to control VFDR-1 3-004 RAEMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide aRCS heat sink for decay heat removal.13 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-13-003 RAEMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide aRCS heat sink for decay heat removal.44 8VA)63 4"A-IR O ERATEDT BLO-KING. 8V 70 63 is opetated loWally to p-.ide VF'DR 4 00 RAVAL VE RBCLCG INLET VALVE cooling at8F to the 6DC pumps and aTO SHUTDOW COO.ING hAt. Sink for the 2DC XC..44 FGV 38 09 A" DIAPIHRArGM OPE0.RATED FCV 32-9 09 is epcated lcally to VFDR 44-006 RAFLO CO -NT-ROL VALVEcoto owtruhDCH3815oSWI-IODAW OG' NG WET rc8ulate RCS_ cool- doemn rat.EXCHAG-ER 41 FLOAW44 FGV a8s 4 8" DIAOPHRAG! M OPEsR-A.TED FCV 39 10 is operated locally to VFDR 4 RAFLOWhCOITROL VALVE con tro fle4A throUh SDG HX 38 132 toSHUTIDOWNA COOLIN HET rogulato RCS- cool doevip rateEXCANCR 12 FLOWMACOriTRO' 'AI ýra44 FCA4 2-44 8" DMI.A.HR-AG1M OPER614ATED FCV 38 11 is opc~ated locally to VFDR 144008 RAFLO-W CONTROL VALVE coent-rol flow~ t-hrough SDC HX 38 1-20 toSHU1TDOWNAgb COOLI0 G HWET rgulate RCS G-eol dkwm rael.ECHANGRO 13I FLONMPI, AprIl 2013 Page G-30 IINMP1, April 2013Page G-30 I Constellation Enerav Nuclear GrouDAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)FireArea Component Component Description Actions VFDR RA/PCS14 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-14-009 RABV-100-69 & 122 MANUAL VALVES opened to periodically refill theEmergency Condenser Makeup tanks.14 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-14-009 RATURBINE FIRE PUMP run as needed to supply fire water toEmergency Condenser Makeup tanks.14 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-1 1 is locally throttled to control VFDR-14-004 RAEMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide aRCS heat sink for decay heat removal.14 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-14-003 RAEMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide aRCS heat sink for decay heat removal.4-6 BV-74-63 44" AIR OPERATED AL OCKIN_ 8Y 70 63 is operot.d lesal!y to pro.de VFDR- 46 00 RAwAV, R, CL- INLT VALVE c..ling "r to tho 0_C puFmpAS and aTO SHUI-TDO)-WN COO0L0 IING heat sinkI forf tho 2DC HXr.4- FCV4 9 8n " DL2.2 3PHRANGAM OPERPAT-ED FCV 38 09 is -pe-ated ecsally to VF OR -- RABLO9 ONUTROL VALVES opentol pwerodicallgh 6 HX 38 136 toHUTDOWNl COO ING HET regubtge RC 9-sool denser rata.EXHNER 411 FlO"41 PGVMP-100- DI APHRAGE OPERATID FV 38 P 10 is mpratnd locally te VFDR-15-9009 RAFO C TROL VALVE ntro I lneeod tough pDC l X 38 122 to9hrgnc Condnse Makeu taksSHUDOW COLIG-EAT rogulato RCSG easol de;.o rato.EXHNER 12 FL ONCONR~O' VAI *944 FGV 39 -4 8" DLA.PHR-A.G1Q OPER-ATED FCY 38 is oporatod locally to VFDR 16 908 RAFLOWA GCONT-ROL VALVE control- flow through 69C H4X 38 1-20 leWHUTDO l OI2NQ HEA roguilat RCA cool dawn Fate,EXCHANGER 13 FLOW0h'AI 49~15 BV-100-68 LOOP I111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-15-009 RABV-100-69 & 122 MANUAL VALVES opened to periodically refill the__________Emergency Condenser Makeup tanks.15 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-15-009 RATURBINE FIRE PUMP run as needed to supply fire water to_________ ___________ ______________________Emergency Condenser Makeup tanks.NMPI, April 2013 Page G-31 IINMP1, April 2013Page G-31 I Constellation Energy Nuclear GroupAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)FireArea Component Component Description Actions VFDR RA/PCS15 VLV-80-1 1 MAKEUP VALVE TO LOOP 12 VLV-60-1 I is locally throttled to control VFDR-15-004 RAEMERGENCY CONDENSERS makeup to EC's 121 & 122 to provide aRCS heat sink for decay heat removal.15 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-15-003 RAEMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide aRCS heat sink for decay heat removal.46A RV A- ^2 44".