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| document type = Letter, License-Application for Facility Operating License (Amend/Renewal) DKT 50
| document type = Letter, License-Application for Facility Operating License (Amend/Renewal) DKT 50
| page count = 46
| page count = 46
| project = TAC:MF2462, TAC:ME3145
| project = TAC:ME3145, TAC:MF2462
| stage = Other
| stage = Other
}}
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==5.0 ENVIRONMENTAL CONSIDERATION==
==5.0 ENVIRONMENTAL CONSIDERATION==
6.0 REFERENCESi of ii ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGETABLE OF CONTENTSATTACHMENTS1. Nine Mile Point Unit 2 Proposed Changes to Technical Specifications (Mark-ups)Note: Attachments 2 through II of the Enclosure to the Nine Mile Point Unit 2 License AmendmentRequest dated November 1, 2013 are not revised or reissued in this revision.2. Nine Mile Point Unit 2 Changes to Bases for Technical Specifications (Mark-ups)3. List of Regulatory Commitments4. MELLLA+ Risk Evaluation5. Nine Mile Point Unit 2 Power/Flow Operating Map for Current Cycle6. General Electric -Hitachi Affidavit Justifying Withholding Proprietary Information inNEDC-33576P7. Global Nuclear Fuel Affidavit Justifying Withholding Proprietary Information in GNF-0000-0156-7490-RO-P8. NEDC-33576NP, Safety Analysis Report for Nine Mile Point Unit 2 Maximum Extended Load LineLimit Analysis Plus (Non-proprietary)9. Global Nuclear Fuel Report GNF-0000-0 1 56-7490-RO-NP, "GNF Additional Information Regardingthe Requested Change to the Technical Specification SLMCPR," dated August 26, 2013 (Non-proprietary)10. NEDC-33576P, Safety Analysis Report for Nine Mile Point Unit 2 Maximum Extended Load LineLimit Analysis Plus (Proprietary)11. Global Nuclear Fuel Report GNF-0000-0 1 56-7490-RO-P, "GNF Additional Information Regardingthe Requested Change to the Technical Specification SLMCPR," dated August 26, 2013 (Proprietary)ii of ii ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE1.0 SUMMARY DESCRIPTIONThis evaluation supports a request to amend Renewed Operating License (OL) NPF-69 for NineMile Point Unit 2 (NMP2). The proposed amendment includes supporting changes to the NMP2Technical Specifications (TSs) necessary to: 1) implement the Maximum Extended Load LineLimit Analysis Plus (MELLLA+) expanded operating domain; 2) change the stability solution toDetect and Suppress Solution -Confirmation Density (DSS-CD); 3) use the TRACG04 analysiscode; and 4) incr.ease the isotpic en.rihment of boro.n 10 in the sodium pentabor.. e so!uti.nutilized in the Standby Liquid Control System (SLS); and 5) increase the Safety Limit MinimumCritical Power Ratio (SLMCPR) for two recirculation loops in operation.The following is a list of the proposed changes to the NMP2 TSs:* Revise Safety Limit (SL) 2.1.1.2 by increasing the SLMCPR for two recirculation loops inoperation from > 1.07 to > 1.09" Revise the acceptance criterion in TS 3.1.7, "Standby Liquid Control (SLC) System,"Surveillance Requirement (SR) 3.1.7.7 by increasing the discharge pressure from> 1,327 pounds per square inch gauge (psig) to > 1,335 psig" Revise the aeceptance cr-iter-ion int TS SR 3.1.7.10 by inremasing the sodium pentaborateboron 10 enr-iehment Fcquir-ement from -&#xfd; 25 atom percent to &#xfd;! 92 atom percent, and make-aCorespending e hange in Ti S Figure 3.1.7 1, "Sodium Pentabortec SolutieoVolumne/Conentration" Requiements"" Revise TS Figure 3.1.7 I to account for the deer-ease in the miniftmum volume of the SLS t"nfrom 4,558.6 gallons and 1,288 gallons at sodium pentabConte ionoentations of 13.6oT and14.49%, respectively, to 1,600 gallons and 1,530 gallons at sodium pentaborete concentrationsof 13.6%4 and 14.41%, r-espectively* Change the Required Actions for Condition F of TS 3.3.1.1, "Reactor Protection System(RPS) Instrumentation"*Change Condition G of TS 3.3. 1.1* Add new Conditions J and K to TS 3.3.1.1* Correct an editorial error in Note 3 to TS SR 3.3.1.1.13 (i.e., "ORRM" is changed to"OPRM")* Eliminate TS SR 3.3.1.1.16 and references to it in TS Table 3.3.1.1-1, "Reactor ProtectionSystem Instrumentation"* Change the allowable value (AV) for TS Table 3.3.1.1-1, Function 2.b, Average PowerRange Monitor (APRM) -Flow Biased Simulated Thermal Power (STP) -Upscale from"5 0.55W+60.5% [Rated Thermal Power] RTP and 5 115.5% RTP" to"< 0.61W + 63.4% RTP and < 115.5% RTP"* Add a new note to TS Table 3.3.1.1-1, Function 2.b that requires the Flow Biased SimulatedThermal Power -Upscale scram setpoint to be reset to the values defined by the CoreOperating Limits Report (COLR) to implement the Automated Backup Stability Protection(BSP) Scram Region in accordance with Required Action F-,2-4F.2 of TS 3.3.1.1I of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE" Add a new note to TS Table 3.3.1.1-1, Function 2.e, Oscillation Power Range Monitor(OPRM) -Upscale to denote that following implementation of DSS-CD, DSS-CD is notrequired to be armed while in the DSS-CD Armed Region during the first reactor startup andduring the first controlled shutdown that passes completely through the DSS-CD ArmedRegion. However, DSS-CD is considered operable and capable of automatically arming foroperation at recirculation drive flow rates above the DSS-CD Armed Region* Change the mode of applicability for TS Table 3.3.1.1-1, Function 2.e, OPRM-Upscale fromMode I to > 18% RTP" Change the allowable value for TS Table 3.3.1.1-1, Function 2.e from "As specified in theCOLR" to "NA", Add a pr-.hibiti'n t. TS Limiting Condition for Operation (LCO) 3.4.1, "Recirculation LoopsOperating," is modified to prohibittimhat-pfhbits operation in the Maximum Extended LoadLine Limit Analysis (MELLLA) domain or MELLLA+ expanded operating domain asdefined in the COLR when in operation with a single recirculation loop* Add Required Action B.2 to TS 3.4.1 to identify that intentional operation in the MELLLAdomain or MELLLA+ domain as defined in the COLR is prohibited when a recirculationloop is declared "not in operation" due to a recirculation loop flow mismatch not within limits" Revise TS 5.6.5.a.4 to replace "Reactor Protection System Instrumentation Setpoint for theOPRM -Upscale Function Allowable Value for Specification 3.3.1.1" with "The ManualBackup Stability Protection (BSP) Scram Region (Region I), the Manual BSP ControlledEntry Region (Region II), the modified APRM Simulated Thermal Power -High setpointsused in the OPRM (Function 2.e), Automated BSP Scram Region, and the BSP Boundary forSpecification 3.3.1.1"" Add TS 5.6.8, "OPRM Report," to define the contents of the report required by new RequiredAction F-.24F.3 of TS 3.3.1.1Nuclear Regulatory Commission (NRC) approval of the requested operating domain expansionwill allow NMP2 to implement operational changes that will increase operational flexibility forpower maneuvering, compensate for fuel depletion, and maintain efficient power distribution inthe reactor core without the need for more frequent rod pattern changes. MELLLA+ supportsoperation of NMP2 at Current Licensed Thermal Power (CLTP) of 3,988 Megawatts -Thermal(MWth) with core flow as low as 85% of rated core flow. By operating in the MELLLA+ domain,a significantly lower number of control rod movements will be required than in the presentoperating domain. This represents a significant improvement in operating flexibility. It alsoprovides safer operation, because reducing the number of control rod manipulations:(a) minimizes the likelihood of fuel failures and (b) reduces the likelihood of accidents initiatedby reactor maneuvers required to achieve an operating condition where control rods can bewithdrawn.Attachments 8 and 10 provide the non-proprietary and proprietary versions of the MELLLA+Safety Analysis Report (MELLLA+ SAR), respectively. The MELLLA+ SAR follows theguidelines contained in GE-Hitachi Nuclear Energy Americas (GEH) Licensing Topical Report(LTR) NEDC-33006P-A, Revision 3, "Maximum Extended Load Line Limit Analysis Plus"(MELLLA+ LTR) (Reference 1). The MELLLA+ SAR provides the technical bases for thisrequest and contains an integrated summary of the results of the underlying safety analyses andevaluations performed specifically for the NMP2 expanded operating domain.The MELLLA+ SAR also provides the analyses to change the NMP2 stability solution fromOption III to DSS-CD and use the GEH analysis code TRACG04. DSS-CD as required by theMELLLA+ LTR Safety Evaluation Report. DSS-CD is being implemented using the guidelines2 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEcontained in GEH LTR NEDC-33075P-A, Revision 7, "General Electric Boiling Water ReactorDetect and Suppress Solution -Confirmation Density," (Reference 2). The use of TRACG04 isbeing implemented using the guidelines contained in GEH LTR "DSS-CD TRACG Application,"NEDE-33147P-A, Revision 4, August 2013 (Reference 3).The proposed change to the SLMCPR value for two recirculation loops in operation is based onan analysis performed by Global Nuclear Fuel (GNF) for NMP2 during Cycle 15 operations withMELLLA+ conditions. The GNF report, GNF-0000-0156-7490-RO, "GNF AdditionalInformation Regarding the Requested Changes to the Technical Specification SLMCPR," datedAugust 26, 2013, supports changing the two recirculation loops in operation value of SLMCPRfrom > 1.07 to >_ 1.09, and maintaining the single recirculation loop in operation value ofSLMCPR at > 1.09. These values are based on NRC approved methods and procedures.Attachments 9 and 11 of this Enclosure provide non-proprietary and proprietary versions of theGNF report, respectively.Attachments 10 and 11 of this Enclosure contain information considered to be proprietary asdefined by 10 CFR 2.390. GEH and GNF, as the owners of the proprietary information inAttachments 10 and 11, respectively, have executed the affidavits provided in Attachments 6and 7 to this Enclosure detailing the reasons for withholding the proprietary information.Attachment 3 delineates the regulatory commitments associated with the proposed change.2.0 DETAILED DESCRIPTION2.1 Background2.1.1 MELLLA+Operation of Boiling Water Reactors (BWRs) requires that reactivity balance be maintained toaccommodate fuel burn-up. BWR operators have two options to maintain this reactivity balance:(a) control rod movements or (b) core flow adjustments. Because of the strong void reactivityfeedback and its distributed effect through the core, flow adjustments are the preferred reactivitycontrol method. Operation at low-flow conditions at rated power level also increases the fuelcapacity factor through spectral shift and the increased flow region compensates for reactivityreduction due to fuel depletion during the operating cycle.At NMP2, an Extended Power Uprate (EPU) was implemented by extending the MELLLAoperating domain up to the EPU power level (3,988 MWth). The extension of the MELLLA lineto EPU power levels reduces the available core flow window. In addition, the increased corepressure drop with EPU limits the recirculation flow capability. Consequently, EPU plantsgenerally operate with a reduced core flow window and compensate for reactivity loss withcontrol rod movement. Operation in the MELLLA+ expanded operating domain will provide alarger core flow window for NMP2.In June 2009, the NRC approved the use of the MELLLA+ LTR (NEDO-33006P-A)(Reference 1) as a basis for MELLLA+ operating domain expansion license amendment requests,subject to limitations specified in the MELLLA+ LTR and in the associated NRC safetyevaluation. The NMP2 request complies with the specified limitations and conditions asdiscussed in Appendix B of Attachments 8 and 10.3 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEIn January 2008, the NRC approved the use of the DSS-CD LTR (NEDC-33075P) as a basis forimplementing DSS-CD as a stability solution to replace the Option III solution in licenseamendment requests, subject to limitations specified in the DSS-CD LTR and in the associatedNRC safety evaluation. The NMP2 request complies with Revision 7 of NEDC-33075P(Reference 2), including the specified limitations and conditions as discussed in Appendix C ofAttachments 8 and 10.The TRACG code for use in DSS-CD applications (NEDE-33147P-A) was approved by NRC inNovember 2007. The NMP2 request complies with Revision 4 of NEDE-33147P-A(Reference 3).In addition, the NRC approved the Applicability of GE Methods to Expanded Operating DomainLicensing Topical Report (NEDC-33173P-A) which imposes limitations and requirements for theuse of GEH Methods in expanded operating domains including power uprates and MELLLA+domains. The NMP2 request complies with Revision 4 of NEDC-33173P-A (Reference 4),including the specified limitations and conditions as discussed in Appendix A of Attachments 8and 10.Detailed evaluations of the reactor, engineered safety features, power conversion, emergencypower, support systems, and design basis accidents were performed and are provided inAttachments 8 and 10. These evaluations demonstrate that NMP2 can safely operate in theMELLLA+ expanded operating domain with DSS-CD as the thermal hydraulic stability solution.2.1.2 Standby Liquid Control System is.topi. Enrichment of Boron 1NMPNS proposes to inr-easc the isotopi. of beroen 10 in the sodium pentaboatesolution used to prcparc the nouttron absortber- solution in the Standby Liquid Control Systom(SLS) to &#xfd;! 92 atom percoent. The proposed berefn 10 enrielifitnt value allows the minimfum netsolution volume storoed in the Slug storage tank to be deer-eased to 1,530 gallons EAt 14.4% sodiumpentaberate eoncentraion and 1,600 gallons at 13.6% sodium pentaborae concontration. Iadditiei-,NMPNS proposes to increase the acceptance criterion for the SLS pump dischargepressure from 1,327 psig to:- 1,335 psig.2.1.3 Safety Limit Minimum Critical Power RatioNMPNS proposes to revise SL 2.1.1.2 by increasing the SLMCPR for two recirculation loops inoperation from > 1.07 to ! 1.09. The proposed change to the SLMCPR value for tworecirculation loops in operation is based on an analysis performed by GNF for NMP2 duringCycle 15 operations with MELLLA+ conditions. The GNF report, GNF-0000-0156-7490-RO-P,"GNF Additional Information Regarding the Requested Changes to the Technical SpecificationSLMCPR," dated August 26, 2013, supports changing the two recirculation loops in operationvalue of SLMCPR from ! 1.07 to ! 1.09, and maintaining the single recirculation loop inoperation value of SLMCPR at &#xfd; 1.09. These values are based on NRC approved methods andprocedures. Attachments 9 and I11 of this Enclosure provide non-proprietary and proprietaryversions of the GNF report, respectively.2.2 Proposed Changes to the Nine Mile Point Unit 2 Technical SpecificationsNMP2 TS changes are required to allow operation in the expanded MELLLA+ operating domain,use of DSS-CD, inerease the isotopic enrzeinhmcnt of boroen 10 in the sodium pentaborate solution4 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEused to prepare the ne.tr.n abs.r.b.r. s.lution in the SLS, and increase the SLMCPR for tworecirculation loops in operation. Attachment 1 of this Enclosure provides a mark-up of the NMP2TS showing the proposed changes. Attachment 2 of this Enclosure provides a mark-up of theNMP2 TS Bases to show the corresponding changes to the TS Bases. Attachment 2 is providedfor information only. A description of each TS change is provided below.Safety Limit 2.1.1.2, Safety Limit Minimum Critical Power RatioSL 2.1.1.2 is revised to increase the SLMCPR for two recirculation loops in operation from> 1.07 to  1.09.TS 3.1.7, Standby Liquid Control (SLC) SystemTS SR 3.1.7.7 is revised to increase the acceptance criterion for the Standby Liquid ControlSystem (SLS) pump discharge pressure from > 1,327 psig to > 1,335 psig.T-S ISR 3.167. 10 is r-evised to iner-ease the boron 10 cnrichmcnt requirement of sodium pcntabrteffrom ? 25 eatm perceent to ? 92 atom perceent. in addition TS Figure 3.1.7 1 is updated to r-eocthe ineroease in the boront 10 cnriehment roquirement.TIS Figure 3.1.7 1, "Sodium Pentab-ratm Solution ati-n Requir.ements," isrevised to aecount for the change in the net volufe in the SL9 tank that arise thecnr-iehment iner-ease. The miniftmum volumne is changed from 4,558.6 gallonts and 4,288 gallons asodium pcntaber-ate coneentrationis of 13.6% and 14.4%, respeetively, to 1,600 gaillons and 1,530gallonis at a sodium pentaberate coneentration of 13.6% and 14.44, r-espeetively.TS 3.3.1.1, Reactor Protection System (RPS) InstrumentationRequired Actions F. 1 and F.2 of TS 3.3.1.1 and their associated Completion Times are replacedwith the following new Required Actions and Completion Times.REQUIRED ACTIONCOMPLETION TIMEF. I Initiate Action to implement the ManualBSP Regions defined in the COLR.ANDF-.2 F._2 Implement the Automated BSPScram Region using the modifiedAPRM Simulated Thermal Power -High scram setpoints defined in theCOLR.ANDR.-,F.__3 Initiate action in accordancewith Specification 5.6.8.Immediately12 hours90 dayslmmediately5 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGECondition G is modified to no longer apply in the event a Required Action and associatedCompletion Time of Condition F is not met.New Condition J (see below) is added to address the action to take in the event a Required Actionand associated Completion Time of Condition F is not met.New Condition K (see below) is added to address the action to takeAction and associated Completion Time of Condition J is not met.in the event a RequiredCONDITIONREQUIRED ACTIONCOMPLETION TIMEJ. Required Action andassociated Completion Timeof Condition F not met.J. 1 Initiate action toimplement the ManualBSP regions defined in theCOLR.ANDJ.2 Reduce operation to belowthe BSP Boundary definedin the COLR.ANDJ.3 ---------NOTE-------LCO 3.0.4 is not applicableImmediately12 hours120 daysRestore required channel tooperable.K. Required Action andassociated Completion Timeof Condition J not met.K.1ReducePOWERRTPTHERMALto less than 18%4 hours"ORRM" is changed to "OPRM" in Note 3 to TS SR 3.3.1.1.13.TS SR 3.3.1.1.16 is eliminated, and references to it in TS Table 3.3.1.1-1 are eliminated.TS Table 3.3.1.1-1, Function 2.b, Flow Biased Simulated Thermal Power -Upscale, contains botha flow-biased AV (_ 0.55W + 60.5% RTP) and a fixed AV at 115.5% RTP. The flow-biased AVwill be changed to (< 0.6 1W + 63.4% RTP).A new note is added to TS Table 3.3.1.1-1, Function 2.b that requires the Flow Biased SimulatedThermal Power -Upscale scram setpoint to be reset to the values defined by the COLR toimplement the Automated Backup Stability Protection (BSP) Scram Region in accordance withRequired Action F.2. F.2 of TS 3.3.1.1.A new note is added to TS Table 3.3.1.1-1, Function 2.e, OPRM -Upscale, to denote thatfollowing DSS-CD implementation, DSS-CD is not required to be armed while in the DSS-CDArmed Region during the first reactor startup and during the first controlled shutdown that passes6 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEcompletely through the DSS-CD Armed Region. However, DSS-CD is considered operable andcapable of automatically arming for operation at recirculation drive flow rates above the DSS-CDArmed Region.The mode of applicability for TS Table 3.3.1.1-1, Function 2.e, OPRM-Upscale is changed fromMode I to > 18% RTP.In addition, the allowable value for Function 2.e is changed from "As specified in the COLR" to"4NA."TS 3.4.1, Recirculation Loops, OperatingLCO 3.4.1 is modified to include an additional provision that will prohibit intentional operation inthe MELLLA domain or the MELLLA+ domain as defined in the COLR when only a singlerecirculation loop is in operation. It will state:"... One recirculation loop shall be in operation provided the plant is not operating in theMELLLA or MELLLA+ domain defined in the COLR and provided the following limits areapplied when the associated LCO is applicable:..."A new Required Action B.2 is added to prohibit intentional operation in the MELLLA domain orthe MELLLA+ domain defined in the COLR in the event a recirculation loop is declared to be"not in operation" due to a recirculation loop flow mismatch. The Completion Time for this newRequired Action is 2 hours.TS 5.6.5, Core Operating Limits Report (COLR)TS 5.6.5.a is modified by replacing the reference to "Reactor Protection System InstrumentationSetpoint for the OPRM -Upscale Function Allowable Value for Specification 3.3.1.1" with areference to "The Manual Backup Stability Protection (BSP) Scram Region (Region 1), theManual BSP Controlled Entry Region (Region II), the modified APRM Simulated ThermalPower -High setpoints used in the OPRM (Function 2.e), Automated BSP Scram Region, andthe BSP Boundary for Specification 3.3.1.1 ."TS. 5.6.8, OPRM ReportThe following new report requirement is added as TS 5.6.8, "OPRM Report:""When a report is required by Required Action F-,2-F.3 of TS 3.3.1.1, "RPSInstrumentation," a report shall be submitted within 90 days of ,nteing CONDITION Fthefollowing 90 days. The report shall outline the preplanned means to provide backup stabilityprotection, the cause of the inoperability, and the plans to- and schedule for restoring therequired instrumentation channels to OPERABLE status."The new TS section is numbered TS 5.6.8, because on November 21, 2012, NMP2 submitted aLicense Amendment Request (LAR) to create a new TS section that is numbered TS 5.6.7 for theReactor Coolant System Pressure and Temperature Limits Report (Reference 5). NMPNSanticipates that LAR will be approved by the NRC and implemented at NMP2 prior to approvalof the MELLLA+ LAR. The numbering of TS 5.6.8 is an administrative consideration. The7 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEMELLLA+ LAR is independent of the LAR submitted on November 21, 2012. NRC approval orrejection of Reference 5 would have no technical impact on the MELLLA+ LAR.2.3 Modification SummaryThe MELLLA+ core operating domain expansion does not require major plant hardwaremodifications. The core operating domain expansion involves changes to the core power/flowmap and a small number of setpoints and alarms. Because there are no increases in the operatingpressure, power, steam flow rate, and feedwater flow rate, there are no major modifications toother plant equipment.The stability solution is being changed from Option III to the DSS-CD solution. The DSS-CDsolution algorithm, licensing basis, and application procedures are generically described in theDSS-CD LTR (Reference 2), and are applicable to NMP2. The DSS-CD solution uses the samehardware as the current Option III solution. DSS-CD requires a revision to the existing stabilitysolution software.The boroen 10 enrichmfent in the sodium pentabor-ate solution in the SLS is incr-eased fo! 25 atom pcrccnt to ! 92 atom perceent. The increase in the boroen 10 enriehment in the scdiumpentaborate solution for- the SLS is sufficient to decrease the sodium pentabernte solution velumesterce in tMle SLS storage tank. In a..iti.n, tc I ne SLS pump discharge pressure acceptancecriterion is changed to > 1,335 psig. Changes to instrumentati. n setp..ots will be made toaccotunt for these changes. The increase in the SLMCPR for two recirculation loops in operationdoes not require any physical modifications to structures, systems, or components.
 
==6.0 REFERENCES==
i of ii ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGETABLE OF CONTENTSATTACHMENTS1. Nine Mile Point Unit 2 Proposed Changes to Technical Specifications (Mark-ups)Note: Attachments 2 through II of the Enclosure to the Nine Mile Point Unit 2 License AmendmentRequest dated November 1, 2013 are not revised or reissued in this revision.2. Nine Mile Point Unit 2 Changes to Bases for Technical Specifications (Mark-ups)3. List of Regulatory Commitments4. MELLLA+ Risk Evaluation5. Nine Mile Point Unit 2 Power/Flow Operating Map for Current Cycle6. General Electric -Hitachi Affidavit Justifying Withholding Proprietary Information inNEDC-33576P7. Global Nuclear Fuel Affidavit Justifying Withholding Proprietary Information in GNF-0000-0156-7490-RO-P8. NEDC-33576NP, Safety Analysis Report for Nine Mile Point Unit 2 Maximum Extended Load LineLimit Analysis Plus (Non-proprietary)9. Global Nuclear Fuel Report GNF-0000-0 1 56-7490-RO-NP, "GNF Additional Information Regardingthe Requested Change to the Technical Specification SLMCPR," dated August 26, 2013 (Non-proprietary)10. NEDC-33576P, Safety Analysis Report for Nine Mile Point Unit 2 Maximum Extended Load LineLimit Analysis Plus (Proprietary)11. Global Nuclear Fuel Report GNF-0000-0 1 56-7490-RO-P, "GNF Additional Information Regardingthe Requested Change to the Technical Specification SLMCPR," dated August 26, 2013 (Proprietary)ii of ii ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE1.0 SUMMARY DESCRIPTIONThis evaluation supports a request to amend Renewed Operating License (OL) NPF-69 for NineMile Point Unit 2 (NMP2). The proposed amendment includes supporting changes to the NMP2Technical Specifications (TSs) necessary to: 1) implement the Maximum Extended Load LineLimit Analysis Plus (MELLLA+) expanded operating domain; 2) change the stability solution toDetect and Suppress Solution -Confirmation Density (DSS-CD); 3) use the TRACG04 analysiscode; and 4) incr.ease the isotpic en.rihment of boro.n 10 in the sodium pentabor.. e so!uti.nutilized in the Standby Liquid Control System (SLS); and 5) increase the Safety Limit MinimumCritical Power Ratio (SLMCPR) for two recirculation loops in operation.The following is a list of the proposed changes to the NMP2 TSs:* Revise Safety Limit (SL) 2.1.1.2 by increasing the SLMCPR for two recirculation loops inoperation from > 1.07 to > 1.09" Revise the acceptance criterion in TS 3.1.7, "Standby Liquid Control (SLC) System,"Surveillance Requirement (SR) 3.1.7.7 by increasing the discharge pressure from> 1,327 pounds per square inch gauge (psig) to > 1,335 psig" Revise the aeceptance cr-iter-ion int TS SR 3.1.7.10 by inremasing the sodium pentaborateboron 10 enr-iehment Fcquir-ement from -&#xfd; 25 atom percent to &#xfd;! 92 atom percent, and make-aCorespending e hange in Ti S Figure 3.1.7 1, "Sodium Pentabortec SolutieoVolumne/Conentration" Requiements"" Revise TS Figure 3.1.7 I to account for the deer-ease in the miniftmum volume of the SLS t"nfrom 4,558.6 gallons and 1,288 gallons at sodium pentabConte ionoentations of 13.6oT and14.49%, respectively, to 1,600 gallons and 1,530 gallons at sodium pentaborete concentrationsof 13.6%4 and 14.41%, r-espectively* Change the Required Actions for Condition F of TS 3.3.1.1, "Reactor Protection System(RPS) Instrumentation"*Change Condition G of TS 3.3. 1.1* Add new Conditions J and K to TS 3.3.1.1* Correct an editorial error in Note 3 to TS SR 3.3.1.1.13 (i.e., "ORRM" is changed to"OPRM")* Eliminate TS SR 3.3.1.1.16 and references to it in TS Table 3.3.1.1-1, "Reactor ProtectionSystem Instrumentation"* Change the allowable value (AV) for TS Table 3.3.1.1-1, Function 2.b, Average PowerRange Monitor (APRM) -Flow Biased Simulated Thermal Power (STP) -Upscale from"5 0.55W+60.5% [Rated Thermal Power] RTP and 5 115.5% RTP" to"< 0.61W + 63.4% RTP and < 115.5% RTP"* Add a new note to TS Table 3.3.1.1-1, Function 2.b that requires the Flow Biased SimulatedThermal Power -Upscale scram setpoint to be reset to the values defined by the CoreOperating Limits Report (COLR) to implement the Automated Backup Stability Protection(BSP) Scram Region in accordance with Required Action F-,2-4F.2 of TS 3.3.1.1I of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE" Add a new note to TS Table 3.3.1.1-1, Function 2.e, Oscillation Power Range Monitor(OPRM) -Upscale to denote that following implementation of DSS-CD, DSS-CD is notrequired to be armed while in the DSS-CD Armed Region during the first reactor startup andduring the first controlled shutdown that passes completely through the DSS-CD ArmedRegion. However, DSS-CD is considered operable and capable of automatically arming foroperation at recirculation drive flow rates above the DSS-CD Armed Region* Change the mode of applicability for TS Table 3.3.1.1-1, Function 2.e, OPRM-Upscale fromMode I to > 18% RTP" Change the allowable value for TS Table 3.3.1.1-1, Function 2.e from "As specified in theCOLR" to "NA", Add a pr-.hibiti'n t. TS Limiting Condition for Operation (LCO) 3.4.1, "Recirculation LoopsOperating," is modified to prohibittimhat-pfhbits operation in the Maximum Extended LoadLine Limit Analysis (MELLLA) domain or MELLLA+ expanded operating domain asdefined in the COLR when in operation with a single recirculation loop* Add Required Action B.2 to TS 3.4.1 to identify that intentional operation in the MELLLAdomain or MELLLA+ domain as defined in the COLR is prohibited when a recirculationloop is declared "not in operation" due to a recirculation loop flow mismatch not within limits" Revise TS 5.6.5.a.4 to replace "Reactor Protection System Instrumentation Setpoint for theOPRM -Upscale Function Allowable Value for Specification 3.3.1.1" with "The ManualBackup Stability Protection (BSP) Scram Region (Region I), the Manual BSP ControlledEntry Region (Region II), the modified APRM Simulated Thermal Power -High setpointsused in the OPRM (Function 2.e), Automated BSP Scram Region, and the BSP Boundary forSpecification 3.3.1.1"" Add TS 5.6.8, "OPRM Report," to define the contents of the report required by new RequiredAction F-.24F.3 of TS 3.3.1.1Nuclear Regulatory Commission (NRC) approval of the requested operating domain expansionwill allow NMP2 to implement operational changes that will increase operational flexibility forpower maneuvering, compensate for fuel depletion, and maintain efficient power distribution inthe reactor core without the need for more frequent rod pattern changes. MELLLA+ supportsoperation of NMP2 at Current Licensed Thermal Power (CLTP) of 3,988 Megawatts -Thermal(MWth) with core flow as low as 85% of rated core flow. By operating in the MELLLA+ domain,a significantly lower number of control rod movements will be required than in the presentoperating domain. This represents a significant improvement in operating flexibility. It alsoprovides safer operation, because reducing the number of control rod manipulations:(a) minimizes the likelihood of fuel failures and (b) reduces the likelihood of accidents initiatedby reactor maneuvers required to achieve an operating condition where control rods can bewithdrawn.Attachments 8 and 10 provide the non-proprietary and proprietary versions of the MELLLA+Safety Analysis Report (MELLLA+ SAR), respectively. The MELLLA+ SAR follows theguidelines contained in GE-Hitachi Nuclear Energy Americas (GEH) Licensing Topical Report(LTR) NEDC-33006P-A, Revision 3, "Maximum Extended Load Line Limit Analysis Plus"(MELLLA+ LTR) (Reference 1). The MELLLA+ SAR provides the technical bases for thisrequest and contains an integrated summary of the results of the underlying safety analyses andevaluations performed specifically for the NMP2 expanded operating domain.The MELLLA+ SAR also provides the analyses to change the NMP2 stability solution fromOption III to DSS-CD and use the GEH analysis code TRACG04. DSS-CD as required by theMELLLA+ LTR Safety Evaluation Report. DSS-CD is being implemented using the guidelines2 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEcontained in GEH LTR NEDC-33075P-A, Revision 7, "General Electric Boiling Water ReactorDetect and Suppress Solution -Confirmation Density," (Reference 2). The use of TRACG04 isbeing implemented using the guidelines contained in GEH LTR "DSS-CD TRACG Application,"NEDE-33147P-A, Revision 4, August 2013 (Reference 3).The proposed change to the SLMCPR value for two recirculation loops in operation is based onan analysis performed by Global Nuclear Fuel (GNF) for NMP2 during Cycle 15 operations withMELLLA+ conditions. The GNF report, GNF-0000-0156-7490-RO, "GNF AdditionalInformation Regarding the Requested Changes to the Technical Specification SLMCPR," datedAugust 26, 2013, supports changing the two recirculation loops in operation value of SLMCPRfrom > 1.07 to >_ 1.09, and maintaining the single recirculation loop in operation value ofSLMCPR at > 1.09. These values are based on NRC approved methods and procedures.Attachments 9 and 11 of this Enclosure provide non-proprietary and proprietary versions of theGNF report, respectively.Attachments 10 and 11 of this Enclosure contain information considered to be proprietary asdefined by 10 CFR 2.390. GEH and GNF, as the owners of the proprietary information inAttachments 10 and 11, respectively, have executed the affidavits provided in Attachments 6and 7 to this Enclosure detailing the reasons for withholding the proprietary information.Attachment 3 delineates the regulatory commitments associated with the proposed change.2.0 DETAILED DESCRIPTION2.1 Background2.1.1 MELLLA+Operation of Boiling Water Reactors (BWRs) requires that reactivity balance be maintained toaccommodate fuel burn-up. BWR operators have two options to maintain this reactivity balance:(a) control rod movements or (b) core flow adjustments. Because of the strong void reactivityfeedback and its distributed effect through the core, flow adjustments are the preferred reactivitycontrol method. Operation at low-flow conditions at rated power level also increases the fuelcapacity factor through spectral shift and the increased flow region compensates for reactivityreduction due to fuel depletion during the operating cycle.At NMP2, an Extended Power Uprate (EPU) was implemented by extending the MELLLAoperating domain up to the EPU power level (3,988 MWth). The extension of the MELLLA lineto EPU power levels reduces the available core flow window. In addition, the increased corepressure drop with EPU limits the recirculation flow capability. Consequently, EPU plantsgenerally operate with a reduced core flow window and compensate for reactivity loss withcontrol rod movement. Operation in the MELLLA+ expanded operating domain will provide alarger core flow window for NMP2.In June 2009, the NRC approved the use of the MELLLA+ LTR (NEDO-33006P-A)(Reference 1) as a basis for MELLLA+ operating domain expansion license amendment requests,subject to limitations specified in the MELLLA+ LTR and in the associated NRC safetyevaluation. The NMP2 request complies with the specified limitations and conditions asdiscussed in Appendix B of Attachments 8 and 10.3 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEIn January 2008, the NRC approved the use of the DSS-CD LTR (NEDC-33075P) as a basis forimplementing DSS-CD as a stability solution to replace the Option III solution in licenseamendment requests, subject to limitations specified in the DSS-CD LTR and in the associatedNRC safety evaluation. The NMP2 request complies with Revision 7 of NEDC-33075P(Reference 2), including the specified limitations and conditions as discussed in Appendix C ofAttachments 8 and 10.The TRACG code for use in DSS-CD applications (NEDE-33147P-A) was approved by NRC inNovember 2007. The NMP2 request complies with Revision 4 of NEDE-33147P-A(Reference 3).In addition, the NRC approved the Applicability of GE Methods to Expanded Operating DomainLicensing Topical Report (NEDC-33173P-A) which imposes limitations and requirements for theuse of GEH Methods in expanded operating domains including power uprates and MELLLA+domains. The NMP2 request complies with Revision 4 of NEDC-33173P-A (Reference 4),including the specified limitations and conditions as discussed in Appendix A of Attachments 8and 10.Detailed evaluations of the reactor, engineered safety features, power conversion, emergencypower, support systems, and design basis accidents were performed and are provided inAttachments 8 and 10. These evaluations demonstrate that NMP2 can safely operate in theMELLLA+ expanded operating domain with DSS-CD as the thermal hydraulic stability solution.2.1.2 Standby Liquid Control System is.topi. Enrichment of Boron 1NMPNS proposes to inr-easc the isotopi. of beroen 10 in the sodium pentaboatesolution used to prcparc the nouttron absortber- solution in the Standby Liquid Control Systom(SLS) to &#xfd;! 92 atom percoent. The proposed berefn 10 enrielifitnt value allows the minimfum netsolution volume storoed in the Slug storage tank to be deer-eased to 1,530 gallons EAt 14.4% sodiumpentaberate eoncentraion and 1,600 gallons at 13.6% sodium pentaborae concontration. Iadditiei-,NMPNS proposes to increase the acceptance criterion for the SLS pump dischargepressure from 1,327 psig to:- 1,335 psig.2.1.3 Safety Limit Minimum Critical Power RatioNMPNS proposes to revise SL 2.1.1.2 by increasing the SLMCPR for two recirculation loops inoperation from > 1.07 to ! 1.09. The proposed change to the SLMCPR value for tworecirculation loops in operation is based on an analysis performed by GNF for NMP2 duringCycle 15 operations with MELLLA+ conditions. The GNF report, GNF-0000-0156-7490-RO-P,"GNF Additional Information Regarding the Requested Changes to the Technical SpecificationSLMCPR," dated August 26, 2013, supports changing the two recirculation loops in operationvalue of SLMCPR from ! 1.07 to ! 1.09, and maintaining the single recirculation loop inoperation value of SLMCPR at &#xfd; 1.09. These values are based on NRC approved methods andprocedures. Attachments 9 and I11 of this Enclosure provide non-proprietary and proprietaryversions of the GNF report, respectively.2.2 Proposed Changes to the Nine Mile Point Unit 2 Technical SpecificationsNMP2 TS changes are required to allow operation in the expanded MELLLA+ operating domain,use of DSS-CD, inerease the isotopic enrzeinhmcnt of boroen 10 in the sodium pentaborate solution4 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEused to prepare the ne.tr.n abs.r.b.r. s.lution in the SLS, and increase the SLMCPR for tworecirculation loops in operation. Attachment 1 of this Enclosure provides a mark-up of the NMP2TS showing the proposed changes. Attachment 2 of this Enclosure provides a mark-up of theNMP2 TS Bases to show the corresponding changes to the TS Bases. Attachment 2 is providedfor information only. A description of each TS change is provided below.Safety Limit 2.1.1.2, Safety Limit Minimum Critical Power RatioSL 2.1.1.2 is revised to increase the SLMCPR for two recirculation loops in operation from> 1.07 to  1.09.TS 3.1.7, Standby Liquid Control (SLC) SystemTS SR 3.1.7.7 is revised to increase the acceptance criterion for the Standby Liquid ControlSystem (SLS) pump discharge pressure from > 1,327 psig to > 1,335 psig.T-S ISR 3.167. 10 is r-evised to iner-ease the boron 10 cnrichmcnt requirement of sodium pcntabrteffrom ? 25 eatm perceent to ? 92 atom perceent. in addition TS Figure 3.1.7 1 is updated to r-eocthe ineroease in the boront 10 cnriehment roquirement.TIS Figure 3.1.7 1, "Sodium Pentab-ratm Solution ati-n Requir.ements," isrevised to aecount for the change in the net volufe in the SL9 tank that arise thecnr-iehment iner-ease. The miniftmum volumne is changed from 4,558.6 gallonts and 4,288 gallons asodium pcntaber-ate coneentrationis of 13.6% and 14.4%, respeetively, to 1,600 gaillons and 1,530gallonis at a sodium pentaberate coneentration of 13.6% and 14.44, r-espeetively.TS 3.3.1.1, Reactor Protection System (RPS) InstrumentationRequired Actions F. 1 and F.2 of TS 3.3.1.1 and their associated Completion Times are replacedwith the following new Required Actions and Completion Times.REQUIRED ACTIONCOMPLETION TIMEF. I Initiate Action to implement the ManualBSP Regions defined in the COLR.ANDF-.2-4-F._2 Implement the Automated BSPScram Region using the modifiedAPRM Simulated Thermal Power -High scram setpoints defined in theCOLR.ANDR.-,F.__3 Initiate action in accordancewith Specification 5.6.8.Immediately12 hours90 dayslmmediately5 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGECondition G is modified to no longer apply in the event a Required Action and associatedCompletion Time of Condition F is not met.New Condition J (see below) is added to address the action to take in the event a Required Actionand associated Completion Time of Condition F is not met.New Condition K (see below) is added to address the action to takeAction and associated Completion Time of Condition J is not met.in the event a RequiredCONDITIONREQUIRED ACTIONCOMPLETION TIMEJ. Required Action andassociated Completion Timeof Condition F not met.J. 1 Initiate action toimplement the ManualBSP regions defined in theCOLR.ANDJ.2 Reduce operation to belowthe BSP Boundary definedin the COLR.ANDJ.3 ---------NOTE-------LCO 3.0.4 is not applicableImmediately12 hours120 daysRestore required channel tooperable.K. Required Action andassociated Completion Timeof Condition J not met.K.1ReducePOWERRTPTHERMALto less than 18%4 hours"ORRM" is changed to "OPRM" in Note 3 to TS SR 3.3.1.1.13.TS SR 3.3.1.1.16 is eliminated, and references to it in TS Table 3.3.1.1-1 are eliminated.TS Table 3.3.1.1-1, Function 2.b, Flow Biased Simulated Thermal Power -Upscale, contains botha flow-biased AV (_ 0.55W + 60.5% RTP) and a fixed AV at 115.5% RTP. The flow-biased AVwill be changed to (< 0.6 1W + 63.4% RTP).A new note is added to TS Table 3.3.1.1-1, Function 2.b that requires the Flow Biased SimulatedThermal Power -Upscale scram setpoint to be reset to the values defined by the COLR toimplement the Automated Backup Stability Protection (BSP) Scram Region in accordance withRequired Action F.2. F.2 of TS 3.3.1.1.A new note is added to TS Table 3.3.1.1-1, Function 2.e, OPRM -Upscale, to denote thatfollowing DSS-CD implementation, DSS-CD is not required to be armed while in the DSS-CDArmed Region during the first reactor startup and during the first controlled shutdown that passes6 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEcompletely through the DSS-CD Armed Region. However, DSS-CD is considered operable andcapable of automatically arming for operation at recirculation drive flow rates above the DSS-CDArmed Region.The mode of applicability for TS Table 3.3.1.1-1, Function 2.e, OPRM-Upscale is changed fromMode I to > 18% RTP.In addition, the allowable value for Function 2.e is changed from "As specified in the COLR" to"4NA."TS 3.4.1, Recirculation Loops, OperatingLCO 3.4.1 is modified to include an additional provision that will prohibit intentional operation inthe MELLLA domain or the MELLLA+ domain as defined in the COLR when only a singlerecirculation loop is in operation. It will state:"... One recirculation loop shall be in operation provided the plant is not operating in theMELLLA or MELLLA+ domain defined in the COLR and provided the following limits areapplied when the associated LCO is applicable:..."A new Required Action B.2 is added to prohibit intentional operation in the MELLLA domain orthe MELLLA+ domain defined in the COLR in the event a recirculation loop is declared to be"not in operation" due to a recirculation loop flow mismatch. The Completion Time for this newRequired Action is 2 hours.TS 5.6.5, Core Operating Limits Report (COLR)TS 5.6.5.a is modified by replacing the reference to "Reactor Protection System InstrumentationSetpoint for the OPRM -Upscale Function Allowable Value for Specification 3.3.1.1" with areference to "The Manual Backup Stability Protection (BSP) Scram Region (Region 1), theManual BSP Controlled Entry Region (Region II), the modified APRM Simulated ThermalPower -High setpoints used in the OPRM (Function 2.e), Automated BSP Scram Region, andthe BSP Boundary for Specification 3.3.1.1 ."TS. 5.6.8, OPRM ReportThe following new report requirement is added as TS 5.6.8, "OPRM Report:""When a report is required by Required Action F-,2-F.3 of TS 3.3.1.1, "RPSInstrumentation," a report shall be submitted within 90 days of ,nteing CONDITION Fthefollowing 90 days. The report shall outline the preplanned means to provide backup stabilityprotection, the cause of the inoperability, and the plans to- and schedule for restoring therequired instrumentation channels to OPERABLE status."The new TS section is numbered TS 5.6.8, because on November 21, 2012, NMP2 submitted aLicense Amendment Request (LAR) to create a new TS section that is numbered TS 5.6.7 for theReactor Coolant System Pressure and Temperature Limits Report (Reference 5). NMPNSanticipates that LAR will be approved by the NRC and implemented at NMP2 prior to approvalof the MELLLA+ LAR. The numbering of TS 5.6.8 is an administrative consideration. The7 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEMELLLA+ LAR is independent of the LAR submitted on November 21, 2012. NRC approval orrejection of Reference 5 would have no technical impact on the MELLLA+ LAR.2.3 Modification SummaryThe MELLLA+ core operating domain expansion does not require major plant hardwaremodifications. The core operating domain expansion involves changes to the core power/flowmap and a small number of setpoints and alarms. Because there are no increases in the operatingpressure, power, steam flow rate, and feedwater flow rate, there are no major modifications toother plant equipment.The stability solution is being changed from Option III to the DSS-CD solution. The DSS-CDsolution algorithm, licensing basis, and application procedures are generically described in theDSS-CD LTR (Reference 2), and are applicable to NMP2. The DSS-CD solution uses the samehardware as the current Option III solution. DSS-CD requires a revision to the existing stabilitysolution software.The boroen 10 enrichmfent in the sodium pentabor-ate solution in the SLS is incr-eased fo! 25 atom pcrccnt to ! 92 atom perceent. The increase in the boroen 10 enriehment in the scdiumpentaborate solution for- the SLS is sufficient to decrease the sodium pentabernte solution velumesterce in tMle SLS storage tank. In a..iti.n, tc I ne SLS pump discharge pressure acceptancecriterion is changed to > 1,335 psig. Changes to instrumentati. n setp..ots will be made toaccotunt for these changes. The increase in the SLMCPR for two recirculation loops in operationdoes not require any physical modifications to structures, systems, or components.