^IR OPERATED BLOCKINGQ BV 70 63 is oporated l.cally to I;om.da VF-DR-4A-00 RA^J 0 _RG 1R-CC IN L1T, T' L V^ 996^0A9 ...c to the S6C pUMPS and aTO SHUTON COLNG hat rinkt for #;A ROC HXc46A FV 80 R" 1 DIAPH RAG OPERA.TED FCV 38-09 i6 oprated locally to VF;-R- 0-l60 RAFLOW-A CONTROL VALVE control- flow throweugh 69C HX 38 1326 toSHTDW cool ING Hr.-AT Fogulate RCS_ caoe I M-NA. ratoEXHNER 11 FLO0WCOTR AI 'IE446A F;QA -344 9_2 D!IA.1HR-AG O. -PER-ATED FCV 38 114 is oporated locally to VFDMR ISA 006 RAFLW OTROL VAL VE control flowA thro-ugh SOC HX 38 1 20 toAHlDW COO, ING HEAT regulate =CS cooel dao40i Fate,46A IV 38 02 AMQQ- R0- PERATM WV 28 02 D~ operated locally ý.ia the VFDR 46A 008 RASHUTDO30WN COOL INGQ OU1TZIT hand whaal to octablich a Q SOC uotioOUTSDE SOL.TIN VLVE fewal; katfom the RCS to accamplich16A BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-1100-68 and BV-1 00-69 are locally VFDR-16A-009 RABV-100-69 & 122 MANUAL VALVES opened to periodically refill the_________Emergency Condenser Makeup tanks.16A PMP-.100-02 DIESEL DRIVEN VERTICAL PMP-1 00-02 is manually started and VFDR-16A.009 RATURBINE FIRE PUMP run as needed to supply fire water to______________________ ______________________Emergency Condenser Makeup tanks.16A VLV-60-1 1 MAKEUP VALVE TO LOOP 12 VLV-60-1 I is locally throttled to control VFDR-1 6A-004 RAEMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide aRCS heat sink for decay heat removal.NMPI, April 2013 Page G-32 IINMP1, April 2013Page G-32 I Constellation Enerav Nuclear GrouDAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)Fire Component Component Description Actions VFDR RAJPCSArea16A VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-16A-003 RAEMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide aRCS heat sink for decay heat removal.46B BV-7-63 44" AIR ^OPERATED B6LOKING 6V 70 53 is ep.rat'd Iocally to provde VFDR-41"16B- RAALER2CLC INLETVW 98HA cooIng Wtr the 60G pumps and aTO HTDOW OLN'G' hcat a-Alk for the SDQC WXA.4468 FGV-3& 40 AV DL'-.PHRAOM-4 OPER14ATED FCY 38 1 0 is cperated locsally to VFR4OB.lg496 RA..O OTROL VAL.VE can.. ..fl0w o ugh .DG 1. X 3.8 41 32 2 toSHTONI coOlIW NG E rogulatS RCS G-0-0l dW_.A FatoEXCHANGER412 FLOW16B BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-16B-007 RABV-100-69 & 122 MANUAL VALVES opened to periodically refill theEmergency Condenser Makeup tanks.16B PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-16B-007 RATURBINE FIRE PUMP run as needed to supply fire water toEmergency Condenser Makeup tanks.16B VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-1 1 is locally throttled to control VFDR-16B-004 RAEMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide aRCS heat sink for decay heat removal.16B VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-16B-003 RAEMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide aRCS heat sink for decay heat removal.47A -0-63 11" AIR O ^ERATED AL OCVKNG 8V 70 63 Is op ..-atcd locally to VZDR47A"007 RAI- PRBCLC INLET ;Vtor to #is 6DC pumps and aTO SHUll-T.DOWAN COO-INIG hcot cink for thA 21DC 1-14cS-YST-EM47A FGV43-09 8- OPERATED FCV 328 00 is operated loWally to VFDR 47-A 00 RAF6LOWNTO VALVE flo, tR, ugh 60C HX 38 136 teHUDW C L HE--AT rOgulato RCS, eol ddo rat4.EXCHAANGER 11 FLOWNMPI, April 2013 Page G-33 IINMPI, April 2013Page G-33 I Constellation Enerav Nuclear GrouDAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)FireArea Component Component Description Actions VFDR RA/PCS7A FGV-138 LOOP" 11 2 A ND OPERATED FCOV 11 BV 10 0 c6 opond ted lBcally