==3.0 TECHNICAL EVALUATION==
==3.0 TECHNICAL EVALUATION==
3.1 MELLLA+Attachments 8 and 10 of this Enclosure provide non-proprietary and proprietary versions of the"Safety Analysis Report for Nine Mile Point Unit 2 Maximum Extended Load Line LimitAnalysis Plus (MELLLA+ SAR)," NEDO-33576NP and NEDC-33576P, respectively. TheMELLLA+ SAR summarizes the results of the significant safety evaluations performed thatjustify the expansion of the core flow operating domain for NMP2. The changes expand theoperating domain in the region of operation with less than rated core flow, but do not increase thelicensed power level or the maximum core flow. The expanded operating domain is identified asMELLLA+.The scope of evaluations required to support the expansion of the core flow operating domain tothe MELLLA+ boundary is contained in NEDC-33006P-A, "Maximum Extended Load LineLimit Analysis Plus," referred to as the MELLLA+ LTR (Reference 1). The MELLLA+ SARprovides a systematic disposition of the MELLLA+ LTR subjects applied to NMP2, includingperformance of plant-specific assessments and confirmation of the applicability of genericassessments to support a MELLLA+ core flow operating domain expansion. The MELLLA+operating domain expansion is applied as an incremental expansion of the operating boundarywithout changing the maximum licensed power, maximum core flow, or the current plant vesseldome pressure. The MELLLA+ SAR supports operation of NMP2 at a licensed thermal power of3,988 MWt with core flow as low as 85% of rated core flow.8 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEThe MELLLA+ core operating domain expansion does not require major plant systemsmodifications. NMP2 will implement the DSS-CD solution in accordance with the applicableLTRs (References 3 and 4), including the applicable limitations and conditions. Implementationof DSS-CD requires a revision to the existing stability solution software.The core operating domain expansion involves changes to the operating power/core flow map andchanges to a small number of instrument setpoints. Because there are no increases in theoperating pressure, power, steam flow rate, and feedwater flow rate, there are no significanteffects on the plant systems outside of the Nuclear Steam Supply System (NSSS). There is apotential increase in the steam moisture content at certain times while operating in theMELLLA+ operating domain. The effects of the potential increase in moisture content on plantsystems have been evaluated and determined to be acceptable. The MELLLA+ operating domainexpansion does not cause additional requirements to be imposed on any of the safety, balance-of-plant, electrical, or auxiliary systems. No changes to the power generation and electricaldistribution systems are required as a result of the MELLLA+ operating domain expansion.This report also addresses applicable limitations and conditions as described in the MELLLA+LTR SER for the GEH LTR NEDC-33173P, "Applicability of GE Methods to ExpandedOperating Domains" (Methods LTR SER) (Reference 4). A complete listing of the applicableLTR SER limitations and conditions and the sections of the MELLLA+ SAR which address themare presented in Appendices A, B, and C of the MELLLA+ SAR.Only previously NRC-approved or industry-accepted methods were used for the analyses ofaccidents and transients. Therefore, because the safety analysis methods have been previouslyaddressed, the details of the methods are not presented for review and approval in the MELLLA+SAR. Also, event and analysis descriptions that are already provided in other licensing reports orthe NMP2 Updated Safety Analysis Report (USAR) are not repeated within the MELLLA+ SAR.Evaluations of the reactor core and fuel performance, reactor coolant and connected systems,engineered safety features, instrumentation and control, electrical power and auxiliary systems,power conversion systems, radwaste systems and radiation sources, reactor safety performanceevaluations were performed. The MELLLA+ SAR summarizes the results of the evaluations thatjustify the MELLLA+ operating domain expansion to a minimum core flow rate of 85% of ratedcore flow at 100% RTP.Section 11.3.1 of Attachments 8 and 10 provides a summary of the modifications that will berequired to implement the MELLLA+ operating domain, DSS-CD, and the changes to the SLS.Section 11.3.2 of Attachments 8 and 10 provides a summary of the MELLLA+ issues including adiscussion of the MELLLA+ analysis basis, fuel thermal limits, makeup water sources, designbasis accidents, challenges to fuel, challenges to the containment, design basis accidentradiological consequences, anticipated operational occurrence analyses, combined effects,non-Loss of Coolant Accident (LOCA) radiological release accidents, equipment qualification,balance-of-plant, and environmental consequences.An assessment of the risk increase, including core damage frequency (CDF) and large earlyrelease frequency (LERF) associated with operation in the MELLLA+ operating domain isprovided in Attachment 4 of this Enclosure and Section 10.5 of Attachments 8 and 10 of thisEnclosure. The estimated risk increase for at-power events due to MELLLA+ is a delta CDF of1E-8 and delta LERF of 3E-9. This represents a very small risk change in RG 1.1749 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE(Reference 6). Based on these results, the proposed MELLLA+ operating domain is acceptableon a risk basis.3.2 DSS-CDThe long-term stability solution is being changed from the currently approved Option III solutionto DSS-CD. The DSS-CD solution algorithm, licensing basis, and application procedures aregenerically described in NEDC-33075P-A (Reference 2) and NEDE-33147P-A (Reference 3),and are applicable to NMP2 including any limitations and conditions associated with their useand approval. Section 2.4 of the MELLLA+ SAR (Attachments 8 and 10) addresses the changeto the DSS-CD stability solution. In addition, a complete listing of the required DSS-CD SERand limitations and conditions and the sections of the MELLLA+ SAR which address them ispresented in Appendix C of the MELLLA+ SAR, respectively.DSS-CD is designed to identify the power oscillation upon inception and initiate control rodinsertion (scram) to terminate the oscillations prior to any significant amplitude growth exceedingthe applicable safety limits. DSS-CD is based on the same hardware design as Option III.However, it introduces an enhanced detection algorithm that detects the inception of poweroscillations and generates an earlier power suppression trip signal exclusively based .n........ :.... :ri.d r-..gni. .The existing Option III algorithms are retained (withgeneric setpoints) to provide defense-in-depth protection for unanticipated reactor instabilityevents.3.3 Standby Liquid Control System Boron 10 En-.hmen-The SLS is described in Section 9.3.5 of the NMP2 USAR. The system provides a backupcapability for shutting down the reactor. The SLS is needed only in the event that sufficientcontrol rods cannot be inserted into the reactor core to accomplish shutdown and cooldown in thenormal manner. To accomplish this function, the SLS injects a sodium pentaborate solution intothe reactor. The SLS consists of a boron solution storage tank, two positive displacement pumps,two explosive valves (provided in parallel for redundancy), and associated piping and valves usedto transfer borated water from the storage tank to the reactor pressure vessel (RPV). The boratedwater solution is discharged into the RPV through the high pressure core spray sparger.The specified neutron absorber solution is sodium pentaborate. It is prepared by dissolvinggranularly-enriched sodium pentaborate in demineralized water (NMP2 USAR 9.3.5.2). Thesodium pentaborate solution is discharged radially over the top of the core through the HighPressure Core Spray (HPCS) sparger. The boron absorbs thermal neutrons and thereby terminatesthe nuclear fission chain reaction in the uranium fuel. The sodium pentaborate also acts as abuffer to maintain the suppression pool pH at or above 7.0 to prevent the re-evolution of iodine,when mixed in the suppression pool following a LOCA accompanied by significant fuel damage(NMP2 USAR Section 9.3.5.1).3.3.1 Reactor Boron Cold Shutdown Concentration RequirementsThe reactor boron concentration requirements for achieving cold shutdown (780 parts per million(ppm) natural boron) is not increased for MELLLA+, because there is no change in fuel type andno change to the operating cycle. The total weight of boron-10 required for cold shutdown(including the 25% margin) does change for MELLLA+, because of a conservative increase inthe assumed weight of the reactor coolant in the applicable analysis.10 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEThe shutdown margin is calculated for each plant reload and is documented in the SupplementalReload Licensing Report (SRLR).Ii ,"b at'-N 1 4 lift Z-- * ----AI nSLflpfl~C in imron ~u ~nricnmcnr in ~ou:um rcnuwarurc10 CFR 50.62(e)(4) rvauir-es:"Each boiling water -retor must have a standby liquid contrl system (SCS) withtheapability of injecting into the reacter- prcssurce vessel a bor-ated water- solutiont at such a florate, level of bor-on concentfation and borefn 10 isotope enrichmcnt, and aecountfing or-eactor pressure vessel voluime, that the resulting reaetivity entrl is at least equivalent tothat r-esulting from injection of 86 gallons per- minuite of 13 weight perceent sodiumpentaborate dceahydrate solutioin at the nAtur-al boroen 10 isotope abundance into a 251 inchinside diameter- r.ea.t. r pressure .ss. l for- a gien ... design..."The NRC approved licensing topical r-eport NEDE 31096 A (Rcfer-enee 7) provides a method*bwhich the boront equivalency r-equirement of 10 CFR 50.62(e)(4) can be demonefstrated.Equation 1 1 of that docuiment was used to demonstfate injectiont capacity equivalency as"elews.(Q/'86) x (M2511') x (C-'13) x (9/'49.8) 1 1Where-Q -expeeted SLS flew rate (gpmn)A425-L mass of water in the rcaetfr ivessel and recircuat~ion system at hot ratedconditions (Ibs) for- a 251 inceh diameter- vessel reference plant-l mass of water in the NMP2 reactor- vessel and r-eeir-eulation system at hot ratedeendifiefts Ibs)-sodium pentaborae solution concentreation (%494)E boron 10 isotope enrichment (atom percent)NMP2 is equipped with a 251 inch diameter reactor- vessel (NMP2 USAR Section 15.)assumptions utilized int the analyses of the changes to the SLS.Substituting the cuffent values definted in Table I in the above equationt yields:82.4/86 X 1 X 13.6413 X 25/19.8 -l.27&#xfd;--4Substituting the new values definted in Table 1 into the above equation yields:!80/86 X 1 X 13.6/13 X 92/19.8 -4.52> 1 for 13.6 v,0/_)This that the equivalent control capacity requirement of 10 CFR 50.62(e).4)is m et, when the changes to the SLS flow rate and the boront 10 isotope em-iehment are inceluded.int addition, the control margin ince s. This is due to -nefeasing the boroen 10 enrichment tennin the equation by a factor- of3.68 (i.e., 92'25 -3.68).-11 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEf41re: -T.I-. .....requre c ..I sm)I -1pump pto nave a o eow ra.. A at ii ' ...gpml.Maintaining the TS SR 3. 1.7.7 aeceptance cr-iter-ia for SLC pump flow rate at 41.2 gpmf providesmar~gin with r-espcct to the r~equired flow rate for- AT-WS mitigation. This issue has beeaddressed for- euffent eperation via the NMP-NS Correcetive Action System.As defined in Table 1, the analyzed SLS injection flow Fate is r-educed to 80 gpm flow r-ate otwo SLC System pumps in .perati.n to a ..ount for- diluti.n ..fe.ts identified by GEH4 SafetyCommunicationt 10 13, Standby Liquid C.ntrol System Dilution Flew, with additional margin.Table 1 Assumptions rega.rding SLS P.rforman.Pafametef Units Current Vaulue New-ValueReactor boron ..n.. ntatin for pp 7. ...0cold shutdown (natur-al boront)Maximuma allowable solutionf V.+% 144. -144conceentrationtMinimum allowable solutio 4 A 4 6conccntfatienSolution eoncentration assumed in Y% 4-3 4446ATWS anal-isi__Mfinimumn boroen 10 enrichment foi Aom 2 2ATWS aftaly~fisDesigni-SLS pumip flew Faite 813M 4-54-Minimum SLS pump flow rate as gPm" 44l-2 4442defined in T-S 3.1.7Sing pump flow rate (two. pup i pin 92-4 80)3-.33.3.2 Change in SLS Pump Discharge Pressure Acceptance CriterionTS SR 3.1.7.7 is revised to increase the acceptance criterion for the SLS pump discharge pressurefrom >_ 1,327 psig to > 1,335 psig. This change is required due to the increase in the peak upperplenum pressure after SLS pump startup to 1,241 pounds per square inch -absolute (psia) asidentified in Tables 9-4 and 9-7 of Attachments 8 and 10 of this Enclosure. Currently, the peakupper plenum pressure after SLS pump startup is 1,236 psia. Thus, the ATWS analysis forMELLLA+ establishes a pressure differential of five psi. The SLS pump discharge pressureacceptance criterion in TS SR 3.1.7.7 is increased by eight psig to address the increase in theupper plenum pressure and provide an additional three psi margin.33.43.3.3 Anticipated Transient without SCRAMSection 9.3.1 of the MELLLA+ SAR (Attachments 8 and 10 of this Enclosure) provides asummary of the plant-specific analyses of Anticipated Transients Without Scram (ATWS) todemonstrate that the ATWS acceptance criteria are met for operation in the MELLLA+ operatingdomain. NMP2 meets the ATWS mitigation requirements in 10 CFR 50.62 for an alternate rodinsertion (ARI) system, SLS boron injection equivalent to 86 gpm, and automatic RPT logic. The12 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEplant-specific ATWS analyses take credit for the ATWS-RPT and SLS. However, ARI is notcredited.Section 9.3.1.1 of the MELLLA+ SAR (Attachments 8 and 10 of this Enclosure) provides alicensing basis ODYN ATWS analysis that demonstrates that the ATWS acceptance criteriawould be met in the event of a NMP2 response to an ATWS event initiated in the MELLLA+operating domain.In addition, a plant-specific ATWS analysis was performed at MELLLA+ conditions thatassumed operation of a single SLS pump. The analysis and the results are discussed in Section9.3.1.1.1 of the MELLLA+ SAR (Attachments 8 and 10 of this Enclosure). It concludes: "AllATWS acceptance criteria are met at MELLLA+ conditions with only a single SLS pumpoperating."3.3.5 Suppression Pool BufferingIn supp.rt of the Alte.ate Source Term (AST-) meth.dology, the SLS also provides suppr.ession.peel buffer-ing following a LOCA accompanied by significant fuel damage, preventing r-eevelution of iodine from the suppr-ession peol by maintaining the pool pH4 above 7.0. Seetion9.3.5.1 of the NMP2 USAR requires a sufficient coneentration and quaftity of sodiumpentaborate to be available for injcetion into the reaetor vessel to controel pH4 in the suppr-esiopool for 30 days fllwing a DBA LOCAi.The r1eduetiont int the mi r u :ired solution volume results in a reduction in. the e ssolutiont available for- injectionl to maneneppeso ol p14I  7.0 for- 30 days post LOCA.The minimum" sodium pentabor-ate solution vouei rqie for injection post LOCA foradequate p14 eontrol is 1,065 lons at the limiting t.e., a sodium pentaborateconcentration of 13.6%4). The minimum requir-ed tank vollumie at4 al concentration of 13.6 %ireduced frim 4,558.6 gallons to 1,600 gallons. While this does r educe the amount of cxeavailable solutiont, adequate margin is maintaincd to ensure that the SLS can perform its requiredAST support funetion.The proposed bor-on 10 enr-ichment changes do not impaet the capability to achieve and maintaina pH4 above 7.0 in the suppression pool following a LOCA, because the chemnical pr-opeffies nonenetration of the sodium pentaborate solution injected into the suppr-essin pool will remainthe same. Given the rduced volume of solution that wil be available, ther will be a two hourr-eduction int the maximum tome available to add boron to the supeso pool to ananpabove 7.0 (noeminal time based on low level alann is within 22- housvess h-- ~ttieowithin 24 hours). Are.vie. of the Emergency Operating Procedures confirmed that the sodiumpentaborate solutiont would be injeeted within 30 minutes following the eccunfeftee of LOCA.The maaximutm 22 hour- timne period proevides a large margin to the minimm requremet formanual operator- action toinec the sodium pentaborate solution of 30 minuatesq. In addition, theSuppression pool pisntepected to drop bielw:7 foseelda.Section 9.3.5.3 of the NMP2 USAR delineates that ontly one of the two SL.S loops was assumefor suppr-ession pool p1H control operation. The proeposed changes to the SLS do not affect thI., ...t (.... C' I1.-I...-- -It% A L ... .+I^ : l ..,i;. .~~~~0~~ ---- ---- --- ---I--I-- --------------------0~.... ' .. .... ....J .... ..... .rrI-of313 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE"J f J"QI- --3.3.C n~ige tin 818 Sterage I unit Soluiuon -VolumeThe pr-opescd bOroni 10 enriehment value allows the minimum solutiont Volumfe stor-ed in theSLsterage tank th be dedreased to 1,530 gallons at a sodium.. pentabrate e.ne.ntr..in of 14.4% and1,600 gallons at a sodium pentablrVate eoinentraion of 13.6%. The mark uip of NMP2 TS Figt3.1.7 1 promvided inf Attaclhment 1 mf this Entlesur delineates thce preposed hange in the net SLSstorage fankl solutioin volume.The required minimum lumes for- the 13.6 p%49 and 14.4 mar solutin vplumres were der tby deteamininag the m einimum salution veluifne and then increasing the allumne to acotun frei1) the dead vsluine not pumped in the reator0 that remains in the SLS and 14PCS piping;a2) instrument accuracy.The minimum net solution volume for- injection meets all consider-ations for AT-WS boroninjection raes, AST suppression pool pH- control, and assures tha the r-eactor- core boronconcentr-ation will be greater- than :780 ppm natural boron equivalent3-.343.3.3 SLS Pump Relief Valve Setpoint MarginThe SLS pump relief valve setpoint margin is the difference between the relief valve nominalsetpoint and the maximum SLS pump discharge pressure. A margin of 78 psi provides sufficientmargin against inadvertent relief valve lifting. The 78 psi is based on an allowance for the reliefvalve setpoint drift (typically 3% (3% of 1,600 psi = 48 psi)) and SLS pump pressure pulsations(30 psi).For MELLLA+ operation during the limiting ATWS event, the relief valve setpoint margin is205.7 psi. This margin is based on a SLS pump relief valve setpoint of 1552 psig (1600 psig -3% tolerance (i.e., 48 psig)) and subtracting a SLS pump discharge pressure of 1346.3 psig (i.e.,1552 psig -1346.3 psig = 205.7 psi). The margin reduces to 175.7 psi if 30 psi for SLS pumppressure pulsations is taken into consideration (i.e., 205.7 psi -30 psi = 175.7 psi).3.3.8 Net Positive Suction Head Available (NIPSH1) for SbS P-unqwrThe propesed changes include a reduction in the minimum volume for- the SLS storage tank. Tiresults inl at eduction in the swaic head available to provide Net Positive Suction Read (NPSH4Yfotthe SLS pumps. The calculation that determnines the SLS pumnp NPSNft did not take any cedifor- the staic head abo-ve the SLS storage tank zero level. The mninimum tank level cof:Fespondinfgto the minimum net volume pefmitted by the proposed change to Figure 3.1.7 1 is greater- thanthree feet above tankl zero.3.4 Safety Limit Minimum Critical Power RatioCycle specific transient analyses are performed to determine the required SLMCPR and thechange in Critical Power Ratio (CPR) [ACPR] for specific transients. To ensure that adequatemargin is maintained, a design requirement based on a statistical analysis was selected, in thatmoderate frequency transients caused by a single operator error or equipment malfunction shallbe limited such that, considering uncertainties in manufacturing and monitoring the coreoperating state, at least 99.9% of the fuel rods would be expected to avoid boiling transition. Thelowest allowable transient MCPR limit which meets the design requirement is termed the fuelcladding integrity SLMCPR.14 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGENUREG-0800, Standard Review Plan, Section 4.4, "Thermal and Hydraulic Design," AcceptanceCriterion No. I.B, states, in part, that the limiting (minimum) value of CPR is to be establishedsuch that at least 99.9% of the fuel rods in the core would not be expected to experience departurefrom nucleate boiling during normal operation or anticipated operational occurrences.A cycle specific Operating Limit MCPR (OLMCPR) is established to provide adequate assurancethat the fuel cladding integrity SLMCPR is not exceeded for any anticipated operationaltransients. The OLMCPR is obtained by adding the maximum value of ACPR for the mostlimiting transient postulated to occur at the plant to the fuel cladding integrity SLMCPR.3.4.1 Analytical Methods, Standards, Data and ResultsNMPNS proposes to revise SL 2.1.1.2 by increasing the SLMCPR for two recirculation loops inoperation from >_ 1.07 to > 1.09. The proposed change to the SLMCPR value for tworecirculation loops in operation is based on an analysis performed by GNF for NMP2 duringCycle 15 operations with MELLLA+ conditions. The GNF report, GNF-0000-0156-7490-RO-P,"GNF Additional Information Regarding the Requested Changes to the Technical SpecificationSLMCPR," dated August 26, 2013, supports changing the two recirculation loops in operationvalue of SLMCPR from >_ 1.07 to >_ 1.09, and maintaining the single recirculation loop inoperation value of SLMCPR at > 1.09. These values are based on NRC approved methods andprocedures. Attachments 9 and 11 of this Enclosure provide non-proprietary and proprietaryversions of the GNF report, respectively.GNF performed the SLMCPR calculation in accordance with Revision 19 of NEDE-2401 1-P-A,"General Electric Standard Application for Reactor Fuel," (Reference 8) using the followingNRC-approved methodologies and uncertainties:" NEDC-32601P-A, "Methodology and Uncertainties for Safety Limit MCPREvaluations," (August 1999) (Reference 9).* NEDC-32694P-A, "Power Distribution Uncertainties for Safety Limit MCPREvaluations" (August 1999) (Reference 10).* NEDC-32505P-A, "R-Factor Calculation Method for GEl 1, GE12 and GE13 Fuel,"(Revision 1, July 1999) (Reference 11).Section 2.9 of Attachments 9 and 11 of this Enclosure require NMPNS to "provide the currentand previous cycle power/flow map in a separate attachment." Figure 1-1 of Attachments 8 and10 of this Enclosure provide the power/flow operating map for MELLLA+. This will be thepower/flow map for NMP2 operations in Cycle 15 following NRC approval of this LicenseAmendment Request. Attachment 5 of this Enclosure provides the NMP2 power/flow operatingmap for the current operating cycle.3.4.2 Major Contributors to SLMCPR ChangeIn general, the calculated safety limit is dominated by two key parameters: (1) flatness of the corebundle-by-bundle MCPR distribution, and (2) flatness of the bundle pin-by-pin power/R-Factordistribution. Greater flatness in either parameter yields more rods susceptible to boiling transitionand thus a higher calculated SLMCPR. The MCPR Importance Parameter (MIP) measures the15 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEcore bundle-by-bundle MCPR distribution and the R-Factor Importance Parameter (RIP)measures the bundle pin-by-pin power/R-Factor distribution. The impact of the fuel loadingpattern on the calculated two recirculation loops in operation SLMCPR has been correlated to theparameter MIPRIP, which combines the MIP and RIP values.Another factor besides core MCPR distribution or bundle R-factor distribution that significantlyimpacts the SLMCPR is the expansion of the analysis domain that comes with the initialapplication of MELLLA+. The rated power / minimum core flow point is analyzed at a lowercore flow (than without MELLLA+) using increased uncertainties that tend to increase theSLMCPR. Also, a new point at off-rated power / off-rated flow was analyzed using the increaseduncertainties.Table 3 of the GNF analysis (Attachments 9 and 11 of this Enclosure) presents the MIP and RIPparameters for the previous cycle and the current cycle along with the two recirculation loops inoperation SLMCPR estimates using MIPRIP correlations. In addition, Table 3 of the GNFanalysis provided in Attachments 9 and 1 1 presents estimated impacts on the two recirculationloops in operation SLMCPR due to methodology deviations, penalties, and/or uncertaintydeviations from approved values. Based on the MIPRIP correlation and any impacts due todeviations from approved values, a final estimated two loops in operation SLMCPR isdetermined. Section 2.2 of the GNF analysis (Attachments 9 and 11 of this Enclosure) provides adetailed discussion of the items in Table 3 of the GNF analysis (Attachments 9 and 11 of thisEnclosure) that result in the increase in the estimated SLMCPR.3.4.3 Considerations Addressed in the GNF Analysis Regarding R-Factor, Core Flow Rate andRandom Effective Tip Reading, and Fuel Axial Power Shape PenaltySection 2.2.1 of the GNF analysis provides a discussion that justifies an increase in the R-Factoruncertainty value. GNF states that it generically increased the GEXL R-Factor uncertainty toaccount for an increase in channel bow due to the emerging unforeseen phenomena called controlblade shadow corrosion-induced channel bow, which is not accounted for in the channel bowuncertainty component of the approved R-Factor uncertainty. NMP2 has experienced controlblade shadow corrosion-induced channel bow. Accounting for the control blade shadowcorrosion-induced channel bow, the NMP2 Cycle 15 analysis shows an expected channel bowuncertainty which is bounded by the increased GEXL R-Factor uncertainty. Thus, the use of theincreased GEXL RFactor uncertainty value adequately accounts for the expected control bladeshadow corrosion-induced channel bow for NMP2 Cycle 15.Section 2.2.2 of the GNF analysis provides a discussion that identifies that the uncertainty valuesfor the core flow rate and the random effective tip reading in the two recirculation loops inoperation calculation were conservatively adjusted by using the single recirculation loop inoperation uncertainty values. The GNF analysis states the treatment of the core flow and randomeffective TIP reading uncertainties is based on the assumption that the signal to noise ratiodeteriorates as core flow is reduced.Section 2.4 of the GNF analysis provides a discussion regarding higher uncertainties and non-conservative bases in the GEXL correlations for the various types of axial power shapes. GNFdetermined that no power shape penalties were required to be applied to the calculated NMP2Cycle 15 SLMCPR values.16 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE3.4.4 ConclusionThe proposed change to revise SL 2.1.1.2 by increasing the SLMCPR for two recirculation loopsin operation from > 1.07 to > 1.09 is acceptable, and continues to maintain the same level ofsafety as the current licensing basis.3.5 NMP2 TS ChangesTable 2 defines the affected NMP2 TS, describes the change, and defines the supportingAttachment to this Enclosure that supports the TS Change.Table 2 -Changes to NMP2 Technical SpecificationsNMP2 TS Description of the Change SupportingAttachmentSL 2.1.1.2 Increase the SLMCPR for two recirculation loops in Attachments 9 and 11operation from > 1.07 to > 1.09TS 3.1.7 -Increase the SLS pump discharge pressure from Section 6.5.3 ofSR 3.1.7.7 > 1,327 psig to >_ 1,335 psig Attachments 8 and 10TS-83.b.7- hinr-easing the sodium pentaberate ber-on0 Seetiont 6.5.1 ofSR 3.A.7. enr-iehment rzguir-ement from ! 25 atom pcr-eent toAttachments 8 and 10and S92 .tem.peree.T-S Figure Reducig the minimum net vlume n to 1,600 gallt Section 6.5.1 of3.i.74i and 1,530 gallens at Scdium pentabonuats i ngentrmaiens Attachments 8 and 10of 13.6% Sad 14.4%, ersperticlhyT-S-Figure lneroasing the COdi ;an pcItabniate brcon 1 Seetio 6.5.1 in3. i7 I enriehmcn egunt tfrom ! 25 atom pereent toAttachiments 8 and 1092acteforPer."itTS 3.3.1.1 The Required Actions for Condition F are modified to: Complies with DSS-1) Initiate Action to implement the Manual BSP CD LTRRegions defined in the COLR; 2) Implement the Section 2.4 ofAutomated BSP Scram Region using the modified Attachments 8 and 10APRM Simulated Thermal Power -High scramsetpoints defined in the COLR; and 3) Initiate action inaccordance with Specification 5.6.8TS 3.3.1.1 Condition G is modified to no longer apply in the event Complies with DSS-a Required Action and associated Completion Time of CD LTRCondition F is not met. Section 2.4 ofAttachments 8 and 10TS 3.3. 1.1 New Condition J is added to address the action to take Complies with DSS-in the event a Required Action and associated CD LTRCompletion Time of Condition F is not met. Section 2.4 ofAttachments 8 and 1017 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGETable 2 -Changes to NMIP2 Technical SpecificationsNMP2 TS Description of the Change SupportingAttachmentTS 3.3.1.1 New Condition K is added to address the action to take Complies with DSS-in the event a Required Action and associated CD LTRCompletion Time of Condition J is not met. Section 2.4 ofAttachments 8 and 10TS SR Correct an editorial error in Note 3 (i.e., ORRM is Editorial correction3.3.1.1.13 changed to OPRM)TS SR Eliminate TS SR 3.3.1.1.16 and references to it in TS Complies with DSS-3.3.1.1.16 Table 3.3.1.1-1 CD LTRand TS Section 2.4 ofTable Attachments 8 and 103.3.1.1-1TS Table Change the AV for APRM -Flow Biased STP -Section 5.3.1 of3.3.1.1-1, Upscale from ":50.55W+60.5% RTP and < 115.5% Attachments 8 and 10Function 2.b RTP" to "< 0.61W + 63.4% RTP and < 115.5% RTP"TS Table Add a new note that requires the Flow Biased Complies with DSS-3.3.1.1-1, Simulated Thermal Power -Upscale scram setpoint to CD LTRFunction 2.b be reset to the values defined by the COLR to Section 2.4 ofimplement the Automated BSP Scram Region in Attachments 8 and 10accordance with Required Action F--.4F.2 of TS3.3.1.1TS Table Add a new note for Function 2.e, OPRM -Upscale, to Complies with DSS-3.3.1.1-1, denote that following implementation of DSS-CD, CD LTRFunction 2.e DSS-CD is not required to be armed while in the DSS- Section 2.4 ofCD Armed Region during the first reactor startup and Attachments 8 and 10during the first controlled shutdown that passescompletely through the DSS-CD Armed Region.However, DSS-CD is considered operable and capableof automatically arming for operation at recirculationdrive flow rates above the DSS-CD Armed RegionTS Table Change the mode of applicability for TS Table 3.3.1.1- Complies with DSS-3.3.1.1-1, 1, Function 2.e, OPRM-Upscale from Mode 1 to > 18% CD LTRFunction 2.e RTP. Section 2.4 ofAttachments 8 and 10TS Table Change the allowable value from "As specified in the Complies with DSS-3.3.1.1-1, COLR" to "NA" CD LTRFunction 2.e Section 2.4 ofAttachments 8 and 1018 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGETable 2 -Changes to NMP2 Technical SpecificationsNMP2 TS Description of the Change SupportingAttachmentTS.LCO Add a new requirement that pr...i..bModify the LCO Complies with DSS-3.4.1 to prohibit operation in the MELLLA domain or CD LTRMELLLA+ expanded operating domain as defined inthe COLR when in operation with a single recirculation Sections 1.2.4 andloop 3.6.3 2.4 ofAttachments 8 and 10address thatMELLLA+ is notanalyzed for singleloop operationIn addition, NMP2does not currentlypermit single loopoperation while in theMELLLA domain,because it is notanalyzed.TS 3.4.1, Add Required Action B.2 to identify that intentional Complies with DSS-Condition B operation in the MELLLA domain or MELLLA+ CD LTRdomain as defined in the COLR is prohibited when arecirculation loop is declared "not in operation" due to Sections 1.2.4 anda recirculation loop flow mismatch not within limits 3.6.3 2.4 ofAttachments 8 and 10address thatMELLLA+ is notanalyzed for singleloop operationIn addition, NMP2does not currentlypermit single loopoperation while in theMELLLA domain,because it is notanalyzed.19 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGETable 2 -Changes to NMP2 Technical SpecificationsNMP2 TS Description of the Change SupportingAttachmentTS 5.6.5 Replace the reference to "Reactor Protection System Complies with DSS-Instrumentation Setpoint for the OPRM -Upscale CD LTRFunction Allowable Value for Specification 3.3.1.1" Section 2.4 ofwith a reference to "The Manual Backup Stability Attachments 8 and 10Protection (BSP) Scram Region (Region I), the ManualBSP Controlled Entry Region (Region II), the modifiedAPRM Simulated Thermal Power -High setpointsused in the OPRM (Function 2.e), Automated BSPScram Region, and the BSP Boundary for Specification3.3.1.1."TS 5.6.8 Add a new TS section (i.e., TS 5.6.8) to define the Complies with DSS-contents of the report required by new Required Action CD LTRF-.2-.F.3 of TS 3.3.1.1 Section 2.4 ofAttachments 8 and 103.6 TSTF-493There are no effects on the current TS or their licensing bases relative to TSTF-493. Two TSReactor Protection System (RPS) functions are changing in this amendment: (1) the OPRM -Upscale function; and (2) the APRM -Flow Biased Simulated Thermal Power (STP) -Upscalefunction. The OPRM setpoints are unique to a particular core design for a particular fuel cycle.The OPRM function setpoints do not have specific TS allowable values (AVs). The APRM STP -High AVs are specified in TS Table 3.3.1.1-1.MELLLA+ changes the OPRM setpoints in that they are now derived from DSS-CD algorithmsversus Option III algorithms; however, their protective function remains the same. The revisedBases for TS 3.3.1.1 provided in Attachment 2 of this Enclosure states: "The OPRM Upscalefunction settings are not traditional instrumentation setpoints determined under an instrumentsetpoint methodology. There is no Allowable Value for this Function, and the OPRM UpscaleFunction is not [Limiting Safety System Setting (LSSS) Safety Limit (SL)]-related and [the DSS-CD Licensing Topical Report, NEDC-33075P-A] confirms that the OPRM Upscale Functionsettings based on DSS-CD also do not have traditional instrumentation setpoints determinedunder an instrument setpoint methodology."MELLLA+ also changes the APRM -Flow Biased Simulated Thermal Power -Upscale AV fortwo loop operations in the MELLLA+ domain and the APRM -Flow Biased Simulated ThermalPower -Upscale function is used for the Automated Backup Stability Protection (ABSP) if theOPRM becomes inoperable. The APRM STP-High AV and setpoint do have setpointmethodology applied as described in TSTF-493. In addition, the TSTF-493 footnotes werepreviously added to this function in Amendment 140 to the NMP2 Renewed Operating LicenseNPF-69 issued on December 22, 2011 (Reference 12).20 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE3.7 Topics Discussed During NRC Pre-MeetingsOn February 27, 2013, representatives from NMPNS met with the NRC to discuss theMELLLA+ LAR. During this meeting, the NRC sought clarification regarding several topics.Table 3 summarizes those topics, and provides a cross reference to the location in Attachments 8and 10 of this Enclosure that addresses the topic. The NRC issued a summary of this meeting onMarch 13, 2013 (Reference 13).Table 3 -Topics Discussed During NRC Pre-Meeting on February 27, 2013Topic as Summarized in NRC Meeting SummaryIssued on March 13, 2013 (Reference 13)MELLLA+ SARReference (Attachments 8and 10 of this Enclosure)Automated Backup Stability ProtectionThe NMP2 submittal is based on Revision 6 of NEDC-33075P-A. The NMP2 is planning to take exception to Rev 6 relative tothe Automatic Backup Stability Protection (ABSP) set points byusing a simplified method that is consistent with the ABSP setpoint methodology described in Revision 7 of NEDC-33075P.Since the NRC staff has not approved Revision 7 of the LicensingTopical Report (LTR) NEDE-33075P, Re: Detect and SuppressSolution-Confirmation Density (DSS-CD) for Automatic BackupStability Protection (ABSP), the License Amendment Request(LAR) should not refer to revision 7 of NEDE-33075P, butprovide the justifications, consistent with revision 7, for anyexceptions taken in the LAR.The NMPNS submittal isbased on Revision 7 ofNEDC-33075P-A.Since the February 27, 2013meeting, the NRC approvedRevision 7 of NEDC-33075P-AJustification provided inSection 2.4.3Emergency Core Cooling System NPSH Information provided inThe NMP2 does not take credit for Containment Accident Section 4.2.6Pressure (CAP) to assure adequate net positive suction head(NPSH). In response to NRC staff, the licensee stated that a re-analysis of CAP is not required as a result of MELLLA+. Basedon feedback from the NRC staff, the NMP2 MELLLA+ submittalwill reference the NMP2 Extended Power Uprate (EPU)submittal Requests for Additional Information (RAr's) related toCAP and describe that the NPSH margins in the NMP2 EPUresponses remain bounding for MELLLA+.DSS-CD Implementation Information provided inImplementation of DSS-CD Stability Solution in Place of Option Section 2.4.1III. The NMP2 MELLLA+ submittal will address theimplementation strategy for DSS-CD, including the need formonitoring the timing for arming the protection associated withDSS-CD and the Oscillation Power Range Monitor (OPRM) dataanalysis already completed.21 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGETable 3 -Topics Discussed During NRC Pre-Meeting on February 27,2013Topic as Summarized in NRC Meeting SummaryIssued on March 13, 2013 (Reference 13)MELLLA+ SARReference (Attachments 8and 10 of this Enclosure)TRACG ATWS with Core Instability (ATWSI)The NMP2 submittal will include anticipated transients withoutSCRAM with instability (ATWSI) sensitivity analysis resultsusing a modified T-min correlation similar to what GeneralElectric Hitachi Nuclear Energy (GEH) provided in response toanother licensee's RAI. Additional information on the model wasrequested if and when it becomes available. However, GEH notedthat there is no additional testing at this time.Information provided inSection 9.3.3Operator TrainingProvide the implementation plan outlining the simulator upgradeand operator training plan to support implementation of the LAR.Information provided inSection 10.6NMPNS has requested thatthe NRC approve this LARby October 2014. Tosupport this schedule,NMPNS plans to upgradethe simulator by the secondquarter of 2014 to supportoperator training in thesecond and third quarters of2014.Reference Core versus Actual Cycle Specific Core See Notes I through 3Cycle Specific Core Design and Associated Safety Analyses, andReload Analysis using PRIME Code. The NMP2 submittal will Information provided indescribe the potential differences in the analytical inputs and Sections 2.1, 2.2, and 2.6.3results between the reference core and the actual reload analysis and Footnote 4 of Appendixthat will be submitted as a supplement to the MELLLA+ Asubmittal.GESTR-M versus PRIME Following the NRCSubsequent to the meeting the NRC staff noted that the licensee's discussions, the MELLLA+presentation stated that the licensee's LAR submission is going to SAR was revised to utilizeinclude the analyses based in GESTR-M Code and it is planning PRIME Thermal-to supplement its LAR with the Analyses based on PRIME Code, Mechanical (T-M)The LAR submission based on GESTR-M Code would not be methodology. In addition,"acceptable," This staff concern has been communicated to the PRIME fuel parameterslicensee on March 12, 2013. have been used in theanalyses requiring fuelIn an email dated March 12, 2013, the NRC staff noted that a performance parameters.LAR submission based on GESTR-M Code would not be"acceptable". A follow-up meeting with the NRC was conducted Information provided inon March 29, 2013. Table 1-1, Sections 2.6.3and 4.322 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGETable 3 -Topics Discussed During NRC Pre-Meeting on February 27, 2013Topic as Summarized in NRC Meeting Summary MELLLA+ SARIssued on March 13, 2013 (Reference 13) Reference (Attachments 8and 10 of this Enclosure)Notes:1. The fuel and cycle-dependent analyses, including the plant-specific thermal limitsassessment, will be submitted for NRC staff confirmation by supplementing the initialMELLLA+ Safety Analysis Report (SAR) in accordance with Limitation and Condition12.4 of the MELLLA+ Licensing Topical Report (LTR) Safety Evaluation Report (SER).Specifically, CENG will provide the cycle specific Supplemental Reload Licensing Report(SRLR) and Fuel Bundle Information Report (FBIR), which includes the supplementalinformation to satisfy MELLLA+ LTR SER Limitation and Condition 12.4. CENG willsubmit this information by February 28, 2014.2. Nine Mile Point Nuclear Station, LLC (NMPNS) will provide a cycle-specific core designloading map along with a summary of differences between the reference design described inthe M+SAR and the cycle-specific core design. This summary will include differences inthe energy requirements, average enrichment, and analytical inputs, a cycle-specific thermallimits assessment, and the actual reload analysis results. Additionally, the SupplementalReload Licensing Report, which includes the cycle specific core map, will be provided.Submittal of the cycle-specific design will satisfy the NRC request made at the MELLLA+LAR pre-meeting on March 13, 2013.3. The NMP2 Cycle 15 specific reload analysis will utilize TRACG rather than ODYN forAOO. Section 9.1.1 of the MELLLA+ SAR (Attachments 8 and 10 of this Enclosure) states:"In the event that the cycle-specific reload analysis is based on TRACG rather ODYN forAOO, no 0.01 added to the OLMCPR is required."
3.1 MELLLA+Attachments 8 and 10 of this Enclosure provide non-proprietary and proprietary versions of the"Safety Analysis Report for Nine Mile Point Unit 2 Maximum Extended Load Line LimitAnalysis Plus (MELLLA+ SAR)," NEDO-33576NP and NEDC-33576P, respectively. TheMELLLA+ SAR summarizes the results of the significant safety evaluations performed thatjustify the expansion of the core flow operating domain for NMP2. The changes expand theoperating domain in the region of operation with less than rated core flow, but do not increase thelicensed power level or the maximum core flow. The expanded operating domain is identified asMELLLA+.The scope of evaluations required to support the expansion of the core flow operating domain tothe MELLLA+ boundary is contained in NEDC-33006P-A, "Maximum Extended Load LineLimit Analysis Plus," referred to as the MELLLA+ LTR (Reference 1). The MELLLA+ SARprovides a systematic disposition of the MELLLA+ LTR subjects applied to NMP2, includingperformance of plant-specific assessments and confirmation of the applicability of genericassessments to support a MELLLA+ core flow operating domain expansion. The MELLLA+operating domain expansion is applied as an incremental expansion of the operating boundarywithout changing the maximum licensed power, maximum core flow, or the current plant vesseldome pressure. The MELLLA+ SAR supports operation of NMP2 at a licensed thermal power of3,988 MWt with core flow as low as 85% of rated core flow.8 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEThe MELLLA+ core operating domain expansion does not require major plant systemsmodifications. NMP2 will implement the DSS-CD solution in accordance with the applicableLTRs (References 3 and 4), including the applicable limitations and conditions. Implementationof DSS-CD requires a revision to the existing stability solution software.The core operating domain expansion involves changes to the operating power/core flow map andchanges to a small number of instrument setpoints. Because there are no increases in theoperating pressure, power, steam flow rate, and feedwater flow rate, there are no significanteffects on the plant systems outside of the Nuclear Steam Supply System (NSSS). There is apotential increase in the steam moisture content at certain times while operating in theMELLLA+ operating domain. The effects of the potential increase in moisture content on plantsystems have been evaluated and determined to be acceptable. The MELLLA+ operating domainexpansion does not cause additional requirements to be imposed on any of the safety, balance-of-plant, electrical, or auxiliary systems. No changes to the power generation and electricaldistribution systems are required as a result of the MELLLA+ operating domain expansion.This report also addresses applicable limitations and conditions as described in the MELLLA+LTR SER for the GEH LTR NEDC-33173P, "Applicability of GE Methods to ExpandedOperating Domains" (Methods LTR SER) (Reference 4). A complete listing of the applicableLTR SER limitations and conditions and the sections of the MELLLA+ SAR which address themare presented in Appendices A, B, and C of the MELLLA+ SAR.Only previously NRC-approved or industry-accepted methods were used for the analyses ofaccidents and transients. Therefore, because the safety analysis methods have been previouslyaddressed, the details of the methods are not presented for review and approval in the MELLLA+SAR. Also, event and analysis descriptions that are already provided in other licensing reports orthe NMP2 Updated Safety Analysis Report (USAR) are not repeated within the MELLLA+ SAR.Evaluations of the reactor core and fuel performance, reactor coolant and connected systems,engineered safety features, instrumentation and control, electrical power and auxiliary systems,power conversion systems, radwaste systems and radiation sources, reactor safety performanceevaluations were performed. The MELLLA+ SAR summarizes the results of the evaluations thatjustify the MELLLA+ operating domain expansion to a minimum core flow rate of 85% of ratedcore flow at 100% RTP.Section 11.3.1 of Attachments 8 and 10 provides a summary of the modifications that will berequired to implement the MELLLA+ operating domain, DSS-CD, and the changes to the SLS.Section 11.3.2 of Attachments 8 and 10 provides a summary of the MELLLA+ issues including adiscussion of the MELLLA+ analysis basis, fuel thermal limits, makeup water sources, designbasis accidents, challenges to fuel, challenges to the containment, design basis accidentradiological consequences, anticipated operational occurrence analyses, combined effects,non-Loss of Coolant Accident (LOCA) radiological release accidents, equipment qualification,balance-of-plant, and environmental consequences.An assessment of the risk increase, including core damage frequency (CDF) and large earlyrelease frequency (LERF) associated with operation in the MELLLA+ operating domain isprovided in Attachment 4 of this Enclosure and Section 10.5 of Attachments 8 and 10 of thisEnclosure. The estimated risk increase for at-power events due to MELLLA+ is a delta CDF of1E-8 and delta LERF of 3E-9. This represents a very small risk change in RG 1.1749 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE(Reference 6). Based on these results, the proposed MELLLA+ operating domain is acceptableon a risk basis.3.2 DSS-CDThe long-term stability solution is being changed from the currently approved Option III solutionto DSS-CD. The DSS-CD solution algorithm, licensing basis, and application procedures aregenerically described in NEDC-33075P-A (Reference 2) and NEDE-33147P-A (Reference 3),and are applicable to NMP2 including any limitations and conditions associated with their useand approval. Section 2.4 of the MELLLA+ SAR (Attachments 8 and 10) addresses the changeto the DSS-CD stability solution. In addition, a complete listing of the required DSS-CD SERand limitations and conditions and the sections of the MELLLA+ SAR which address them ispresented in Appendix C of the MELLLA+ SAR, respectively.DSS-CD is designed to identify the power oscillation upon inception and initiate control rodinsertion (scram) to terminate the oscillations prior to any significant amplitude growth exceedingthe applicable safety limits. DSS-CD is based on the same hardware design as Option III.However, it introduces an enhanced detection algorithm that detects the inception of poweroscillations and generates an earlier power suppression trip signal exclusively based .n........ :.... :ri.d r-..gni. .The existing Option III algorithms are retained (withgeneric setpoints) to provide defense-in-depth protection for unanticipated reactor instabilityevents.3.3 Standby Liquid Control System Boron 10 En-.hmen-The SLS is described in Section 9.3.5 of the NMP2 USAR. The system provides a backupcapability for shutting down the reactor. The SLS is needed only in the event that sufficientcontrol rods cannot be inserted into the reactor core to accomplish shutdown and cooldown in thenormal manner. To accomplish this function, the SLS injects a sodium pentaborate solution intothe reactor. The SLS consists of a boron solution storage tank, two positive displacement pumps,two explosive valves (provided in parallel for redundancy), and associated piping and valves usedto transfer borated water from the storage tank to the reactor pressure vessel (RPV). The boratedwater solution is discharged into the RPV through the high pressure core spray sparger.The specified neutron absorber solution is sodium pentaborate. It is prepared by dissolvinggranularly-enriched sodium pentaborate in demineralized water (NMP2 USAR 9.3.5.2). Thesodium pentaborate solution is discharged radially over the top of the core through the HighPressure Core Spray (HPCS) sparger. The boron absorbs thermal neutrons and thereby terminatesthe nuclear fission chain reaction in the uranium fuel. The sodium pentaborate also acts as abuffer to maintain the suppression pool pH at or above 7.0 to prevent the re-evolution of iodine,when mixed in the suppression pool following a LOCA accompanied by significant fuel damage(NMP2 USAR Section 9.3.5.1).3.3.1 Reactor Boron Cold Shutdown Concentration RequirementsThe reactor boron concentration requirements for achieving cold shutdown (780 parts per million(ppm) natural boron) is not increased for MELLLA+, because there is no change in fuel type andno change to the operating cycle. The total weight of boron-10 required for cold shutdown(including the 25% margin) does change for MELLLA+, because of a conservative increase inthe assumed weight of the reactor coolant in the applicable analysis.10 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEThe shutdown margin is calculated for each plant reload and is documented in the SupplementalReload Licensing Report (SRLR).Ii ,"b at'-N 1 4 lift Z-- * ----AI nSLflpfl~C in imron ~u ~nricnmcnr in ~ou:um rcnuwarurc10 CFR 50.62(e)(4) rvauir-es:"Each boiling water -retor must have a standby liquid contrl system (SCS) withtheapability of injecting into the reacter- prcssurce vessel a bor-ated water- solutiont at such a florate, level of bor-on concentfation and borefn 10 isotope enrichmcnt, and aecountfing or-eactor pressure vessel voluime, that the resulting reaetivity entrl is at least equivalent tothat r-esulting from injection of 86 gallons per- minuite of 13 weight perceent sodiumpentaborate dceahydrate solutioin at the nAtur-al boroen 10 isotope abundance into a 251 inchinside diameter- r.ea.t. r pressure .ss. l for- a gien ... design..."The NRC approved licensing topical r-eport NEDE 31096 A (Rcfer-enee 7) provides a method*bwhich the boront equivalency r-equirement of 10 CFR 50.62(e)(4) can be demonefstrated.Equation 1 1 of that docuiment was used to demonstfate injectiont capacity equivalency as"elews.(Q/'86) x (M2511') x (C-'13) x (9/'49.8) 1 1Where-Q -expeeted SLS flew rate (gpmn)A425-L mass of water in the rcaetfr ivessel and recircuat~ion system at hot ratedconditions (Ibs) for- a 251 inceh diameter- vessel reference plant-l mass of water in the NMP2 reactor- vessel and r-eeir-eulation system at hot ratedeendifiefts Ibs)-sodium pentaborae solution concentreation (%494)E boron 10 isotope enrichment (atom percent)NMP2 is equipped with a 251 inch diameter reactor- vessel (NMP2 USAR Section 15.)assumptions utilized int the analyses of the changes to the SLS.Substituting the cuffent values definted in Table I in the above equationt yields:82.4/86 X 1 X 13.6413 X 25/19.8 -l.27&#xfd;--4Substituting the new values definted in Table 1 into the above equation yields:!80/86 X 1 X 13.6/13 X 92/19.8 -4.52> 1 for 13.6 v,0/_)This that the equivalent control capacity requirement of 10 CFR 50.62(e).4)is m et, when the changes to the SLS flow rate and the boront 10 isotope em-iehment are inceluded.int addition, the control margin ince s. This is due to -nefeasing the boroen 10 enrichment tennin the equation by a factor- of3.68 (i.e., 92'25 -3.68).-11 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEf41re: -T.I-. .....requre c ..I sm)I -1pump pto nave a o eow ra.. A at ii ' ...gpml.Maintaining the TS SR 3. 1.7.7 aeceptance cr-iter-ia for SLC pump flow rate at 41.2 gpmf providesmar~gin with r-espcct to the r~equired flow rate for- AT-WS mitigation. This issue has beeaddressed for- euffent eperation via the NMP-NS Correcetive Action System.As defined in Table 1, the analyzed SLS injection flow Fate is r-educed to 80 gpm flow r-ate otwo SLC System pumps in .perati.n to a ..ount for- diluti.n ..fe.ts identified by GEH4 SafetyCommunicationt 10 13, Standby Liquid C.ntrol System Dilution Flew, with additional margin.Table 1 Assumptions rega.rding SLS P.rforman.Pafametef Units Current Vaulue New-ValueReactor boron ..n.. ntatin for pp 7. ...0cold shutdown (natur-al boront)Maximuma allowable solutionf V.+% 144. -144conceentrationtMinimum allowable solutio 4-3-A 4-3-6conccntfatienSolution eoncentration assumed in Y% 4-3 4446ATWS anal-isi__Mfinimumn boroen 10 enrichment foi Aom 2- 9-2ATWS aftaly~fisDesigni-SLS pumip flew Faite 813M 4-54-Minimum SLS pump flow rate as gPm" 44l-2 4442defined in T-S 3.1.7Sing pump flow rate (two. pup i pin 92-4 80)3-.33.3.2 Change in SLS Pump Discharge Pressure Acceptance CriterionTS SR 3.1.7.7 is revised to increase the acceptance criterion for the SLS pump discharge pressurefrom >_ 1,327 psig to > 1,335 psig. This change is required due to the increase in the peak upperplenum pressure after SLS pump startup to 1,241 pounds per square inch -absolute (psia) asidentified in Tables 9-4 and 9-7 of Attachments 8 and 10 of this Enclosure. Currently, the peakupper plenum pressure after SLS pump startup is 1,236 psia. Thus, the ATWS analysis forMELLLA+ establishes a pressure differential of five psi. The SLS pump discharge pressureacceptance criterion in TS SR 3.1.7.7 is increased by eight psig to address the increase in theupper plenum pressure and provide an additional three psi margin.33.43.3.3 Anticipated Transient without SCRAMSection 9.3.1 of the MELLLA+ SAR (Attachments 8 and 10 of this Enclosure) provides asummary of the plant-specific analyses of Anticipated Transients Without Scram (ATWS) todemonstrate that the ATWS acceptance criteria are met for operation in the MELLLA+ operatingdomain. NMP2 meets the ATWS mitigation requirements in 10 CFR 50.62 for an alternate rodinsertion (ARI) system, SLS boron injection equivalent to 86 gpm, and automatic RPT logic. The12 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEplant-specific ATWS analyses take credit for the ATWS-RPT and SLS. However, ARI is notcredited.Section 9.3.1.1 of the MELLLA+ SAR (Attachments 8 and 10 of this Enclosure) provides alicensing basis ODYN ATWS analysis that demonstrates that the ATWS acceptance criteriawould be met in the event of a NMP2 response to an ATWS event initiated in the MELLLA+operating domain.In addition, a plant-specific ATWS analysis was performed at MELLLA+ conditions thatassumed operation of a single SLS pump. The analysis and the results are discussed in Section9.3.1.1.1 of the MELLLA+ SAR (Attachments 8 and 10 of this Enclosure). It concludes: "AllATWS acceptance criteria are met at MELLLA+ conditions with only a single SLS pumpoperating."3.3.5 Suppression Pool BufferingIn supp.rt of the Alte.ate Source Term (AST-) meth.dology, the SLS also provides suppr.ession.peel buffer-ing following a LOCA accompanied by significant fuel damage, preventing r-eevelution of iodine from the suppr-ession peol by maintaining the pool pH4 above 7.0. Seetion9.3.5.1 of the NMP2 USAR requires a sufficient coneentration and quaftity of sodiumpentaborate to be available for injcetion into the reaetor vessel to controel pH4 in the suppr-esiopool for 30 days fllwing a DBA LOCAi.The r1eduetiont int the mi r u :ired solution volume results in a reduction in. the e ssolutiont available for- injectionl to maneneppeso ol p14I  7.0 for- 30 days post LOCA.The minimum" sodium pentabor-ate solution vouei rqie for injection post LOCA foradequate p14 eontrol is 1,065 lons at the limiting t.e., a sodium pentaborateconcentration of 13.6%4). The minimum requir-ed tank vollumie at4 al concentration of 13.6 %ireduced frim 4,558.6 gallons to 1,600 gallons. While this does r educe the amount of cxeavailable solutiont, adequate margin is maintaincd to ensure that the SLS can perform its requiredAST support funetion.The proposed bor-on 10 enr-ichment changes do not impaet the capability to achieve and maintaina pH4 above 7.0 in the suppression pool following a LOCA, because the chemnical pr-opeffies nonenetration of the sodium pentaborate solution injected into the suppr-essin pool will remainthe same. Given the rduced volume of solution that wil be available, ther will be a two hourr-eduction int the maximum tome available to add boron to the supeso pool to ananpabove 7.0 (noeminal time based on low level alann is within 22- housvess h-- ~ttieowithin 24 hours). Are.vie. of the Emergency Operating Procedures confirmed that the sodiumpentaborate solutiont would be injeeted within 30 minutes following the eccunfeftee of LOCA.The maaximutm 22 hour- timne period proevides a large margin to the minimm requremet formanual operator- action toinec the sodium pentaborate solution of 30 minuatesq. In addition, theSuppression pool pisntepected to drop bielw:7 foseelda.Section 9.3.5.3 of the NMP2 USAR delineates that ontly one of the two SL.S loops was assumefor suppr-ession pool p1H control operation. The proeposed changes to the SLS do not affect thI., ...t (.... C' I1.-I...-- -It% A L ... .+I^ : l ..,i;. .~~~~0~~ ---- ---- --- ---I--I-- --------------------0~.... ' .. .... ....J .... ..... .rrI-of313 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE"J f J"QI- --3.3.C n~ige tin 818 Sterage I unit Soluiuon -VolumeThe pr-opescd bOroni 10 enriehment value allows the minimum solutiont Volumfe stor-ed in theSLsterage tank th be dedreased to 1,530 gallons at a sodium.. pentabrate e.ne.ntr..in of 14.4% and1,600 gallons at a sodium pentablrVate eoinentraion of 13.6%. The mark uip of NMP2 TS Figt3.1.7 1 promvided inf Attaclhment 1 mf this Entlesur delineates thce preposed hange in the net SLSstorage fankl solutioin volume.The required minimum lumes for- the 13.6 p%49 and 14.4 mar solutin vplumres were der tby deteamininag the m einimum salution veluifne and then increasing the allumne to acotun frei1) the dead vsluine not pumped in the reator0 that remains in the SLS and 14PCS piping;a2) instrument accuracy.The minimum net solution volume for- injection meets all consider-ations for AT-WS boroninjection raes, AST suppression pool pH- control, and assures tha the r-eactor- core boronconcentr-ation will be greater- than :780 ppm natural boron equivalent3-.343.3.3 SLS Pump Relief Valve Setpoint MarginThe SLS pump relief valve setpoint margin is the difference between the relief valve nominalsetpoint and the maximum SLS pump discharge pressure. A margin of 78 psi provides sufficientmargin against inadvertent relief valve lifting. The 78 psi is based on an allowance for the reliefvalve setpoint drift (typically 3% (3% of 1,600 psi = 48 psi)) and SLS pump pressure pulsations(30 psi).For MELLLA+ operation during the limiting ATWS event, the relief valve setpoint margin is205.7 psi. This margin is based on a SLS pump relief valve setpoint of 1552 psig (1600 psig -3% tolerance (i.e., 48 psig)) and subtracting a SLS pump discharge pressure of 1346.3 psig (i.e.,1552 psig -1346.3 psig = 205.7 psi). The margin reduces to 175.7 psi if 30 psi for SLS pumppressure pulsations is taken into consideration (i.e., 205.7 psi -30 psi = 175.7 psi).3.3.8 Net Positive Suction Head Available (NIPSH1) for SbS P-unqwrThe propesed changes include a reduction in the minimum volume for- the SLS storage tank. Tiresults inl at eduction in the swaic head available to provide Net Positive Suction Read (NPSH4Yfotthe SLS pumps. The calculation that determnines the SLS pumnp NPSNft did not take any cedifor- the staic head abo-ve the SLS storage tank zero level. The mninimum tank level cof:Fespondinfgto the minimum net volume pefmitted by the proposed change to Figure 3.1.7 1 is greater- thanthree feet above tankl zero.3.4 Safety Limit Minimum Critical Power RatioCycle specific transient analyses are performed to determine the required SLMCPR and thechange in Critical Power Ratio (CPR) [ACPR] for specific transients. To ensure that adequatemargin is maintained, a design requirement based on a statistical analysis was selected, in thatmoderate frequency transients caused by a single operator error or equipment malfunction shallbe limited such that, considering uncertainties in manufacturing and monitoring the coreoperating state, at least 99.9% of the fuel rods would be expected to avoid boiling transition. Thelowest allowable transient MCPR limit which meets the design requirement is termed the fuelcladding integrity SLMCPR.14 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGENUREG-0800, Standard Review Plan, Section 4.4, "Thermal and Hydraulic Design," AcceptanceCriterion No. I.B, states, in part, that the limiting (minimum) value of CPR is to be establishedsuch that at least 99.9% of the fuel rods in the core would not be expected to experience departurefrom nucleate boiling during normal operation or anticipated operational occurrences.A cycle specific Operating Limit MCPR (OLMCPR) is established to provide adequate assurancethat the fuel cladding integrity SLMCPR is not exceeded for any anticipated operationaltransients. The OLMCPR is obtained by adding the maximum value of ACPR for the mostlimiting transient postulated to occur at the plant to the fuel cladding integrity SLMCPR.3.4.1 Analytical Methods, Standards, Data and ResultsNMPNS proposes to revise SL 2.1.1.2 by increasing the SLMCPR for two recirculation loops inoperation from >_ 1.07 to > 1.09. The proposed change to the SLMCPR value for tworecirculation loops in operation is based on an analysis performed by GNF for NMP2 duringCycle 15 operations with MELLLA+ conditions. The GNF report, GNF-0000-0156-7490-RO-P,"GNF Additional Information Regarding the Requested Changes to the Technical SpecificationSLMCPR," dated August 26, 2013, supports changing the two recirculation loops in operationvalue of SLMCPR from >_ 1.07 to >_ 1.09, and maintaining the single recirculation loop inoperation value of SLMCPR at > 1.09. These values are based on NRC approved methods andprocedures. Attachments 9 and 11 of this Enclosure provide non-proprietary and proprietaryversions of the GNF report, respectively.GNF performed the SLMCPR calculation in accordance with Revision 19 of NEDE-2401 1-P-A,"General Electric Standard Application for Reactor Fuel," (Reference 8) using the followingNRC-approved methodologies and uncertainties:" NEDC-32601P-A, "Methodology and Uncertainties for Safety Limit MCPREvaluations," (August 1999) (Reference 9).* NEDC-32694P-A, "Power Distribution Uncertainties for Safety Limit MCPREvaluations" (August 1999) (Reference 10).* NEDC-32505P-A, "R-Factor Calculation Method for GEl 1, GE12 and GE13 Fuel,"(Revision 1, July 1999) (Reference 11).Section 2.9 of Attachments 9 and 11 of this Enclosure require NMPNS to "provide the currentand previous cycle power/flow map in a separate attachment." Figure 1-1 of Attachments 8 and10 of this Enclosure provide the power/flow operating map for MELLLA+. This will be thepower/flow map for NMP2 operations in Cycle 15 following NRC approval of this LicenseAmendment Request. Attachment 5 of this Enclosure provides the NMP2 power/flow operatingmap for the current operating cycle.3.4.2 Major Contributors to SLMCPR ChangeIn general, the calculated safety limit is dominated by two key parameters: (1) flatness of the corebundle-by-bundle MCPR distribution, and (2) flatness of the bundle pin-by-pin power/R-Factordistribution. Greater flatness in either parameter yields more rods susceptible to boiling transitionand thus a higher calculated SLMCPR. The MCPR Importance Parameter (MIP) measures the15 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEcore bundle-by-bundle MCPR distribution and the R-Factor Importance Parameter (RIP)measures the bundle pin-by-pin power/R-Factor distribution. The impact of the fuel loadingpattern on the calculated two recirculation loops in operation SLMCPR has been correlated to theparameter MIPRIP, which combines the MIP and RIP values.Another factor besides core MCPR distribution or bundle R-factor distribution that significantlyimpacts the SLMCPR is the expansion of the analysis domain that comes with the initialapplication of MELLLA+. The rated power / minimum core flow point is analyzed at a lowercore flow (than without MELLLA+) using increased uncertainties that tend to increase theSLMCPR. Also, a new point at off-rated power / off-rated flow was analyzed using the increaseduncertainties.Table 3 of the GNF analysis (Attachments 9 and 11 of this Enclosure) presents the MIP and RIPparameters for the previous cycle and the current cycle along with the two recirculation loops inoperation SLMCPR estimates using MIPRIP correlations. In addition, Table 3 of the GNFanalysis provided in Attachments 9 and 1 1 presents estimated impacts on the two recirculationloops in operation SLMCPR due to methodology deviations, penalties, and/or uncertaintydeviations from approved values. Based on the MIPRIP correlation and any impacts due todeviations from approved values, a final estimated two loops in operation SLMCPR isdetermined. Section 2.2 of the GNF analysis (Attachments 9 and 11 of this Enclosure) provides adetailed discussion of the items in Table 3 of the GNF analysis (Attachments 9 and 11 of thisEnclosure) that result in the increase in the estimated SLMCPR.3.4.3 Considerations Addressed in the GNF Analysis Regarding R-Factor, Core Flow Rate andRandom Effective Tip Reading, and Fuel Axial Power Shape PenaltySection 2.2.1 of the GNF analysis provides a discussion that justifies an increase in the R-Factoruncertainty value. GNF states that it generically increased the GEXL R-Factor uncertainty toaccount for an increase in channel bow due to the emerging unforeseen phenomena called controlblade shadow corrosion-induced channel bow, which is not accounted for in the channel bowuncertainty component of the approved R-Factor uncertainty. NMP2 has experienced controlblade shadow corrosion-induced channel bow. Accounting for the control blade shadowcorrosion-induced channel bow, the NMP2 Cycle 15 analysis shows an expected channel bowuncertainty which is bounded by the increased GEXL R-Factor uncertainty. Thus, the use of theincreased GEXL RFactor uncertainty value adequately accounts for the expected control bladeshadow corrosion-induced channel bow for NMP2 Cycle 15.Section 2.2.2 of the GNF analysis provides a discussion that identifies that the uncertainty valuesfor the core flow rate and the random effective tip reading in the two recirculation loops inoperation calculation were conservatively adjusted by using the single recirculation loop inoperation uncertainty values. The GNF analysis states the treatment of the core flow and randomeffective TIP reading uncertainties is based on the assumption that the signal to noise ratiodeteriorates as core flow is reduced.Section 2.4 of the GNF analysis provides a discussion regarding higher uncertainties and non-conservative bases in the GEXL correlations for the various types of axial power shapes. GNFdetermined that no power shape penalties were required to be applied to the calculated NMP2Cycle 15 SLMCPR values.16 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE3.4.4 ConclusionThe proposed change to revise SL 2.1.1.2 by increasing the SLMCPR for two recirculation loopsin operation from > 1.07 to > 1.09 is acceptable, and continues to maintain the same level ofsafety as the current licensing basis.3.5 NMP2 TS ChangesTable 2 defines the affected NMP2 TS, describes the change, and defines the supportingAttachment to this Enclosure that supports the TS Change.Table 2 -Changes to NMP2 Technical SpecificationsNMP2 TS Description of the Change SupportingAttachmentSL 2.1.1.2 Increase the SLMCPR for two recirculation loops in Attachments 9 and 11operation from > 1.07 to > 1.09TS 3.1.7 -Increase the SLS pump discharge pressure from Section 6.5.3 ofSR 3.1.7.7 > 1,327 psig to >_ 1,335 psig Attachments 8 and 10TS-83.b.7- hinr-easing the sodium pentaberate ber-on0 Seetiont 6.5.1 ofSR 3.A.7. enr-iehment rzguir-ement from ! 25 atom pcr-eent toAttachments 8 and 10and S92 .tem.peree.T-S Figure Reducig the minimum net vlume n to 1,600 gallt Section 6.5.1 of3.i.74i and 1,530 gallens at Scdium pentabonuats i ngentrmaiens Attachments 8 and 10of 13.6% Sad 14.4%, ersperticlhyT-S-Figure lneroasing the COdi ;an pcItabniate brcon 1 Seetio 6.5.1 in3. i7 I enriehmcn egunt tfrom ! 25 atom pereent toAttachiments 8 and 1092acteforPer."itTS 3.3.1.1 The Required Actions for Condition F are modified to: Complies with DSS-1) Initiate Action to implement the Manual BSP CD LTRRegions defined in the COLR; 2) Implement the Section 2.4 ofAutomated BSP Scram Region using the modified Attachments 8 and 10APRM Simulated Thermal Power -High scramsetpoints defined in the COLR; and 3) Initiate action inaccordance with Specification 5.6.8TS 3.3.1.1 Condition G is modified to no longer apply in the event Complies with DSS-a Required Action and associated Completion Time of CD LTRCondition F is not met. Section 2.4 ofAttachments 8 and 10TS 3.3. 1.1 New Condition J is added to address the action to take Complies with DSS-in the event a Required Action and associated CD LTRCompletion Time of Condition F is not met. Section 2.4 ofAttachments 8 and 1017 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGETable 2 -Changes to NMIP2 Technical SpecificationsNMP2 TS Description of the Change SupportingAttachmentTS 3.3.1.1 New Condition K is added to address the action to take Complies with DSS-in the event a Required Action and associated CD LTRCompletion Time of Condition J is not met. Section 2.4 ofAttachments 8 and 10TS SR Correct an editorial error in Note 3 (i.e., ORRM is Editorial correction3.3.1.1.13 changed to OPRM)TS SR Eliminate TS SR 3.3.1.1.16 and references to it in TS Complies with DSS-3.3.1.1.16 Table 3.3.1.1-1 CD LTRand TS Section 2.4 ofTable Attachments 8 and 103.3.1.1-1TS Table Change the AV for APRM -Flow Biased STP -Section 5.3.1 of3.3.1.1-1, Upscale from ":50.55W+60.5% RTP and < 115.5% Attachments 8 and 10Function 2.b RTP" to "< 0.61W + 63.4% RTP and < 115.5% RTP"TS Table Add a new note that requires the Flow Biased Complies with DSS-3.3.1.1-1, Simulated Thermal Power -Upscale scram setpoint to CD LTRFunction 2.b be reset to the values defined by the COLR to Section 2.4 ofimplement the Automated BSP Scram Region in Attachments 8 and 10accordance with Required Action F--.4F.2 of TS3.3.1.1TS Table Add a new note for Function 2.e, OPRM -Upscale, to Complies with DSS-3.3.1.1-1, denote that following implementation of DSS-CD, CD LTRFunction 2.e DSS-CD is not required to be armed while in the DSS- Section 2.4 ofCD Armed Region during the first reactor startup and Attachments 8 and 10during the first controlled shutdown that passescompletely through the DSS-CD Armed Region.However, DSS-CD is considered operable and capableof automatically arming for operation at recirculationdrive flow rates above the DSS-CD Armed RegionTS Table Change the mode of applicability for TS Table 3.3.1.1- Complies with DSS-3.3.1.1-1, 1, Function 2.e, OPRM-Upscale from Mode 1 to > 18% CD LTRFunction 2.e RTP. Section 2.4 ofAttachments 8 and 10TS Table Change the allowable value from "As specified in the Complies with DSS-3.3.1.1-1, COLR" to "NA" CD LTRFunction 2.e Section 2.4 ofAttachments 8 and 1018 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGETable 2 -Changes to NMP2 Technical SpecificationsNMP2 TS Description of the Change SupportingAttachmentTS.LCO Add a new requirement that pr...i..bModify the LCO Complies with DSS-3.4.1 to prohibit operation in the MELLLA domain or CD LTRMELLLA+ expanded operating domain as defined inthe COLR when in operation with a single recirculation Sections 1.2.4 andloop 3.6.3 2.4 ofAttachments 8 and 10address thatMELLLA+ is notanalyzed for singleloop operationIn addition, NMP2does not currentlypermit single loopoperation while in theMELLLA domain,because it is notanalyzed.TS 3.4.1, Add Required Action B.2 to identify that intentional Complies with DSS-Condition B operation in the MELLLA domain or MELLLA+ CD LTRdomain as defined in the COLR is prohibited when arecirculation loop is declared "not in operation" due to Sections 1.2.4 anda recirculation loop flow mismatch not within limits 3.6.3 2.4 ofAttachments 8 and 10address thatMELLLA+ is notanalyzed for singleloop operationIn addition, NMP2does not currentlypermit single loopoperation while in theMELLLA domain,because it is notanalyzed.19 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGETable 2 -Changes to NMP2 Technical SpecificationsNMP2 TS Description of the Change SupportingAttachmentTS 5.6.5 Replace the reference to "Reactor Protection System Complies with DSS-Instrumentation Setpoint for the OPRM -Upscale CD LTRFunction Allowable Value for Specification 3.3.1.1" Section 2.4 ofwith a reference to "The Manual Backup Stability Attachments 8 and 10Protection (BSP) Scram Region (Region I), the ManualBSP Controlled Entry Region (Region II), the modifiedAPRM Simulated Thermal Power -High setpointsused in the OPRM (Function 2.e), Automated BSPScram Region, and the BSP Boundary for Specification3.3.1.1."TS 5.6.8 Add a new TS section (i.e., TS 5.6.8) to define the Complies with DSS-contents of the report required by new Required Action CD LTRF-.2-.F.3 of TS 3.3.1.1 Section 2.4 ofAttachments 8 and 103.6 TSTF-493There are no effects on the current TS or their licensing bases relative to TSTF-493. Two TSReactor Protection System (RPS) functions are changing in this amendment: (1) the OPRM -Upscale function; and (2) the APRM -Flow Biased Simulated Thermal Power (STP) -Upscalefunction. The OPRM setpoints are unique to a particular core design for a particular fuel cycle.The OPRM function setpoints do not have specific TS allowable values (AVs). The APRM STP -High AVs are specified in TS Table 3.3.1.1-1.MELLLA+ changes the OPRM setpoints in that they are now derived from DSS-CD algorithmsversus Option III algorithms; however, their protective function remains the same. The revisedBases for TS 3.3.1.1 provided in Attachment 2 of this Enclosure states: "The OPRM Upscalefunction settings are not traditional instrumentation setpoints determined under an instrumentsetpoint methodology. There is no Allowable Value for this Function, and the OPRM UpscaleFunction is not [Limiting Safety System Setting (LSSS) Safety Limit (SL)]-related and [the DSS-CD Licensing Topical Report, NEDC-33075P-A] confirms that the OPRM Upscale Functionsettings based on DSS-CD also do not have traditional instrumentation setpoints determinedunder an instrument setpoint methodology."MELLLA+ also changes the APRM -Flow Biased Simulated Thermal Power -Upscale AV fortwo loop operations in the MELLLA+ domain and the APRM -Flow Biased Simulated ThermalPower -Upscale function is used for the Automated Backup Stability Protection (ABSP) if theOPRM becomes inoperable. The APRM STP-High AV and setpoint do have setpointmethodology applied as described in TSTF-493. In addition, the TSTF-493 footnotes werepreviously added to this function in Amendment 140 to the NMP2 Renewed Operating LicenseNPF-69 issued on December 22, 2011 (Reference 12).20 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE3.7 Topics Discussed During NRC Pre-MeetingsOn February 27, 2013, representatives from NMPNS met with the NRC to discuss theMELLLA+ LAR. During this meeting, the NRC sought clarification regarding several topics.Table 3 summarizes those topics, and provides a cross reference to the location in Attachments 8and 10 of this Enclosure that addresses the topic. The NRC issued a summary of this meeting onMarch 13, 2013 (Reference 13).Table 3 -Topics Discussed During NRC Pre-Meeting on February 27, 2013Topic as Summarized in NRC Meeting SummaryIssued on March 13, 2013 (Reference 13)MELLLA+ SARReference (Attachments 8and 10 of this Enclosure)Automated Backup Stability ProtectionThe NMP2 submittal is based on Revision 6 of NEDC-33075P-A. The NMP2 is planning to take exception to Rev 6 relative tothe Automatic Backup Stability Protection (ABSP) set points byusing a simplified method that is consistent with the ABSP setpoint methodology described in Revision 7 of NEDC-33075P.Since the NRC staff has not approved Revision 7 of the LicensingTopical Report (LTR) NEDE-33075P, Re: Detect and SuppressSolution-Confirmation Density (DSS-CD) for Automatic BackupStability Protection (ABSP), the License Amendment Request(LAR) should not refer to revision 7 of NEDE-33075P, butprovide the justifications, consistent with revision 7, for anyexceptions taken in the LAR.The NMPNS submittal isbased on Revision 7 ofNEDC-33075P-A.Since the February 27, 2013meeting, the NRC approvedRevision 7 of NEDC-33075P-AJustification provided inSection 2.4.3Emergency Core Cooling System NPSH Information provided inThe NMP2 does not take credit for Containment Accident Section 4.2.6Pressure (CAP) to assure adequate net positive suction head(NPSH). In response to NRC staff, the licensee stated that a re-analysis of CAP is not required as a result of MELLLA+. Basedon feedback from the NRC staff, the NMP2 MELLLA+ submittalwill reference the NMP2 Extended Power Uprate (EPU)submittal Requests for Additional Information (RAr's) related toCAP and describe that the NPSH margins in the NMP2 EPUresponses remain bounding for MELLLA+.DSS-CD Implementation Information provided inImplementation of DSS-CD Stability Solution in Place of Option Section 2.4.1III. The NMP2 MELLLA+ submittal will address theimplementation strategy for DSS-CD, including the need formonitoring the timing for arming the protection associated withDSS-CD and the Oscillation Power Range Monitor (OPRM) dataanalysis already completed.21 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGETable 3 -Topics Discussed During NRC Pre-Meeting on February 27,2013Topic as Summarized in NRC Meeting SummaryIssued on March 13, 2013 (Reference 13)MELLLA+ SARReference (Attachments 8and 10 of this Enclosure)TRACG ATWS with Core Instability (ATWSI)The NMP2 submittal will include anticipated transients withoutSCRAM with instability (ATWSI) sensitivity analysis resultsusing a modified T-min correlation similar to what GeneralElectric Hitachi Nuclear Energy (GEH) provided in response toanother licensee's RAI. Additional information on the model wasrequested if and when it becomes available. However, GEH notedthat there is no additional testing at this time.Information provided inSection 9.3.3Operator TrainingProvide the implementation plan outlining the simulator upgradeand operator training plan to support implementation of the LAR.Information provided inSection 10.6NMPNS has requested thatthe NRC approve this LARby October 2014. Tosupport this schedule,NMPNS plans to upgradethe simulator by the secondquarter of 2014 to supportoperator training in thesecond and third quarters of2014.Reference Core versus Actual Cycle Specific Core See Notes I through 3Cycle Specific Core Design and Associated Safety Analyses, andReload Analysis using PRIME Code. The NMP2 submittal will Information provided indescribe the potential differences in the analytical inputs and Sections 2.1, 2.2, and 2.6.3results between the reference core and the actual reload analysis and Footnote 4 of Appendixthat will be submitted as a supplement to the MELLLA+ Asubmittal.GESTR-M versus PRIME Following the NRCSubsequent to the meeting the NRC staff noted that the licensee's discussions, the MELLLA+presentation stated that the licensee's LAR submission is going to SAR was revised to utilizeinclude the analyses based in GESTR-M Code and it is planning PRIME Thermal-to supplement its LAR with the Analyses based on PRIME Code, Mechanical (T-M)The LAR submission based on GESTR-M Code would not be methodology. In addition,"acceptable," This staff concern has been communicated to the PRIME fuel parameterslicensee on March 12, 2013. have been used in theanalyses requiring fuelIn an email dated March 12, 2013, the NRC staff noted that a performance parameters.LAR submission based on GESTR-M Code would not be"acceptable". A follow-up meeting with the NRC was conducted Information provided inon March 29, 2013. Table 1-1, Sections 2.6.3and 4.322 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGETable 3 -Topics Discussed During NRC Pre-Meeting on February 27, 2013Topic as Summarized in NRC Meeting Summary MELLLA+ SARIssued on March 13, 2013 (Reference 13) Reference (Attachments 8and 10 of this Enclosure)Notes:1. The fuel and cycle-dependent analyses, including the plant-specific thermal limitsassessment, will be submitted for NRC staff confirmation by supplementing the initialMELLLA+ Safety Analysis Report (SAR) in accordance with Limitation and Condition12.4 of the MELLLA+ Licensing Topical Report (LTR) Safety Evaluation Report (SER).Specifically, CENG will provide the cycle specific Supplemental Reload Licensing Report(SRLR) and Fuel Bundle Information Report (FBIR), which includes the supplementalinformation to satisfy MELLLA+ LTR SER Limitation and Condition 12.4. CENG willsubmit this information by February 28, 2014.2. Nine Mile Point Nuclear Station, LLC (NMPNS) will provide a cycle-specific core designloading map along with a summary of differences between the reference design described inthe M+SAR and the cycle-specific core design. This summary will include differences inthe energy requirements, average enrichment, and analytical inputs, a cycle-specific thermallimits assessment, and the actual reload analysis results. Additionally, the SupplementalReload Licensing Report, which includes the cycle specific core map, will be provided.Submittal of the cycle-specific design will satisfy the NRC request made at the MELLLA+LAR pre-meeting on March 13, 2013.3. The NMP2 Cycle 15 specific reload analysis will utilize TRACG rather than ODYN forAOO. Section 9.1.1 of the MELLLA+ SAR (Attachments 8 and 10 of this Enclosure) states:"In the event that the cycle-specific reload analysis is based on TRACG rather ODYN forAOO, no 0.01 added to the OLMCPR is required."