to VFDR-17A-006 RAFLO CO2NTARL VALVES GnAI flo W through SDC HX 3r 12i toSWUTDOWN01 COOEING H eregulate RCS odenso r Fakpta.EXHNER 13 FLOW941A 04-100-02 DIO R 0VEN VRATEID IV 3P 02 is mpoanted lally saea VFDR-47A-O09 RA;HTON COOL1ING OUTLEfT hand wheel te es&t-ablish a 2-D-C Al 'GiQnOUTSIDE 1I01 ATION VALV flcathl *aom the RCS to accmplichdecay hoo-t rcm -0-a.17A BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-1 00-68 and BV-1 00-69 are locally VFDR-1 7A-009 RABV-100-69 & 122 MANUAL VALVES opened to periodically refill theEmergency Condenser Makeup tanks.17A PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-17A-009 RATURBINE FIRE PUMP run as needed to supply fire water to)__________Emergency Condenser Makeup tanks.17A VLV-60-1 1 MAKEUP VALVE TO LOOP 12 VLV-60-1 1 is locally throttled to control VFDR-17A-004 RAEMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide aRCS heat sink for decay heat removal.17A VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-17A-003 RAEMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide aRCS heat sink for decay heat removal.1-TB AV70 62 14" AIR OPERIATED0 W1LOCKING RV 7-0 53 is operated locally to pro'~de vFDR-17-8 00O RAVALVE RBCLC INLET VA'VE cooling Vtcr 1to the SDC pumps and aS-Y-STEMTOV COOL ttING ha ~kfrtoSCHs47-8 F-GV as&40 8 DI.APHWR-AGM O. -PEERATED FCV 38 10 Is oporatod locally to VFDR-4-78-0()5 RAFLW OTROL VALVE contWrolI flow.A t-hro-ugh 8DC H4X 38 1322 toHT OWN COOLING HEA regulato RCS cool down Fate.EXCW.N--G....R 12 FLOW17B BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-17B-007 RABV-100-69 & 122 MANUAL VALVES opened to periodically refill theEmergency Condenser Makeup tanks.17B PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-17B-007 RATURBINE FIRE PUMP run as needed to supply fire water toEmergency Condenser Makeup tanks.NMPII April 2013 Page G-34 IINMP1, April 2013Page G-34 I Constellation Energy Nuclear GroupAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)Fire Component Component Description Actions VFDR RA/PCSArea17B VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-11 is locally throttled to control VFDR-1 78-004 RAEMERGENCY CONDENSERS makeup to EC's 121 & 122 to provide aRCS heat sink for decay heat removal.17B VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-1 7B-003 RAEMERGENCY CONDENSERS makeup to EC's 111 & 112 to provide aRCS heat sink for decay heat removal.18 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-18-011 RABV-100-69 & 122 MANUAL VALVES opened to periodically refill theEmergency Condenser Makeup tanks.18 FCV-39-15 EMERGENCY CONDENSER FCV-39-15 operated locally via the VFDR-18-005 RASTEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory lossCONTROL VALVE -FLOW flow path from the EC's 111 & 112CONTROL VALVE 11 STEAM RCS return path drain line.LINE DRAIN18 FCV-39-16 EMERGENCY CONDENSER FCV-39-16 operated locally via the VFDR-18-004 RASTEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory lossCONTROL VALVE -FLOW flow path from the ECs 121 & 122 RCSCONTROL VALVE 12 STEAM return path drain line.LINE DRAIN18 IV-01-04 24" AIR OPERATED (ANGLE) AIr vented manually at MSIV IV-01-04 VFDR-18-002 RAISOLATION VALVE WITH to dose valve to prevent inventory loss.SOLENOID VALVES -MAINSTEAM OUT ISOLATION VALVE418 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-18-011 RATURBINE FIRE PUMP run as needed to supply fire water toEmergency Condenser Makeup tanks.18 VLV-05-31 LOOP 11 VENT LINE MANUAL VLV-05-31 is operated locally via the VFDR-18-003 RAISOLATION VALVE DOWN hand wheel to isolate an inventory lossSTREAM OF 05-01R AND 05-11 flow path from the ECs 111 & 112 RCSsteam supply vent line.18 VLV-05-32 LOOP 12 VENT