==4.0 REGULATORY EVALUATION==
==4.0 REGULATORY EVALUATION==
Line 46: Line 48:


==5.0 ENVIRONMENTAL CONSIDERATION==
==5.0 ENVIRONMENTAL CONSIDERATION==
A review has determined that the proposed amendment would change a requirement with respectto installation or use of a facility component located within the restricted area, as defined in10 CFR 20, or would change an inspection or surveillance requirement. However, the proposedamendment does not involve: (i) a significant hazards consideration; (ii) a significant change inthe types or significant increase in the amounts of any effluent that may be released offsite; or(iii) a significant increase in individual or cumulative occupational radiation exposure.Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion setforth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impactstatement or environmental assessment need be prepared in connection with the proposedamendment.6.0 REFERENCES1. GE Hitachi Nuclear Energy, "General Electric Boiling Water Reactor Maximum ExtendedLoad Line Limit Analysis Plus Licensing Topical Report," NEDC-33006P-A, Revision 3,June 2009 and NEDO-33006-A, Revision 3, June 2009.2. GE Hitachi Nuclear Energy, "GE Hitachi Boiling Water Reactor, Detect And SuppressSolution -Confirmation Density," NEDC-33075P, Revision 7, June 2011; and Anthony J.Mendiola (NRC) to Jerald G. Head (GEH), "Revised Draft Safety Evaluation for GE-HitachiNuclear Energy Americas, LLC Topical Report NEDC-33075P, Revision 7, 'GE HitachiBoiling Water Reactor Detect and Suppress Solution -Confirmation Density' (TAC No.ME6577)," dated August 6, 2013.3. GE Hitachi Nuclear Energy, "DSS-CD TRACG Application," NEDE-33147P-A, Revision 4,August 2013.35 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE4. a. GE Hitachi Nuclear Energy, "Applicability of GE Methods to Expanded OperatingDomains," NEDC-33173P-A, Revision 4, November 2012.b. Letter from R. Kingston (GEH) to NRC, "Clarification of Stability Evaluations -NEDC-33173P," MFN 08-541, June 25, 2008.c. Letter from J. Harrison (GEH) to NRC, "Implementation of Methods Limitations -NEDC-33173," MFN 08-693, September 18, 2008.d. Letter from J. Harrison (GEH) to NRC, "NEDC-33173P -Implementation of Limitation12," MFN 09-143, February 27, 2009.e. GE Hitachi Nuclear Energy, "Implementation of PRIME Models and Data inDownstream Methods," NEDO-33173, Supplement 4-A, Revision 1, November 2012.5. Letter from K. Langdon (NMPNS) to the Document Control Desk (NRC), LicenseAmendment Request Pursuant to 10 CFR 50.90: Relocation of Pressure and TemperatureLimit Curves to the Pressure and Temperature Limits Report, dated November 21, 2012(ADAMS Accession Number ML123380336).6. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," U.S. NRC,Revision 2, May 2011.7. NEDE 31096 A, "Anticipated Transients Without Senm Respense to NRC ATWS Rl..CFR5..62," Fcbrua, y ..87..Not used.8. GE Hitachi Nuclear Energy, "General Electric Standard Application for Reactor Fuel,"NEDE-2401 1-P-A, and NEDE-2401 1-P-A-US, Revision 19 April 2012.9 NEDC-32601P-A, "Methodology and Uncertainties for Safety Limit MCPR Evaluations,"August 1999.10 NEDC-32694P-A, "Power Distribution Uncertainties for Safety Limit MCPR Evaluations"August 1999.11 NEDC-32505P-A, "R-Factor Calculation Method for GEl 1, GE12 and GE13 Fuel,"Revision 1, July 1999.12. Letter from (NRC) to K. Langdon (NMPNS), Nine Mile Point Nuclear Station, Unit No. 2 -Issuance of Amendment Re: Extended Power Uprate (TAC No. ME1476), dated December22, 2011 (ADAMS Accession Number ML1 133000041).13. B. Vaidya (NRC), Summary of February 27,_2013, Meeting with Nine Mile Point NuclearStation, Unit 2, to discuss Planned Amendment Request on Implementation of MELLLA+(TAC No. MF0587), dated March 13, 2013 (ADAMS Accession Number ML13059A374)14. NRR Office Instruction LIC-109, "Acceptance Review Procedures," Revision 1.36 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE1 1; I at..... fxs 12A U..C [Muit !NWJC tgA DA~n4m~nt IRequest Pursuant te 10 CFR 50.90! Standby Liquid Control System Inerease in lsctepieEnrichment of Beroen 10, dated July 1C, 2013 (ADAMS Accessiont NumbefMb 1 31l9-7A224-l-)Not used.16. Information Notice 2009-23, "Nuclear Fuel Thermal Conductivity Degradation," U.S.Nuclear Regulatory Commission, dated October 8, 2009.17. Information Notice 2009-23, Supplement 1, "Nuclear Fuel Thermal ConductivityDegradation," U.S. Nuclear Regulatory Commission, dated October 26, 2012.18. Information Notice 2011-21, "Realistic Emergency Core Cooling System Evaluation ModelEffects Resulting from Nuclear Fuel Thermal Conductivity Degradation," U.S. NuclearRegulatory Commission, dated December 13, 2011.19. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating DesignBasis Accidents at Nuclear Power Reactors," July 2000 (ADAMS Accession No.ML003716792).20. Standard Review Plan (SRP) 15.0.1, "Radiological Consequence Analyses Using AlternativeSource Terms," Revision 0, July 2000 (ADAMS Accession No. ML003721661).21. Information Notice 2001-13, "Inadequate Standby Liquid Control System Relief ValveMargin," U.S. Nuclear Regulatory Commission, August 10, 2001.22. Letter from T. Beltz (NRC) to K. Fili (Monticello Nuclear Generating Plant), MoniticelloNuclear Generating Plant -Issuance of Amendment No. 180 to Renewed Facility OperatingLicense repardine Maximum Extended Load Line Limit Anavsis Plus (TAC No. ME3145),23.24.dated March 28, 2014 (ADAMS Accession No. ML14035A248).lcttcr from T. J. O'Connor(Monticcfll Nuclear- Gencrating Plant) to U. S. Nuelear Regulatory Commission, -1iensAmendment Request.: M.. imum Etended Load Line Limit Analysis Pluts," dated Ja"ualry,21, 2010 (ADAMS Aeecssiefn No. ML100280558).Lcter- from M. C. Thadani (NRC) te M. E. Reddemann (Energy Ncrthwest), "CcltimbiaGencr-ating Station issuanee of Amendment Re. iner-easd Ber-en ConeentrMien in StandbyLiquid Control System (TAC No. ME14789)," dated May 18, 2011 (ADAMS Aeeession No.Mbin 1 tl-703 70)7Not used.Letter- from R. V. Guizmani (NRC) to 13. T. McKintney (PPL Susquehanna, LLC),"Susquehanna Steam Ekcetric Staion, units I and 2 Issuance Cf Amnendment -c StandbLiquid Control System (TAG Nos. MD1424 and MD!425)," dated Febrdar-y 28, 200-7I A+.. +r.,A SC A .T L "-, k ... uT DIV % IV, D "T. ' .. kJ T'-- +. D l .+ ..I.- .I 'Aeees;Ste" e. r"bt0-t196Y1d:&.t:1"r.1NVt USU,25. Letter from N. DiFrancesco (NRC) to M. J. Pacillo (Exelon Nuclear), LaSalle CountyStation, Unit 2 -Issuance of Amendment No. 192 Regarding Technical Specification Changefor Safety Limit Minimum Critical Power Ratio (TAC No. ME9769), dated February27, 2013 (ADAMS Accession No. ML13050A637).26. Letter from L. E. Wilkins (NRC) to B. J. O'Grady (Nebraska Public Power District), CooperNuclear Station -Issuance of Amendment Re: Revision of Technical Specifications -Safety37 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGELimit Minimum Critical Power Ratio (TAC No. ME8853), dated November 9, 2012(ADAMS Accession No. ML12299A092).38 of 38 ATTACHMENT 1NINE MILE POINT UNIT 2PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS(MARK-UPS)The current version of the following Technical Specification (TS) page had been marked-up to reflect theproposed changes and replaces the corresponding page previously submitted on November 1, 2013(ADAMS Accession No. ML 13316B 107):3.1.7-3TS page 3.1.7-4 and associated TS Insert 1 -Figure 3.1.7-1 were implemented with approval of NMP2Amendment 143, and are to be removed from the original Attachment 1 submitted on November 1, 2013(ADAMS Accession No. ML 13316B 107).The remaining pages in Attachment 1, submitted on NovemberML13316B107) and modified by the submittal on May 14,ML14139A416), are not changed with this submittal.1, 2013 (ADAMS Accession No.2014 (ADAMS Accession No.Nine Mile Point Nuclear Station, LLCJune 13, 2014 SLC System3.1.7SURVEILLANCE REQUIREMENTS (continued)SURVEILLANCE FREQUENCYSR 3.1.7.7 Verify each pump develops a flow rate In accordance> 41.2 gpm at a discharge pressure with the> 4-3H psig. InserviceTestingProgramSR 3.1.7.8 Verify flow through one SLC subsystem 24 months on afrom pump into reactor pressure vessel. STAGGERED TESTBASISSR 3.1.7.9 Verify all heat traced piping between 24 monthsstorage tank and pump suction valve isunblocked. ANDOnce within24 hours afterpipingtemperature isrestored to> 70&deg;FSR 3.1.7.10 Verify sodium pentaborate enrichment Prior tois > 92 atom percent B-10. addition toSLC tankINMP23.1.7-3Amendment W, 111, 117, 123, 140,44-37  
A review has determined that the proposed amendment would change a requirement with respectto installation or use of a facility component located within the restricted area, as defined in10 CFR 20, or would change an inspection or surveillance requirement. However, the proposedamendment does not involve: (i) a significant hazards consideration; (ii) a significant change inthe types or significant increase in the amounts of any effluent that may be released offsite; or(iii) a significant increase in individual or cumulative occupational radiation exposure.Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion setforth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impactstatement or environmental assessment need be prepared in connection with the proposedamendment.
}}
 
==6.0 REFERENCES==
: 1. GE Hitachi Nuclear Energy, "General Electric Boiling Water Reactor Maximum ExtendedLoad Line Limit Analysis Plus Licensing Topical Report," NEDC-33006P-A, Revision 3,June 2009 and NEDO-33006-A, Revision 3, June 2009.2. GE Hitachi Nuclear Energy, "GE Hitachi Boiling Water Reactor, Detect And SuppressSolution -Confirmation Density," NEDC-33075P, Revision 7, June 2011; and Anthony J.Mendiola (NRC) to Jerald G. Head (GEH), "Revised Draft Safety Evaluation for GE-HitachiNuclear Energy Americas, LLC Topical Report NEDC-33075P, Revision 7, 'GE HitachiBoiling Water Reactor Detect and Suppress Solution -Confirmation Density' (TAC No.ME6577)," dated August 6, 2013.3. GE Hitachi Nuclear Energy, "DSS-CD TRACG Application," NEDE-33147P-A, Revision 4,August 2013.35 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE4. a. GE Hitachi Nuclear Energy, "Applicability of GE Methods to Expanded OperatingDomains," NEDC-33173P-A, Revision 4, November 2012.b. Letter from R. Kingston (GEH) to NRC, "Clarification of Stability Evaluations -NEDC-33173P," MFN 08-541, June 25, 2008.c. Letter from J. Harrison (GEH) to NRC, "Implementation of Methods Limitations -NEDC-33173," MFN 08-693, September 18, 2008.d. Letter from J. Harrison (GEH) to NRC, "NEDC-33173P -Implementation of Limitation12," MFN 09-143, February 27, 2009.e. GE Hitachi Nuclear Energy, "Implementation of PRIME Models and Data inDownstream Methods," NEDO-33173, Supplement 4-A, Revision 1, November 2012.5. Letter from K. Langdon (NMPNS) to the Document Control Desk (NRC), LicenseAmendment Request Pursuant to 10 CFR 50.90: Relocation of Pressure and TemperatureLimit Curves to the Pressure and Temperature Limits Report, dated November 21, 2012(ADAMS Accession Number ML123380336).6. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," U.S. NRC,Revision 2, May 2011.7. NEDE 31096 A, "Anticipated Transients Without Senm Respense to NRC ATWS Rl..CFR5..62," Fcbrua, y ..87..Not used.8. GE Hitachi Nuclear Energy, "General Electric Standard Application for Reactor Fuel,"NEDE-2401 1-P-A, and NEDE-2401 1-P-A-US, Revision 19 April 2012.9 NEDC-32601P-A, "Methodology and Uncertainties for Safety Limit MCPR Evaluations,"August 1999.10 NEDC-32694P-A, "Power Distribution Uncertainties for Safety Limit MCPR Evaluations"August 1999.11 NEDC-32505P-A, "R-Factor Calculation Method for GEl 1, GE12 and GE13 Fuel,"Revision 1, July 1999.12. Letter from (NRC) to K. Langdon (NMPNS), Nine Mile Point Nuclear Station, Unit No. 2 -Issuance of Amendment Re: Extended Power Uprate (TAC No. ME1476), dated December22, 2011 (ADAMS Accession Number ML1 133000041).13. B. Vaidya (NRC), Summary of February 27,_2013, Meeting with Nine Mile Point NuclearStation, Unit 2, to discuss Planned Amendment Request on Implementation of MELLLA+(TAC No. MF0587), dated March 13, 2013 (ADAMS Accession Number ML13059A374)14. NRR Office Instruction LIC-109, "Acceptance Review Procedures," Revision 1.36 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE1 1; I at..... fxs 12A U..C [Muit !NWJC tgA DA~n4m~nt IRequest Pursuant te 10 CFR 50.90! Standby Liquid Control System Inerease in lsctepieEnrichment of Beroen 10, dated July 1C, 2013 (ADAMS Accessiont NumbefMb 1 31l9-7A224-l-)Not used.16. Information Notice 2009-23, "Nuclear Fuel Thermal Conductivity Degradation," U.S.Nuclear Regulatory Commission, dated October 8, 2009.17. Information Notice 2009-23, Supplement 1, "Nuclear Fuel Thermal ConductivityDegradation," U.S. Nuclear Regulatory Commission, dated October 26, 2012.18. Information Notice 2011-21, "Realistic Emergency Core Cooling System Evaluation ModelEffects Resulting from Nuclear Fuel Thermal Conductivity Degradation," U.S. NuclearRegulatory Commission, dated December 13, 2011.19. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating DesignBasis Accidents at Nuclear Power Reactors," July 2000 (ADAMS Accession No.ML003716792).20. Standard Review Plan (SRP) 15.0.1, "Radiological Consequence Analyses Using AlternativeSource Terms," Revision 0, July 2000 (ADAMS Accession No. ML003721661).21. Information Notice 2001-13, "Inadequate Standby Liquid Control System Relief ValveMargin," U.S. Nuclear Regulatory Commission, August 10, 2001.22. Letter from T. Beltz (NRC) to K. Fili (Monticello Nuclear Generating Plant), MoniticelloNuclear Generating Plant -Issuance of Amendment No. 180 to Renewed Facility OperatingLicense repardine Maximum Extended Load Line Limit Anavsis Plus (TAC No. ME3145),23.24.dated March 28, 2014 (ADAMS Accession No. ML14035A248).lcttcr from T. J. O'Connor(Monticcfll Nuclear- Gencrating Plant) to U. S. Nuelear Regulatory Commission, -1iensAmendment Request.: M.. imum Etended Load Line Limit Analysis Pluts," dated Ja"ualry,21, 2010 (ADAMS Aeecssiefn No. ML100280558).Lcter- from M. C. Thadani (NRC) te M. E. Reddemann (Energy Ncrthwest), "CcltimbiaGencr-ating Station issuanee of Amendment Re. iner-easd Ber-en ConeentrMien in StandbyLiquid Control System (TAC No. ME14789)," dated May 18, 2011 (ADAMS Aeeession No.Mbin 1 tl-703 70)7Not used.Letter- from R. V. Guizmani (NRC) to 13. T. McKintney (PPL Susquehanna, LLC),"Susquehanna Steam Ekcetric Staion, units I and 2 Issuance Cf Amnendment -c StandbLiquid Control System (TAG Nos. MD1424 and MD!425)," dated Febrdar-y 28, 200-7I A+.. +r.,A SC A .T L "-, k ... uT DIV % IV, D "T. ' .. kJ T'-- +. D l .+ ..I.- .I 'Aeees;Ste" e. r"bt0-t196Y1d:&.t:1"r.1NVt USU,25. Letter from N. DiFrancesco (NRC) to M. J. Pacillo (Exelon Nuclear), LaSalle CountyStation, Unit 2 -Issuance of Amendment No. 192 Regarding Technical Specification Changefor Safety Limit Minimum Critical Power Ratio (TAC No. ME9769), dated February27, 2013 (ADAMS Accession No. ML13050A637).26. Letter from L. E. Wilkins (NRC) to B. J. O'Grady (Nebraska Public Power District), CooperNuclear Station -Issuance of Amendment Re: Revision of Technical Specifications -Safety37 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGELimit Minimum Critical Power Ratio (TAC No. ME8853), dated November 9, 2012(ADAMS Accession No. ML12299A092).38 of 38 ATTACHMENT 1NINE MILE POINT UNIT 2PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS(MARK-UPS)The current version of the following Technical Specification (TS) page had been marked-up to reflect theproposed changes and replaces the corresponding page previously submitted on November 1, 2013(ADAMS Accession No. ML 13316B 107):3.1.7-3TS page 3.1.7-4 and associated TS Insert 1 -Figure 3.1.7-1 were implemented with approval of NMP2Amendment 143, and are to be removed from the original Attachment 1 submitted on November 1, 2013(ADAMS Accession No. ML 13316B 107).The remaining pages in Attachment 1, submitted on NovemberML13316B107) and modified by the submittal on May 14,ML14139A416), are not changed with this submittal.1, 2013 (ADAMS Accession No.2014 (ADAMS Accession No.Nine Mile Point Nuclear Station, LLCJune 13, 2014 SLC System3.1.7SURVEILLANCE REQUIREMENTS (continued)SURVEILLANCE FREQUENCYSR 3.1.7.7 Verify each pump develops a flow rate In accordance> 41.2 gpm at a discharge pressure with the> 4-3H psig. InserviceTestingProgramSR 3.1.7.8 Verify flow through one SLC subsystem 24 months on afrom pump into reactor pressure vessel. STAGGERED TESTBASISSR 3.1.7.9 Verify all heat traced piping between 24 monthsstorage tank and pump suction valve isunblocked. ANDOnce within24 hours afterpipingtemperature isrestored to> 70&deg;FSR 3.1.7.10 Verify sodium pentaborate enrichment Prior tois > 92 atom percent B-10. addition toSLC tankINMP23.1.7-3Amendment W, 111, 117, 123, 140,44-37}}

Revision as of 02:52, 28 June 2018

Nine Mile Point, Unit 2 - License Amendment Request Pursuant to 10 CFR 50.90: Maximum Extended Load Line Limit Analysis Plus - Revision 1
ML14169A034
Person / Time
Site: Nine Mile Point Constellation icon.png
Issue date: 06/13/2014
From: Costanzo C R
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC ME3145, TAC MF2462
Download: ML14169A034 (46)
v • d • e


Text

AChris CostanzoAM P--i Site Vice President -Nine Mile PointExeonG P.O. Box 63Lycoming, NY 13093315-349-5200 Officewww.exeloncorp.comChristopher.costanzo@exeloncorp.comJune 13, 2014U. S. Nuclear Regulatory CommissionATTN: Document Control DeskWashington, DC 20555-0001Nine Mile Point Nuclear Station, Unit 2Renewed Facility Operating License No. NPF-69Docket No. 50-410

Subject:

License Amendment Request Pursuant to 10 CFR 50.90: MaximumExtended Load Line Limit Analysis Plus -Revision 1

References:

(1) Letter from P. Swift (NMPNS) to Document Control Desk (USNRC),License Amendment Request Pursuant to 10 CFR 50.90: MaximumExtended Load Line Limit Analysis Plus, dated November 1, 2013(ADAMS Accession No. ML1 3316B1307)(2) Letter from B. Vaidya (USNRC) to C. Costanzo (NMPNS), Nine MilePoint Nuclear Station, Unit No. 2 -Issuance of Amendment, Re:License Amendment Request Pursuant to 10 CFR 50.90: StandbyLiquid Control System -Increase in Isotopic Enrichment of Boron-10(TAC No. MF2462), dated March 14, 2014 (ADAMS Accession No.M L 1 4036A005)(3) Letter from J. Stanley (NMPNS) to Document Control Desk(USNRC), License Amendment Request Pursuant to 10 CFR 50.90:Maximum Extended Load Line Limit Analysis Plus -Response toRAI STSB-1 and RAI STSB-2, dated May 14, 2014 (ADAMSAccession No. ML14139A146)(4) Letter from T. Beltz (NRC) to K. Fili (Monticello Nuclear GeneratingPlant), Moniticello Nuclear Generating Plant -Issuance ofAmendment No. 180 to Renewed Facility Operating Licenseregarding Maximum Extended Load Line Limit Anaysis Plus (TACNo. ME3145), dated March 28, 2014 (ADAMS Accession No.ML1 4035A248)Nine Mile Point Nuclear Station, LLC (NMPNS) hereby transmits Revision 1 to the NineMile Point Unit 2 (NMP2) License Amendment Request originally submitted onNovember 1, 2013 (Reference 1). The request to amend the NMP2 Renewed FacilityOperating License No. NPF-69 included a proposed expansion of the operating boundaryto allow operation in the Maximum Extended Load Line Limit Analysis Plus (MELLLA U. S. Nuclear Regulatory CommissionJune 13, 2014Page 2Plus) domain and the use of the General Electric Hitachi Nuclear Energy (GEH) analysiscode TRACG04.Revision 1 of the Enclosure is modified using track changes with all changes in color.Revision 1 of the Enclosure modifies the original request to reflect:1) The USNRC's issuance of Amendment No. 143 to NMP2 Renewed FacilityOperating License No. NPF-69 regarding an increase in the isotopic enrichmentof Boron-10 in the Standby Liquid Control System (Reference 2);2) Changes to the original LAR made in a response to an USNRC Staff Requestfor Additional Information (Reference 3);3) A correction to the No Significant Hazards Consideration regarding thediscussion of the change to the long term stability solution; and4) The USNRC's issuance of Amendment No. 180 to Monticello NuclearGenerating Plant's Renewed Facility Operating License No. DPR-22.Revisions to the Technical Specification (TS) and TS Bases were previously submitted inReference (3). The changes in the TS and TS Bases have been incorporated intorevision 1 of the Enclosure to provide a marked up copy of the Enclosure encompassingall changes to date. Attachment 1 to the Enclosure includes a revision to TS page 3.1.7-3 which replaces the corresponding page previously submitted in Reference (1). TSpage 3.1.7-4 and associated TS Insert 1 -Figure 3.1.7-1 were implemented with approvalof NMP2 Amendment 143, and are to be removed from the original Attachment 1submitted in Reference (1). The remaining pages in Attachment 1 to the Enclosure,submitted on November 1, 2013 (Reference 1) and modified by the submittal on May 14,2014 (Reference 3), are not changed with this submittal. Attachments 2 through 11 of theEnclosure to the original request submitted on November 1, 2013 (Reference 1) are notrevised or reissued.This submittal revises the No Significant Hazards Determination analysis provided byNMPNS in Reference (1). Pursuant to 10 CFR 50.91 (b)(1), NMPNS has provided a copyof this supplemental information to the appropriate state representative.This letter contains no new regulatory commitments.Should you have any questions regarding the information in this submittal, please contactEverett (Chip) Perkins, Director Licensing, at (315) 349-5219.I declare under penalty of perjury that the foregoing is true and correct. Executed on the13th day of June, 2014.Sincerely,Christopher R. Costanzo U. S. Nuclear Regulatory CommissionJune 13, 2014Page 3CRC/KJK

Enclosure:

Revision 1 -Evaluation of the Proposed Changescc: Regional Administrator, Region I, USNRCProject Manager, USNRCResident Inspector, USNRCA. L. Peterson, NYSERDA ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGESNine Mile Point Nuclear Station, LLCJune 13, 2014 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGETABLE OF CONTENTS1.0 SUMMARY DESCRIPTION2.0 DETAILED DESCRIPTION2.1 Background2.2 Proposed Changes to the Nine Mile Point Unit 2 Technical Specifications2.3 Modification Summary

3.0 TECHNICAL EVALUATION

3.1 MELLLA+3.2 DSS-CD3.3 Standby Liquid Control System Bororn 10 Enrichment3.4 Safety Limit Minimum Critical Power Ratio3.5 NMP2 TS Changes3.6 TSTF-4933.7 Topics Discussed During NRC Pre-Meetings