LINE MANUAL VLV-05-32 is operated locally via the VFDR-18-003 RAISOLATION VALVE hand wheel to isolate an inventory lossDOWNSTREAM OF 05-04R AND flow path from the ECs 121 & 122 RCS05-12 steam supply vent line.NMPI, April 2013 Page G-35 I!NMP1, April 2013Page G-35 I Constellation Energy Nuclear GroupAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)FireArea Component Component Description Actions VFDR RA/PCS18 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-1 1 is locally throttled to control VFDR-18-008 RAEMERGENCY CONDENSERS makeup to EC's 121 & 122 to provide aRCS heat sink for decay heat removal.18 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-18-008 RAEMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide aRCS heat sink for decay heat removal.4-9 aV-TO-43 11 AIR OpRN..ATE 91LOC.K,, 1V 79 53 is op-.ated locally to preo".do VFDR 49993 RA~~I~~j~~IV RBL ILTGI.I~ liAg Vtr to the 89G pumps and aTSH DOWNl COG'ING hMat sink for tho ROG HXc.49 FV 3. CID9 8" OPE.RATED FCV 33 09 is operated locally to VFDR40-0Q9 RAFLO 1COTRO VALVE eantrol flowi t-hrough SDC HX 38 132 to lSHUTDO0WN COOLINAG HEAT rogulato RCS- ca-l d-ewim Fate.ECHANG;R 441 FLOWA-44D FGV as 11 i "DAW r-MP..;..OP-PR.TED FC-3R IQ 19 BV-100-8 DIOPH111&2A- OE TD 2 CV-8 an i V-100-9are locally to VFDR-19-007 RAFLOW- C2NUTROL VALVES openedl flow henough SDc HX 3r 120 toGOH I-T 0DOW NCOOLI61N G HEA rogulate RCS see' down Fate.E-XCHANG-ER 12 FLOWEmrgncronenerMkep ans9 I-108-02 MOETOR OPERIATECD IV 3P 02 is mpatnd lcally sta rte VFDR49-0076 RASHU1-TDOWNA9A COOIN -OTLET- hand vpheel to octa-blich a- SDC suctio201ID IfOATION' VALVE flev~ath from the RCS to acamplish________________________docay heat rmoeval.19 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-1 00-68 and BV-1 00-69 are locally VFDR-19-007 RABV-100-69 & 122 MANUAL VALVES opened to periodically refill the_________ ______________________Emergency Condenser Makeup tanks.19 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-19-007 RATURBINE FIRE PUMP run as needed to supply fire water toEmergency Condenser Makeup tanks.19 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-1 1 is locally throttled to control VFDR-19-001 RAEMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide aRCS heat sink for decay heat removal.NMPI, April 2013 Page G-36 ImNMPI, April 2013Page G-36 I Constellation Energy Nuclear GroupAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)FireArea Component Component Description Actions VFDR RA/PCS19 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-19-001 RAEMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide aRCS heat sink for decay heat removal.20 BV- 44-- " AIR 0OPE6RATED ' LOCKING BV. 70 53 62 epwrted !oly to pwid. VFDR RAV.ALkE R6LCINE VLVE- coo1ing W011F to9 the 69C pumpcD and a~O H NTAD l NCOO' ING heat ank for the 20C

20 FGV 38-09 92 DLIA.PI.IRAGM OPE6RATE-D FCY 38 00 is oporotod locally to WFOR2-20-QG RAO CONITROL ,AI AVE cntral fow throuh C HX L 8 4126 tc9HhDOW COOL ING HEAT rcgulato RC- coo' dwoni Fato4 4 rI /60ACIZNITROL VAI20._._G 8 4.. " I, A U. dA.P.R.. OPO ..TE' D r,-, uateC l do fl r. .2Q RAErXCH.ANIG-E-R 412 FLOW920 IV 38 02 AMOTR OPERATED IV 38 02 is ep-r.t.d lea!.ly Y.1, "h. VFDR 20 00Q1 RASHUTi'lDOWNti COOL 0-1ING OUTL6ET hand- v4,col to- e ctoblisch a 2 SDC-A GcienOUTMID ....ATION VALVE fipath f., om t-. .RCS to ase.m p!lih__________ _____________decoy heat romovi2'.20 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-20-008 RABV-100-69 & 122 MANUAL VALVES opened to periodically refill theEmergency Condenser Makeup tanks.20 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-20-008 RATURBINE FIRE PUMP run as needed to supply fire water toEmergency Condenser Makeup tanks.20 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-11 is locally throttled to control VFDR-20-002 RAEMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide aRCS heat sink for decay heat removal.20 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-20-002 RAEMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide aRCS heat sink for decay heat removal.NMPI, AprIl 2013 Page G-37 IINMPI, April 2013Page G-37 I Constellation Energy Nuclear GroupAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)Fire Component Component Description Actions VFDR RA/PCSArea2-1- RV4 53 14" AIR OPERATED BLOC81 "KING W 70 53 he oporat.d lceally to pr"ldo VFDR 24 094 RAI*)AjI~ BCC ILE VLVE coln -to the SDC pumps and aTO )HTOWN COOLING hat cA~k. forf the 89G HXs.24- FGV--8-0A 8" DIA-PHRPAG1.2.. OPERA-TED FCV 38 09 is oporated locally to VFW2.-4-006 RAFOW C-ONTROL VALVE c...ntrl Amow thruh--..* 6DC HX 38 125 teSHUTDOW COINGQ HEAT rgulato RCS cool do;A rato.E-=CH.AGR 11 FLOW24- FGV s .4. 8 .DIAPHR.A.GM.. OPER-A.TE-D FOY 389 2. 4. oporoted Wcally R, 24 06 RAFL WCOTROL VALVE contro flowA. thrOu6gh SIX HX 38 4120 toSHUTDON COOLING I-EA rogulato RCS eeol devm Fate.EX=.CHANIGQrER 13 FL-OW24- IV 38 02 OTROPERATED IV 38 02 Ic oporated locally ýAa toVFDR 21 004- RA9HU TDO WNv C , I OUTLE.T hand .ea. l tno .ctabl!h a SDC ScUotroOU1 TSID IOR 'IN'.A flwopath from the RCS to accomplich________________________decay heat reomwal,21 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-21-008 RABV-100-69 & 122 MANUAL VALVES opened to periodically refill theEmergency Condenser Makeup tanks.21 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-21-008 RATURBINE FIRE PUMP run as needed to supply fire water toEmergency Condenser Makeup tanks.21 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-1 1 is locally throttled to control VFDR-21-002 RAEMERGENCY CONDENSERS makeup to EC's 121 & 122 to provide aRCS heat sink for decay heat removal.21 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-21-002 RAEMERGENCY CONDENSERS makeup to EC's 111 & 112 to provide aRCS heat sink for decay heat removal.22 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-22-008 RABV-100-69 & 122 MANUAL VALVES opened to periodically refill theEmergency Condenser Makeup tanks.22 FCV-39-16 EMERGENCY CONDENSER FCV-39-16 operated locally via the VFDR-22-004 RASTEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory lossCONTROL VALVE -FLOW flow path from the EC's 121 & 122CONTROL VALVE 12 STEAM RCS return path drain line.lNMP1, April 2013Page G-38 I Constellation Enerav Nuclear GroupAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)FireArea Component Component Description Actions VFDR RA/PCSLINE DRAIN22 IV-01-04 24- AIR OPERATED (ANGLE) Air vented manually at MSIV IV-01-04 VFDR-22-002 RAISOLATION VALVE WITH to close valve to prevent inventory loss.SOLENOID VALVES -MAINSTEAM OUT ISOLATION VALVE422 PMP-1 00-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-22-008 RATURBINE FIRE PUMP run as needed to supply fire water toEmergency Condenser Makeup tanks.22 VLV-05-32 LOOP 12 VENT LINE MANUAL VLV-05-32 is operated locally via the VFDR-22-003 RAISOLATION VALVE hand wheel to isolate an inventory lossDOWNSTREAM OF 05-04R AND flow path from the ECs 121 & 12205-12 RCS steam supply vent line.22 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-11 is locally throttled to control VFDR-22-007 RAEMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide aRCS heat sink for decay heat removal.22 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-22-007 RAEMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide aRCS heat sink for decay heat removal.23 