4.0 REGULATORY EVALUATION

4.1 Evaluation of NMP2 License Amendment Requests to Establish that They Are NotLinked4.2 Applicable Regulatory Requirements/Criteria4.3 Precedent4.4 Significant Hazards Consideration4.5 Conclusions

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

i of ii ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGETABLE OF CONTENTSATTACHMENTS1. Nine Mile Point Unit 2 Proposed Changes to Technical Specifications (Mark-ups)Note: Attachments 2 through II of the Enclosure to the Nine Mile Point Unit 2 License AmendmentRequest dated November 1, 2013 are not revised or reissued in this revision.2. Nine Mile Point Unit 2 Changes to Bases for Technical Specifications (Mark-ups)3. List of Regulatory Commitments4. MELLLA+ Risk Evaluation5. Nine Mile Point Unit 2 Power/Flow Operating Map for Current Cycle6. General Electric -Hitachi Affidavit Justifying Withholding Proprietary Information inNEDC-33576P7. Global Nuclear Fuel Affidavit Justifying Withholding Proprietary Information in GNF-0000-0156-7490-RO-P8. NEDC-33576NP, Safety Analysis Report for Nine Mile Point Unit 2 Maximum Extended Load LineLimit Analysis Plus (Non-proprietary)9. Global Nuclear Fuel Report GNF-0000-0 1 56-7490-RO-NP, "GNF Additional Information Regardingthe Requested Change to the Technical Specification SLMCPR," dated August 26, 2013 (Non-proprietary)10. NEDC-33576P, Safety Analysis Report for Nine Mile Point Unit 2 Maximum Extended Load LineLimit Analysis Plus (Proprietary)11. Global Nuclear Fuel Report GNF-0000-0 1 56-7490-RO-P, "GNF Additional Information Regardingthe Requested Change to the Technical Specification SLMCPR," dated August 26, 2013 (Proprietary)ii of ii ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE1.0 SUMMARY DESCRIPTIONThis evaluation supports a request to amend Renewed Operating License (OL) NPF-69 for NineMile Point Unit 2 (NMP2). The proposed amendment includes supporting changes to the NMP2Technical Specifications (TSs) necessary to: 1) implement the Maximum Extended Load LineLimit Analysis Plus (MELLLA+) expanded operating domain; 2) change the stability solution toDetect and Suppress Solution -Confirmation Density (DSS-CD); 3) use the TRACG04 analysiscode; and 4) incr.ease the isotpic en.rihment of boro.n 10 in the sodium pentabor.. e so!uti.nutilized in the Standby Liquid Control System (SLS); and 5) increase the Safety Limit MinimumCritical Power Ratio (SLMCPR) for two recirculation loops in operation.The following is a list of the proposed changes to the NMP2 TSs:* Revise Safety Limit (SL) 2.1.1.2 by increasing the SLMCPR for two recirculation loops inoperation from > 1.07 to > 1.09" Revise the acceptance criterion in TS 3.1.7, "Standby Liquid Control (SLC) System,"Surveillance Requirement (SR) 3.1.7.7 by increasing the discharge pressure from> 1,327 pounds per square inch gauge (psig) to > 1,335 psig" Revise the aeceptance cr-iter-ion int TS SR 3.1.7.10 by inremasing the sodium pentaborateboron 10 enr-iehment Fcquir-ement from -ý 25 atom percent to ý! 92 atom percent, and make-aCorespending e hange in Ti S Figure 3.1.7 1, "Sodium Pentabortec SolutieoVolumne/Conentration" Requiements"" Revise TS Figure 3.1.7 I to account for the deer-ease in the miniftmum volume of the SLS t"nfrom 4,558.6 gallons and 1,288 gallons at sodium pentabConte ionoentations of 13.6oT and14.49%, respectively, to 1,600 gallons and 1,530 gallons at sodium pentaborete concentrationsof 13.6%4 and 14.41%, r-espectively* Change the Required Actions for Condition F of TS 3.3.1.1, "Reactor Protection System(RPS) Instrumentation"*Change Condition G of TS 3.3. 1.1* Add new Conditions J and K to TS 3.3.1.1* Correct an editorial error in Note 3 to TS SR 3.3.1.1.13 (i.e., "ORRM" is changed to"OPRM")* Eliminate TS SR 3.3.1.1.16 and references to it in TS Table 3.3.1.1-1, "Reactor ProtectionSystem Instrumentation"* Change the allowable value (AV) for TS Table 3.3.1.1-1, Function 2.b, Average PowerRange Monitor (APRM) -Flow Biased Simulated Thermal Power (STP) -Upscale from"5 0.55W+60.5% [Rated Thermal Power] RTP and 5 115.5% RTP" to"< 0.61W + 63.4% RTP and < 115.5% RTP"* Add a new note to TS Table 3.3.1.1-1, Function 2.b that requires the Flow Biased SimulatedThermal Power -Upscale scram setpoint to be reset to the values defined by the CoreOperating Limits Report (COLR) to implement the Automated Backup Stability Protection(BSP) Scram Region in accordance with Required Action F-,2-4F.2 of TS 3.3.1.1I of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE" Add a new note to TS Table 3.3.1.1-1, Function 2.e, Oscillation Power Range Monitor(OPRM) -Upscale to denote that following implementation of DSS-CD, DSS-CD is notrequired to be armed while in the DSS-CD Armed Region during the first reactor startup andduring the first controlled shutdown that passes completely through the DSS-CD ArmedRegion. However, DSS-CD is considered operable and capable of automatically arming foroperation at recirculation drive flow rates above the DSS-CD Armed Region* Change the mode of applicability for TS Table 3.3.1.1-1, Function 2.e, OPRM-Upscale fromMode I to > 18% RTP" Change the allowable value for TS Table 3.3.1.1-1, Function 2.e from "As specified in theCOLR" to "NA", Add a pr-.hibiti'n t. TS Limiting Condition for Operation (LCO) 3.4.1, "Recirculation LoopsOperating," is modified to prohibittimhat-pfhbits operation in the Maximum Extended LoadLine Limit Analysis (MELLLA) domain or MELLLA+ expanded operating domain asdefined in the COLR when in operation with a single recirculation loop* Add Required Action B.2 to TS 3.4.1 to identify that intentional operation in the MELLLAdomain or MELLLA+ domain as defined in the COLR is prohibited when a recirculationloop is declared "not in operation" due to a recirculation loop flow mismatch not within limits" Revise TS 5.6.5.a.4 to replace "Reactor Protection System Instrumentation Setpoint for theOPRM -Upscale Function Allowable Value for Specification 3.3.1.1" with "The ManualBackup Stability Protection (BSP) Scram Region (Region I), the Manual BSP ControlledEntry Region (Region II), the modified APRM Simulated Thermal Power -High setpointsused in the OPRM (Function 2.e), Automated BSP Scram Region, and the BSP Boundary forSpecification 3.3.1.1"" Add TS 5.6.8, "OPRM Report," to define the contents of the report required by new RequiredAction F-.24F.3 of TS 3.3.1.1Nuclear Regulatory Commission (NRC) approval of the requested operating domain expansionwill allow NMP2 to implement operational changes that will increase operational flexibility forpower maneuvering, compensate for fuel depletion, and maintain efficient power distribution inthe reactor core without the need for more frequent rod pattern changes. MELLLA+ supportsoperation of NMP2 at Current Licensed Thermal Power (CLTP) of 3,988 Megawatts -Thermal(MWth) with core flow as low as 85% of rated core flow. By operating in the MELLLA+ domain,a significantly lower number of control rod movements will be required than in the presentoperating domain. This represents a significant improvement in operating flexibility. It alsoprovides safer operation, because reducing the number of control rod manipulations:(a) minimizes the likelihood of fuel failures and (b) reduces the likelihood of accidents initiatedby reactor maneuvers required to achieve an operating condition where control rods can bewithdrawn.Attachments 8 and 10 provide the non-proprietary and proprietary versions of the MELLLA+Safety Analysis Report (MELLLA+ SAR), respectively. The MELLLA+ SAR follows theguidelines contained in GE-Hitachi Nuclear Energy Americas (GEH) Licensing Topical Report(LTR) NEDC-33006P-A, Revision 3, "Maximum Extended Load Line Limit Analysis Plus"(MELLLA+ LTR) (Reference 1). The MELLLA+ SAR provides the technical bases for thisrequest and contains an integrated summary of the results of the underlying safety analyses andevaluations performed specifically for the NMP2 expanded operating domain.The MELLLA+ SAR also provides the analyses to change the NMP2 stability solution fromOption III to DSS-CD and use the GEH analysis code TRACG04. DSS-CD as required by theMELLLA+ LTR Safety Evaluation Report. DSS-CD is being implemented using the guidelines2 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEcontained in GEH LTR NEDC-33075P-A, Revision 7, "General Electric Boiling Water ReactorDetect and Suppress Solution -Confirmation Density," (Reference 2). The use of TRACG04 isbeing implemented using the guidelines contained in GEH LTR "DSS-CD TRACG Application,"NEDE-33147P-A, Revision 4, August 2013 (Reference 3).The proposed change to the SLMCPR value for two recirculation loops in operation is based onan analysis performed by Global Nuclear Fuel (GNF) for NMP2 during Cycle 15 operations withMELLLA+ conditions. The GNF report, GNF-0000-0156-7490-RO, "GNF AdditionalInformation Regarding the Requested Changes to the Technical Specification SLMCPR," datedAugust 26, 2013, supports changing the two recirculation loops in operation value of SLMCPRfrom > 1.07 to >_ 1.09, and maintaining the single recirculation loop in operation value ofSLMCPR at > 1.09. These values are based on NRC approved methods and procedures.Attachments 9 and 11 of this Enclosure provide non-proprietary and proprietary versions of theGNF report, respectively.Attachments 10 and 11 of this Enclosure contain information considered to be proprietary asdefined by 10 CFR 2.390. GEH and GNF, as the owners of the proprietary information inAttachments 10 and 11, respectively, have executed the affidavits provided in Attachments 6and 7 to this Enclosure detailing the reasons for withholding the proprietary information.Attachment 3 delineates the regulatory commitments associated with the proposed change.2.0 DETAILED DESCRIPTION2.1 Background2.1.1 MELLLA+Operation of Boiling Water Reactors (BWRs) requires that reactivity balance be maintained toaccommodate fuel burn-up. BWR operators have two options to maintain this reactivity balance:(a) control rod movements or (b) core flow adjustments. Because of the strong void reactivityfeedback and its distributed effect through the core, flow adjustments are the preferred reactivitycontrol method. Operation at low-flow conditions at rated power level also increases the fuelcapacity factor through spectral shift and the increased flow region compensates for reactivityreduction due to fuel depletion during the operating cycle.At NMP2, an Extended Power Uprate (EPU) was implemented by extending the MELLLAoperating domain up to the EPU power level (3,988 MWth). The extension of the MELLLA lineto EPU power levels reduces the available core flow window. In addition, the increased corepressure drop with EPU limits the recirculation flow capability. Consequently, EPU plantsgenerally operate with a reduced core flow window and compensate for reactivity loss withcontrol rod movement. Operation in the MELLLA+ expanded operating domain will provide alarger core flow window for NMP2.In June 2009, the NRC approved the use of the MELLLA+ LTR (NEDO-33006P-A)(Reference 1) as a basis for MELLLA+ operating domain expansion license amendment requests,subject to limitations specified in the MELLLA+ LTR and in the associated NRC safetyevaluation. The NMP2 request complies with the specified limitations and conditions asdiscussed in Appendix B of Attachments 8 and 10.3 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEIn January 2008, the NRC approved the use of the DSS-CD LTR (NEDC-33075P) as a basis forimplementing DSS-CD as a stability solution to replace the Option III solution in licenseamendment requests, subject to limitations specified in the DSS-CD LTR and in the associatedNRC safety evaluation. The NMP2 request complies with Revision 7 of NEDC-33075P(Reference 2), including the specified limitations and conditions as discussed in Appendix C ofAttachments 8 and 10.The TRACG code for use in DSS-CD applications (NEDE-33147P-A) was approved by NRC inNovember 2007. The NMP2 request complies with Revision 4 of NEDE-33147P-A(Reference 3).In addition, the NRC approved the Applicability of GE Methods to Expanded Operating DomainLicensing Topical Report (NEDC-33173P-A) which imposes limitations and requirements for theuse of GEH Methods in expanded operating domains including power uprates and MELLLA+domains. The NMP2 request complies with Revision 4 of NEDC-33173P-A (Reference 4),including the specified limitations and conditions as discussed in Appendix A of Attachments 8and 10.Detailed evaluations of the reactor, engineered safety features, power conversion, emergencypower, support systems, and design basis accidents were performed and are provided inAttachments 8 and 10. These evaluations demonstrate that NMP2 can safely operate in theMELLLA+ expanded operating domain with DSS-CD as the thermal hydraulic stability solution.2.1.2 Standby Liquid Control System is.topi. Enrichment of Boron 1NMPNS proposes to inr-easc the isotopi. of beroen 10 in the sodium pentaboatesolution used to prcparc the nouttron absortber- solution in the Standby Liquid Control Systom(SLS) to ý! 92 atom percoent. The proposed berefn 10 enrielifitnt value allows the minimfum netsolution volume storoed in the Slug storage tank to be deer-eased to 1,530 gallons EAt 14.4% sodiumpentaberate eoncentraion and 1,600 gallons at 13.6% sodium pentaborae concontration. Iadditiei-,NMPNS proposes to increase the acceptance criterion for the SLS pump dischargepressure from 1,327 psig to:- 1,335 psig.2.1.3 Safety Limit Minimum Critical Power RatioNMPNS proposes to revise SL 2.1.1.2 by increasing the SLMCPR for two recirculation loops inoperation from > 1.07 to ! 1.09. The proposed change to the SLMCPR value for tworecirculation loops in operation is based on an analysis performed by GNF for NMP2 duringCycle 15 operations with MELLLA+ conditions. The GNF report, GNF-0000-0156-7490-RO-P,"GNF Additional Information Regarding the Requested Changes to the Technical SpecificationSLMCPR," dated August 26, 2013, supports changing the two recirculation loops in operationvalue of SLMCPR from ! 1.07 to ! 1.09, and maintaining the single recirculation loop inoperation value of SLMCPR at ý 1.09. These values are based on NRC approved methods andprocedures. Attachments 9 and I11 of this Enclosure provide non-proprietary and proprietaryversions of the GNF report, respectively.2.2 Proposed Changes to the Nine Mile Point Unit 2 Technical SpecificationsNMP2 TS changes are required to allow operation in the expanded MELLLA+ operating domain,use of DSS-CD, inerease the isotopic enrzeinhmcnt of boroen 10 in the sodium pentaborate solution4 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEused to prepare the ne.tr.n abs.r.b.r. s.lution in the SLS, and increase the SLMCPR for tworecirculation loops in operation. Attachment 1 of this Enclosure provides a mark-up of the NMP2TS showing the proposed changes. Attachment 2 of this Enclosure provides a mark-up of theNMP2 TS Bases to show the corresponding changes to the TS Bases. Attachment 2 is providedfor information only. A description of each TS change is provided below.Safety Limit 2.1.1.2, Safety Limit Minimum Critical Power RatioSL 2.1.1.2 is revised to increase the SLMCPR for two recirculation loops in operation from> 1.07 to 1.09.TS 3.1.7, Standby Liquid Control (SLC) SystemTS SR 3.1.7.7 is revised to increase the acceptance criterion for the Standby Liquid ControlSystem (SLS) pump discharge pressure from > 1,327 psig to > 1,335 psig.T-S ISR 3.167. 10 is r-evised to iner-ease the boron 10 cnrichmcnt requirement of sodium pcntabrteffrom ? 25 eatm perceent to ? 92 atom perceent. in addition TS Figure 3.1.7 1 is updated to r-eocthe ineroease in the boront 10 cnriehment roquirement.TIS Figure 3.1.7 1, "Sodium Pentab-ratm Solution ati-n Requir.ements," isrevised to aecount for the change in the net volufe in the SL9 tank that arise thecnr-iehment iner-ease. The miniftmum volumne is changed from 4,558.6 gallonts and 4,288 gallons asodium pcntaber-ate coneentrationis of 13.6% and 14.4%, respeetively, to 1,600 gaillons and 1,530gallonis at a sodium pentaberate coneentration of 13.6% and 14.44, r-espeetively.TS 3.3.1.1, Reactor Protection System (RPS) InstrumentationRequired Actions F. 1 and F.2 of TS 3.3.1.1 and their associated Completion Times are replacedwith the following new Required Actions and Completion Times.REQUIRED ACTIONCOMPLETION TIMEF. I Initiate Action to implement the ManualBSP Regions defined in the COLR.ANDF-.2-4-F._2 Implement the Automated BSPScram Region using the modifiedAPRM Simulated Thermal Power -High scram setpoints defined in theCOLR.ANDR.-,F.__3 Initiate action in accordancewith Specification 5.6.8.Immediately12 hours90 dayslmmediately5 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGECondition G is modified to no longer apply in the event a Required Action and associatedCompletion Time of Condition F is not met.New Condition J (see below) is added to address the action to take in the event a Required Actionand associated Completion Time of Condition F is not met.New Condition K (see below) is added to address the action to takeAction and associated Completion Time of Condition J is not met.in the event a RequiredCONDITIONREQUIRED ACTIONCOMPLETION TIMEJ. Required Action andassociated Completion Timeof Condition F not met.J. 1 Initiate action toimplement the ManualBSP regions defined in theCOLR.ANDJ.2 Reduce operation to belowthe BSP Boundary definedin the COLR.ANDJ.3 ---------NOTE-------LCO 3.0.4 is not applicableImmediately12 hours120 daysRestore required channel tooperable.K. Required Action andassociated Completion Timeof Condition J not met.K.1ReducePOWERRTPTHERMALto less than 18%4 hours"ORRM" is changed to "OPRM" in Note 3 to TS SR 3.3.1.1.13.TS SR 3.3.1.1.16 is eliminated, and references to it in TS Table 3.3.1.1-1 are eliminated.TS Table 3.3.1.1-1, Function 2.b, Flow Biased Simulated Thermal Power -Upscale, contains botha flow-biased AV (_ 0.55W + 60.5% RTP) and a fixed AV at 115.5% RTP. The flow-biased AVwill be changed to (< 0.6 1W + 63.4% RTP).A new note is added to TS Table 3.3.1.1-1, Function 2.b that requires the Flow Biased SimulatedThermal Power -Upscale scram setpoint to be reset to the values defined by the COLR toimplement the Automated Backup Stability Protection (BSP) Scram Region in accordance withRequired Action F.2. F.2 of TS 3.3.1.1.A new note is added to TS Table 3.3.1.1-1, Function 2.e, OPRM -Upscale, to denote thatfollowing DSS-CD implementation, DSS-CD is not required to be armed while in the DSS-CDArmed Region during the first reactor startup and during the first controlled shutdown that passes6 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEcompletely through the DSS-CD Armed Region. However, DSS-CD is considered operable andcapable of automatically arming for operation at recirculation drive flow rates above the DSS-CDArmed Region.The mode of applicability for TS Table 3.3.1.1-1, Function 2.e, OPRM-Upscale is changed fromMode I to > 18% RTP.In addition, the allowable value for Function 2.e is changed from "As specified in the COLR" to"4NA."TS 3.4.1, Recirculation Loops, OperatingLCO 3.4.1 is modified to include an additional provision that will prohibit intentional operation inthe MELLLA domain or the MELLLA+ domain as defined in the COLR when only a singlerecirculation loop is in operation. It will state:"... One recirculation loop shall be in operation provided the plant is not operating in theMELLLA or MELLLA+ domain defined in the COLR and provided the following limits areapplied when the associated LCO is applicable:..."A new Required Action B.2 is added to prohibit intentional operation in the MELLLA domain orthe MELLLA+ domain defined in the COLR in the event a recirculation loop is declared to be"not in operation" due to a recirculation loop flow mismatch. The Completion Time for this newRequired Action is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.TS 5.6.5, Core Operating Limits Report (COLR)TS 5.6.5.a is modified by replacing the reference to "Reactor Protection System InstrumentationSetpoint for the OPRM -Upscale Function Allowable Value for Specification 3.3.1.1" with areference to "The Manual Backup Stability Protection (BSP) Scram Region (Region 1), theManual BSP Controlled Entry Region (Region II), the modified APRM Simulated ThermalPower -High setpoints used in the OPRM (Function 2.e), Automated BSP Scram Region, andthe BSP Boundary for Specification 3.3.1.1 ."TS. 5.6.8, OPRM ReportThe following new report requirement is added as TS 5.6.8, "OPRM Report:""When a report is required by Required Action F-,2-F.3 of TS 3.3.1.1, "RPSInstrumentation," a report shall be submitted within 90 days of ,nteing CONDITION Fthefollowing 90 days. The report shall outline the preplanned means to provide backup stabilityprotection, the cause of the inoperability, and the plans to- and schedule for restoring therequired instrumentation channels to OPERABLE status."The new TS section is numbered TS 5.6.8, because on November 21, 2012, NMP2 submitted aLicense Amendment Request (LAR) to create a new TS section that is numbered TS 5.6.7 for theReactor Coolant System Pressure and Temperature Limits Report (Reference 5). NMPNSanticipates that LAR will be approved by the NRC and implemented at NMP2 prior to approvalof the MELLLA+ LAR. The numbering of TS 5.6.8 is an administrative consideration. The7 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEMELLLA+ LAR is independent of the LAR submitted on November 21, 2012. NRC approval orrejection of Reference 5 would have no technical impact on the MELLLA+ LAR.2.3 Modification SummaryThe MELLLA+ core operating domain expansion does not require major plant hardwaremodifications. The core operating domain expansion involves changes to the core power/flowmap and a small number of setpoints and alarms. Because there are no increases in the operatingpressure, power, steam flow rate, and feedwater flow rate, there are no major modifications toother plant equipment.The stability solution is being changed from Option III to the DSS-CD solution. The DSS-CDsolution algorithm, licensing basis, and application procedures are generically described in theDSS-CD LTR (Reference 2), and are applicable to NMP2. The DSS-CD solution uses the samehardware as the current Option III solution. DSS-CD requires a revision to the existing stabilitysolution software.The boroen 10 enrichmfent in the sodium pentabor-ate solution in the SLS is incr-eased fo! 25 atom pcrccnt to ! 92 atom perceent. The increase in the boroen 10 enriehment in the scdiumpentaborate solution for- the SLS is sufficient to decrease the sodium pentabernte solution velumesterce in tMle SLS storage tank. In a..iti.n, tc I ne SLS pump discharge pressure acceptancecriterion is changed to > 1,335 psig. Changes to instrumentati. n setp..ots will be made toaccotunt for these changes. The increase in the SLMCPR for two recirculation loops in operationdoes not require any physical modifications to structures, systems, or components.