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-23-008 RABV-100-69 & 122 MANUAL VALVES opened to periodically refill theEmergency Condenser Makeup tanks.23 FCV-39-16 EMERGENCY CONDENSER FCV-39-16 operated locally via the VFDR-23-003 RASTEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory lossCONTROL VALVE -FLOW flow path from the ECs 121 & 122 RCSCONTROL VALVE 12 STEAM return path drain line.LINE DRAIN23 IV-01-04 24- AIR OPERATED (ANGLE) AIr vented manually at MSIV IV-01-04 VFDR-23-002 RAISOLATION VALVE WITH to close valve to prevent inventory loss.SOLENOID VALVES -MAINSTEAM OUT ISOLATION VALVE4NMPI, AprIl 2013 Page G-39 IINMP1, April 2013Page G-39 I Constellation Enerav Nuclear GrouDAttachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)FireArea Component Component Description Actions VFDR RA/PCS23 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-23-008 RATURBINE FIRE PUMP run as needed to supply fire water toEmergency Condenser Makeup tanks.23 VLV-05-32 LOOP 12 VENT LINE MANUAL VLV-05-32 is operated locally via the VFDR-23-004 RAISOLATION VALVE hand wheel to isolate an inventory lossDOWNSTREAM OF 05-04R AND flow path from the ECs 121 & 122 RCS05-12 steam supply vent line.23 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-1 1 is locally throttled to control VFDR-23-006 RAEMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide aRCS heat sink for decay heat removal.23 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-23-006 RAEMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide aRCS heat sink for decay heat removal.24 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-1 00-68 and BV-1 00-69 are locally VFDR-24-007 RABV-100-69 & 122 MANUAL VALVES opened to periodically refill theEmergency Condenser Makeup tanks.24 FCV-39-15 EMERGENCY CONDENSER FCV-39-15 operated locally via the VFDR-24-003 RASTEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory lossCONTROL VALVE -FLOW flow path from the ECs 111 & 112 RCSCONTROL VALVE 11 STEAM return path drain line.LINE DRAIN24 IV-01-04 24" AIR OPERATED ( ANGLE) Air vented manually at MSIV IV-01-04 VFDR-24-002 RAISOLATION VALVE WITH to close valve to prevent inventory loss.SOLENOID VALVES -MAINSTEAM OUT ISOLATION VALVE424 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-24-007 RATURBINE FIRE PUMP run as needed to supply fire water toEmergency Condenser Makeup tanks.24 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-1 1 is locally throttled to control VFDR-24-004 RAEMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide aRCS heat sink for decay heat removal.24 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-24-004 RAEMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide aRCS heat sink for decay heat removal.NMPI, April 2013 Page G-40 IINMPI, April 2013Page G-40 I Constellation Energy Nuclear Group Attachment G -Recovery Actions TransitionTable G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)FireArea Component Component Description Actions VFDR RA/PCS24 VLV-93-13 12" GATE VALVE -121 VLV-93-13 dosed locally via the hand VFDR-24-006 RACONTAINMENT SPRAY RAW wheel to isolate CTSRW flow toWATER PUMP DISCHARGE Containment Spray Header #11.VALVENMPArl21PaeG4!NMPI, April 2013Page G-41 I

v t e Letter Sequence Response to RAI
TAC:ME8899, (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request Acceptance Supplement Administration Questions 13 November 2012 9 October 2013 9 October 2013 12 February 2014 Answers 27 February 2013 27 March 2013 30 April 2013 30 April 2013 30 April 2013 15 April 2014 Results Approval, Approval Other: ML14181B086
MONTHYEAR2013-04-30 [Table View]

ML13127A397

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