3.0 TECHNICAL EVALUATION

3.1 MELLLA+Attachments 8 and 10 of this Enclosure provide non-proprietary and proprietary versions of the"Safety Analysis Report for Nine Mile Point Unit 2 Maximum Extended Load Line LimitAnalysis Plus (MELLLA+ SAR)," NEDO-33576NP and NEDC-33576P, respectively. TheMELLLA+ SAR summarizes the results of the significant safety evaluations performed thatjustify the expansion of the core flow operating domain for NMP2. The changes expand theoperating domain in the region of operation with less than rated core flow, but do not increase thelicensed power level or the maximum core flow. The expanded operating domain is identified asMELLLA+.The scope of evaluations required to support the expansion of the core flow operating domain tothe MELLLA+ boundary is contained in NEDC-33006P-A, "Maximum Extended Load LineLimit Analysis Plus," referred to as the MELLLA+ LTR (Reference 1). The MELLLA+ SARprovides a systematic disposition of the MELLLA+ LTR subjects applied to NMP2, includingperformance of plant-specific assessments and confirmation of the applicability of genericassessments to support a MELLLA+ core flow operating domain expansion. The MELLLA+operating domain expansion is applied as an incremental expansion of the operating boundarywithout changing the maximum licensed power, maximum core flow, or the current plant vesseldome pressure. The MELLLA+ SAR supports operation of NMP2 at a licensed thermal power of3,988 MWt with core flow as low as 85% of rated core flow.8 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEThe MELLLA+ core operating domain expansion does not require major plant systemsmodifications. NMP2 will implement the DSS-CD solution in accordance with the applicableLTRs (References 3 and 4), including the applicable limitations and conditions. Implementationof DSS-CD requires a revision to the existing stability solution software.The core operating domain expansion involves changes to the operating power/core flow map andchanges to a small number of instrument setpoints. Because there are no increases in theoperating pressure, power, steam flow rate, and feedwater flow rate, there are no significanteffects on the plant systems outside of the Nuclear Steam Supply System (NSSS). There is apotential increase in the steam moisture content at certain times while operating in theMELLLA+ operating domain. The effects of the potential increase in moisture content on plantsystems have been evaluated and determined to be acceptable. The MELLLA+ operating domainexpansion does not cause additional requirements to be imposed on any of the safety, balance-of-plant, electrical, or auxiliary systems. No changes to the power generation and electricaldistribution systems are required as a result of the MELLLA+ operating domain expansion.This report also addresses applicable limitations and conditions as described in the MELLLA+LTR SER for the GEH LTR NEDC-33173P, "Applicability of GE Methods to ExpandedOperating Domains" (Methods LTR SER) (Reference 4). A complete listing of the applicableLTR SER limitations and conditions and the sections of the MELLLA+ SAR which address themare presented in Appendices A, B, and C of the MELLLA+ SAR.Only previously NRC-approved or industry-accepted methods were used for the analyses ofaccidents and transients. Therefore, because the safety analysis methods have been previouslyaddressed, the details of the methods are not presented for review and approval in the MELLLA+SAR. Also, event and analysis descriptions that are already provided in other licensing reports orthe NMP2 Updated Safety Analysis Report (USAR) are not repeated within the MELLLA+ SAR.Evaluations of the reactor core and fuel performance, reactor coolant and connected systems,engineered safety features, instrumentation and control, electrical power and auxiliary systems,power conversion systems, radwaste systems and radiation sources, reactor safety performanceevaluations were performed. The MELLLA+ SAR summarizes the results of the evaluations thatjustify the MELLLA+ operating domain expansion to a minimum core flow rate of 85% of ratedcore flow at 100% RTP.Section 11.3.1 of Attachments 8 and 10 provides a summary of the modifications that will berequired to implement the MELLLA+ operating domain, DSS-CD, and the changes to the SLS.Section 11.3.2 of Attachments 8 and 10 provides a summary of the MELLLA+ issues including adiscussion of the MELLLA+ analysis basis, fuel thermal limits, makeup water sources, designbasis accidents, challenges to fuel, challenges to the containment, design basis accidentradiological consequences, anticipated operational occurrence analyses, combined effects,non-Loss of Coolant Accident (LOCA) radiological release accidents, equipment qualification,balance-of-plant, and environmental consequences.An assessment of the risk increase, including core damage frequency (CDF) and large earlyrelease frequency (LERF) associated with operation in the MELLLA+ operating domain isprovided in Attachment 4 of this Enclosure and Section 10.5 of Attachments 8 and 10 of thisEnclosure. The estimated risk increase for at-power events due to MELLLA+ is a delta CDF of1E-8 and delta LERF of 3E-9. This represents a very small risk change in RG 1.1749 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE(Reference 6). Based on these results, the proposed MELLLA+ operating domain is acceptableon a risk basis.3.2 DSS-CDThe long-term stability solution is being changed from the currently approved Option III solutionto DSS-CD. The DSS-CD solution algorithm, licensing basis, and application procedures aregenerically described in NEDC-33075P-A (Reference 2) and NEDE-33147P-A (Reference 3),and are applicable to NMP2 including any limitations and conditions associated with their useand approval. Section 2.4 of the MELLLA+ SAR (Attachments 8 and 10) addresses the changeto the DSS-CD stability solution. In addition, a complete listing of the required DSS-CD SERand limitations and conditions and the sections of the MELLLA+ SAR which address them ispresented in Appendix C of the MELLLA+ SAR, respectively.DSS-CD is designed to identify the power oscillation upon inception and initiate control rodinsertion (scram) to terminate the oscillations prior to any significant amplitude growth exceedingthe applicable safety limits. DSS-CD is based on the same hardware design as Option III.However, it introduces an enhanced detection algorithm that detects the inception of poweroscillations and generates an earlier power suppression trip signal exclusively based .n........ :.... :ri.d r-..gni. .The existing Option III algorithms are retained (withgeneric setpoints) to provide defense-in-depth protection for unanticipated reactor instabilityevents.3.3 Standby Liquid Control System Boron 10 En-.hmen-The SLS is described in Section 9.3.5 of the NMP2 USAR. The system provides a backupcapability for shutting down the reactor. The SLS is needed only in the event that sufficientcontrol rods cannot be inserted into the reactor core to accomplish shutdown and cooldown in thenormal manner. To accomplish this function, the SLS injects a sodium pentaborate solution intothe reactor. The SLS consists of a boron solution storage tank, two positive displacement pumps,two explosive valves (provided in parallel for redundancy), and associated piping and valves usedto transfer borated water from the storage tank to the reactor pressure vessel (RPV). The boratedwater solution is discharged into the RPV through the high pressure core spray sparger.The specified neutron absorber solution is sodium pentaborate. It is prepared by dissolvinggranularly-enriched sodium pentaborate in demineralized water (NMP2 USAR 9.3.5.2). Thesodium pentaborate solution is discharged radially over the top of the core through the HighPressure Core Spray (HPCS) sparger. The boron absorbs thermal neutrons and thereby terminatesthe nuclear fission chain reaction in the uranium fuel. The sodium pentaborate also acts as abuffer to maintain the suppression pool pH at or above 7.0 to prevent the re-evolution of iodine,when mixed in the suppression pool following a LOCA accompanied by significant fuel damage(NMP2 USAR Section 9.3.5.1).3.3.1 Reactor Boron Cold Shutdown Concentration RequirementsThe reactor boron concentration requirements for achieving cold shutdown (780 parts per million(ppm) natural boron) is not increased for MELLLA+, because there is no change in fuel type andno change to the operating cycle. The total weight of boron-10 required for cold shutdown(including the 25% margin) does change for MELLLA+, because of a conservative increase inthe assumed weight of the reactor coolant in the applicable analysis.10 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEThe shutdown margin is calculated for each plant reload and is documented in the SupplementalReload Licensing Report (SRLR).Ii ,"b at'-N 1 4 lift Z-- * ----AI nSLflpfl~C in imron ~u ~nricnmcnr in ~ou:um rcnuwarurc10 CFR 50.62(e)(4) rvauir-es:"Each boiling water -retor must have a standby liquid contrl system (SCS) withtheapability of injecting into the reacter- prcssurce vessel a bor-ated water- solutiont at such a florate, level of bor-on concentfation and borefn 10 isotope enrichmcnt, and aecountfing or-eactor pressure vessel voluime, that the resulting reaetivity entrl is at least equivalent tothat r-esulting from injection of 86 gallons per- minuite of 13 weight perceent sodiumpentaborate dceahydrate solutioin at the nAtur-al boroen 10 isotope abundance into a 251 inchinside diameter- r.ea.t. r pressure .ss. l for- a gien ... design..."The NRC approved licensing topical r-eport NEDE 31096 A (Rcfer-enee 7) provides a method*bwhich the boront equivalency r-equirement of 10 CFR 50.62(e)(4) can be demonefstrated.Equation 1 1 of that docuiment was used to demonstfate injectiont capacity equivalency as"elews.(Q/'86) x (M2511') x (C-'13) x (9/'49.8) 1 1Where-Q -expeeted SLS flew rate (gpmn)A425-L mass of water in the rcaetfr ivessel and recircuat~ion system at hot ratedconditions (Ibs) for- a 251 inceh diameter- vessel reference plant-l mass of water in the NMP2 reactor- vessel and r-eeir-eulation system at hot ratedeendifiefts Ibs)-sodium pentaborae solution concentreation (%494)E boron 10 isotope enrichment (atom percent)NMP2 is equipped with a 251 inch diameter reactor- vessel (NMP2 USAR Section 15.)assumptions utilized int the analyses of the changes to the SLS.Substituting the cuffent values definted in Table I in the above equationt yields:82.4/86 X 1 X 13.6413 X 25/19.8 -l.27ý--4Substituting the new values definted in Table 1 into the above equation yields:!80/86 X 1 X 13.6/13 X 92/19.8 -4.52> 1 for 13.6 v,0/_)This that the equivalent control capacity requirement of 10 CFR 50.62(e).4)is m et, when the changes to the SLS flow rate and the boront 10 isotope em-iehment are inceluded.int addition, the control margin ince s. This is due to -nefeasing the boroen 10 enrichment tennin the equation by a factor- of3.68 (i.e., 92'25 -3.68).-11 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEf41re: -T.I-. .....requre c ..I sm)I -1pump pto nave a o eow ra.. A at ii ' ...gpml.Maintaining the TS SR 3. 1.7.7 aeceptance cr-iter-ia for SLC pump flow rate at 41.2 gpmf providesmar~gin with r-espcct to the r~equired flow rate for- AT-WS mitigation. This issue has beeaddressed for- euffent eperation via the NMP-NS Correcetive Action System.As defined in Table 1, the analyzed SLS injection flow Fate is r-educed to 80 gpm flow r-ate otwo SLC System pumps in .perati.n to a ..ount for- diluti.n ..fe.ts identified by GEH4 SafetyCommunicationt 10 13, Standby Liquid C.ntrol System Dilution Flew, with additional margin.Table 1 Assumptions rega.rding SLS P.rforman.Pafametef Units Current Vaulue New-ValueReactor boron ..n.. ntatin for pp 7. ...0cold shutdown (natur-al boront)Maximuma allowable solutionf V.+% 144. -144conceentrationtMinimum allowable solutio 4-3-A 4-3-6conccntfatienSolution eoncentration assumed in Y% 4-3 4446ATWS anal-isi__Mfinimumn boroen 10 enrichment foi Aom 2- 9-2ATWS aftaly~fisDesigni-SLS pumip flew Faite 813M 4-54-Minimum SLS pump flow rate as gPm" 44l-2 4442defined in T-S 3.1.7Sing pump flow rate (two. pup i pin 92-4 80)3-.33.3.2 Change in SLS Pump Discharge Pressure Acceptance CriterionTS SR 3.1.7.7 is revised to increase the acceptance criterion for the SLS pump discharge pressurefrom >_ 1,327 psig to > 1,335 psig. This change is required due to the increase in the peak upperplenum pressure after SLS pump startup to 1,241 pounds per square inch -absolute (psia) asidentified in Tables 9-4 and 9-7 of Attachments 8 and 10 of this Enclosure. Currently, the peakupper plenum pressure after SLS pump startup is 1,236 psia. Thus, the ATWS analysis forMELLLA+ establishes a pressure differential of five psi. The SLS pump discharge pressureacceptance criterion in TS SR 3.1.7.7 is increased by eight psig to address the increase in theupper plenum pressure and provide an additional three psi margin.33.43.3.3 Anticipated Transient without SCRAMSection 9.3.1 of the MELLLA+ SAR (Attachments 8 and 10 of this Enclosure) provides asummary of the plant-specific analyses of Anticipated Transients Without Scram (ATWS) todemonstrate that the ATWS acceptance criteria are met for operation in the MELLLA+ operatingdomain. NMP2 meets the ATWS mitigation requirements in 10 CFR 50.62 for an alternate rodinsertion (ARI) system, SLS boron injection equivalent to 86 gpm, and automatic RPT logic. The12 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEplant-specific ATWS analyses take credit for the ATWS-RPT and SLS. However, ARI is notcredited.Section 9.3.1.1 of the MELLLA+ SAR (Attachments 8 and 10 of this Enclosure) provides alicensing basis ODYN ATWS analysis that demonstrates that the ATWS acceptance criteriawould be met in the event of a NMP2 response to an ATWS event initiated in the MELLLA+operating domain.In addition, a plant-specific ATWS analysis was performed at MELLLA+ conditions thatassumed operation of a single SLS pump. The analysis and the results are discussed in Section9.3.1.1.1 of the MELLLA+ SAR (Attachments 8 and 10 of this Enclosure). It concludes: "AllATWS acceptance criteria are met at MELLLA+ conditions with only a single SLS pumpoperating."3.3.5 Suppression Pool BufferingIn supp.rt of the Alte.ate Source Term (AST-) meth.dology, the SLS also provides suppr.ession.peel buffer-ing following a LOCA accompanied by significant fuel damage, preventing r-eevelution of iodine from the suppr-ession peol by maintaining the pool pH4 above 7.0. Seetion9.3.5.1 of the NMP2 USAR requires a sufficient coneentration and quaftity of sodiumpentaborate to be available for injcetion into the reaetor vessel to controel pH4 in the suppr-esiopool for 30 days fllwing a DBA LOCAi.The r1eduetiont int the mi r u :ired solution volume results in a reduction in. the e ssolutiont available for- injectionl to maneneppeso ol p14I 7.0 for- 30 days post LOCA.The minimum" sodium pentabor-ate solution vouei rqie for injection post LOCA foradequate p14 eontrol is 1,065 lons at the limiting t.e., a sodium pentaborateconcentration of 13.6%4). The minimum requir-ed tank vollumie at4 al concentration of 13.6 %ireduced frim 4,558.6 gallons to 1,600 gallons. While this does r educe the amount of cxeavailable solutiont, adequate margin is maintaincd to ensure that the SLS can perform its requiredAST support funetion.The proposed bor-on 10 enr-ichment changes do not impaet the capability to achieve and maintaina pH4 above 7.0 in the suppression pool following a LOCA, because the chemnical pr-opeffies nonenetration of the sodium pentaborate solution injected into the suppr-essin pool will remainthe same. Given the rduced volume of solution that wil be available, ther will be a two hourr-eduction int the maximum tome available to add boron to the supeso pool to ananpabove 7.0 (noeminal time based on low level alann is within 22- housvess h-- ~ttieowithin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />). Are.vie. of the Emergency Operating Procedures confirmed that the sodiumpentaborate solutiont would be injeeted within 30 minutes following the eccunfeftee of LOCA.The maaximutm 22 hour- timne period proevides a large margin to the minimm requremet formanual operator- action toinec the sodium pentaborate solution of 30 minuatesq. In addition, theSuppression pool pisntepected to drop bielw:7 foseelda.Section 9.3.5.3 of the NMP2 USAR delineates that ontly one of the two SL.S loops was assumefor suppr-ession pool p1H control operation. The proeposed changes to the SLS do not affect thI., ...t (.... C' I1.-I...-- -It% A L ... .+I^ : l ..,i;. .~~~~0~~ ---- ---- --- ---I--I-- --------------------0~.... ' .. .... ....J .... ..... .rrI-of313 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE"J f J"QI- --3.3.C n~ige tin 818 Sterage I unit Soluiuon -VolumeThe pr-opescd bOroni 10 enriehment value allows the minimum solutiont Volumfe stor-ed in theSLsterage tank th be dedreased to 1,530 gallons at a sodium.. pentabrate e.ne.ntr..in of 14.4% and1,600 gallons at a sodium pentablrVate eoinentraion of 13.6%. The mark uip of NMP2 TS Figt3.1.7 1 promvided inf Attaclhment 1 mf this Entlesur delineates thce preposed hange in the net SLSstorage fankl solutioin volume.The required minimum lumes for- the 13.6 p%49 and 14.4 mar solutin vplumres were der tby deteamininag the m einimum salution veluifne and then increasing the allumne to acotun frei1) the dead vsluine not pumped in the reator0 that remains in the SLS and 14PCS piping;a2) instrument accuracy.The minimum net solution volume for- injection meets all consider-ations for AT-WS boroninjection raes, AST suppression pool pH- control, and assures tha the r-eactor- core boronconcentr-ation will be greater- than :780 ppm natural boron equivalent3-.343.3.3 SLS Pump Relief Valve Setpoint MarginThe SLS pump relief valve setpoint margin is the difference between the relief valve nominalsetpoint and the maximum SLS pump discharge pressure. A margin of 78 psi provides sufficientmargin against inadvertent relief valve lifting. The 78 psi is based on an allowance for the reliefvalve setpoint drift (typically 3% (3% of 1,600 psi = 48 psi)) and SLS pump pressure pulsations(30 psi).For MELLLA+ operation during the limiting ATWS event, the relief valve setpoint margin is205.7 psi. This margin is based on a SLS pump relief valve setpoint of 1552 psig (1600 psig -3% tolerance (i.e., 48 psig)) and subtracting a SLS pump discharge pressure of 1346.3 psig (i.e.,1552 psig -1346.3 psig = 205.7 psi). The margin reduces to 175.7 psi if 30 psi for SLS pumppressure pulsations is taken into consideration (i.e., 205.7 psi -30 psi = 175.7 psi).3.3.8 Net Positive Suction Head Available (NIPSH1) for SbS P-unqwrThe propesed changes include a reduction in the minimum volume for- the SLS storage tank. Tiresults inl at eduction in the swaic head available to provide Net Positive Suction Read (NPSH4Yfotthe SLS pumps. The calculation that determnines the SLS pumnp NPSNft did not take any cedifor- the staic head abo-ve the SLS storage tank zero level. The mninimum tank level cof:Fespondinfgto the minimum net volume pefmitted by the proposed change to Figure 3.1.7 1 is greater- thanthree feet above tankl zero.3.4 Safety Limit Minimum Critical Power RatioCycle specific transient analyses are performed to determine the required SLMCPR and thechange in Critical Power Ratio (CPR) [ACPR] for specific transients. To ensure that adequatemargin is maintained, a design requirement based on a statistical analysis was selected, in thatmoderate frequency transients caused by a single operator error or equipment malfunction shallbe limited such that, considering uncertainties in manufacturing and monitoring the coreoperating state, at least 99.9% of the fuel rods would be expected to avoid boiling transition. Thelowest allowable transient MCPR limit which meets the design requirement is termed the fuelcladding integrity SLMCPR.14 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGENUREG-0800, Standard Review Plan, Section 4.4, "Thermal and Hydraulic Design," AcceptanceCriterion No. I.B, states, in part, that the limiting (minimum) value of CPR is to be establishedsuch that at least 99.9% of the fuel rods in the core would not be expected to experience departurefrom nucleate boiling during normal operation or anticipated operational occurrences.A cycle specific Operating Limit MCPR (OLMCPR) is established to provide adequate assurancethat the fuel cladding integrity SLMCPR is not exceeded for any anticipated operationaltransients. The OLMCPR is obtained by adding the maximum value of ACPR for the mostlimiting transient postulated to occur at the plant to the fuel cladding integrity SLMCPR.3.4.1 Analytical Methods, Standards, Data and ResultsNMPNS proposes to revise SL 2.1.1.2 by increasing the SLMCPR for two recirculation loops inoperation from >_ 1.07 to > 1.09. The proposed change to the SLMCPR value for tworecirculation loops in operation is based on an analysis performed by GNF for NMP2 duringCycle 15 operations with MELLLA+ conditions. The GNF report, GNF-0000-0156-7490-RO-P,"GNF Additional Information Regarding the Requested Changes to the Technical SpecificationSLMCPR," dated August 26, 2013, supports changing the two recirculation loops in operationvalue of SLMCPR from >_ 1.07 to >_ 1.09, and maintaining the single recirculation loop inoperation value of SLMCPR at > 1.09. These values are based on NRC approved methods andprocedures. Attachments 9 and 11 of this Enclosure provide non-proprietary and proprietaryversions of the GNF report, respectively.GNF performed the SLMCPR calculation in accordance with Revision 19 of NEDE-2401 1-P-A,"General Electric Standard Application for Reactor Fuel," (Reference 8) using the followingNRC-approved methodologies and uncertainties:" NEDC-32601P-A, "Methodology and Uncertainties for Safety Limit MCPREvaluations," (August 1999) (Reference 9).* NEDC-32694P-A, "Power Distribution Uncertainties for Safety Limit MCPREvaluations" (August 1999) (Reference 10).* NEDC-32505P-A, "R-Factor Calculation Method for GEl 1, GE12 and GE13 Fuel,"(Revision 1, July 1999) (Reference 11).Section 2.9 of Attachments 9 and 11 of this Enclosure require NMPNS to "provide the currentand previous cycle power/flow map in a separate attachment." Figure 1-1 of Attachments 8 and10 of this Enclosure provide the power/flow operating map for MELLLA+. This will be thepower/flow map for NMP2 operations in Cycle 15 following NRC approval of this LicenseAmendment Request. Attachment 5 of this Enclosure provides the NMP2 power/flow operatingmap for the current operating cycle.3.4.2 Major Contributors to SLMCPR ChangeIn general, the calculated safety limit is dominated by two key parameters: (1) flatness of the corebundle-by-bundle MCPR distribution, and (2) flatness of the bundle pin-by-pin power/R-Factordistribution. Greater flatness in either parameter yields more rods susceptible to boiling transitionand thus a higher calculated SLMCPR. The MCPR Importance Parameter (MIP) measures the15 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEcore bundle-by-bundle MCPR distribution and the R-Factor Importance Parameter (RIP)measures the bundle pin-by-pin power/R-Factor distribution. The impact of the fuel loadingpattern on the calculated two recirculation loops in operation SLMCPR has been correlated to theparameter MIPRIP, which combines the MIP and RIP values.Another factor besides core MCPR distribution or bundle R-factor distribution that significantlyimpacts the SLMCPR is the expansion of the analysis domain that comes with the initialapplication of MELLLA+. The rated power / minimum core flow point is analyzed at a lowercore flow (than without MELLLA+) using increased uncertainties that tend to increase theSLMCPR. Also, a new point at off-rated power / off-rated flow was analyzed using the increaseduncertainties.Table 3 of the GNF analysis (Attachments 9 and 11 of this Enclosure) presents the MIP and RIPparameters for the previous cycle and the current cycle along with the two recirculation loops inoperation SLMCPR estimates using MIPRIP correlations. In addition, Table 3 of the GNFanalysis provided in Attachments 9 and 1 1 presents estimated impacts on the two recirculationloops in operation SLMCPR due to methodology deviations, penalties, and/or uncertaintydeviations from approved values. Based on the MIPRIP correlation and any impacts due todeviations from approved values, a final estimated two loops in operation SLMCPR isdetermined. Section 2.2 of the GNF analysis (Attachments 9 and 11 of this Enclosure) provides adetailed discussion of the items in Table 3 of the GNF analysis (Attachments 9 and 11 of thisEnclosure) that result in the increase in the estimated SLMCPR.3.4.3 Considerations Addressed in the GNF Analysis Regarding R-Factor, Core Flow Rate andRandom Effective Tip Reading, and Fuel Axial Power Shape PenaltySection 2.2.1 of the GNF analysis provides a discussion that justifies an increase in the R-Factoruncertainty value. GNF states that it generically increased the GEXL R-Factor uncertainty toaccount for an increase in channel bow due to the emerging unforeseen phenomena called controlblade shadow corrosion-induced channel bow, which is not accounted for in the channel bowuncertainty component of the approved R-Factor uncertainty. NMP2 has experienced controlblade shadow corrosion-induced channel bow. Accounting for the control blade shadowcorrosion-induced channel bow, the NMP2 Cycle 15 analysis shows an expected channel bowuncertainty which is bounded by the increased GEXL R-Factor uncertainty. Thus, the use of theincreased GEXL RFactor uncertainty value adequately accounts for the expected control bladeshadow corrosion-induced channel bow for NMP2 Cycle 15.Section 2.2.2 of the GNF analysis provides a discussion that identifies that the uncertainty valuesfor the core flow rate and the random effective tip reading in the two recirculation loops inoperation calculation were conservatively adjusted by using the single recirculation loop inoperation uncertainty values. The GNF analysis states the treatment of the core flow and randomeffective TIP reading uncertainties is based on the assumption that the signal to noise ratiodeteriorates as core flow is reduced.Section 2.4 of the GNF analysis provides a discussion regarding higher uncertainties and non-conservative bases in the GEXL correlations for the various types of axial power shapes. GNFdetermined that no power shape penalties were required to be applied to the calculated NMP2Cycle 15 SLMCPR values.16 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE3.4.4 ConclusionThe proposed change to revise SL 2.1.1.2 by increasing the SLMCPR for two recirculation loopsin operation from > 1.07 to > 1.09 is acceptable, and continues to maintain the same level ofsafety as the current licensing basis.3.5 NMP2 TS ChangesTable 2 defines the affected NMP2 TS, describes the change, and defines the supportingAttachment to this Enclosure that supports the TS Change.Table 2 -Changes to NMP2 Technical SpecificationsNMP2 TS Description of the Change SupportingAttachmentSL 2.1.1.2 Increase the SLMCPR for two recirculation loops in Attachments 9 and 11operation from > 1.07 to > 1.09TS 3.1.7 -Increase the SLS pump discharge pressure from Section 6.5.3 ofSR 3.1.7.7 > 1,327 psig to >_ 1,335 psig Attachments 8 and 10TS-83.b.7- hinr-easing the sodium pentaberate ber-on0 Seetiont 6.5.1 ofSR 3.A.7. enr-iehment rzguir-ement from ! 25 atom pcr-eent toAttachments 8 and 10and S92 .tem.peree.T-S Figure Reducig the minimum net vlume n to 1,600 gallt Section 6.5.1 of3.i.74i and 1,530 gallens at Scdium pentabonuats i ngentrmaiens Attachments 8 and 10of 13.6% Sad 14.4%, ersperticlhyT-S-Figure lneroasing the COdi ;an pcItabniate brcon 1 Seetio 6.5.1 in3. i7 I enriehmcn egunt tfrom ! 25 atom pereent toAttachiments 8 and 1092acteforPer."itTS 3.3.1.1 The Required Actions for Condition F are modified to: Complies with DSS-1) Initiate Action to implement the Manual BSP CD LTRRegions defined in the COLR; 2) Implement the Section 2.4 ofAutomated BSP Scram Region using the modified Attachments 8 and 10APRM Simulated Thermal Power -High scramsetpoints defined in the COLR; and 3) Initiate action inaccordance with Specification 5.6.8TS 3.3.1.1 Condition G is modified to no longer apply in the event Complies with DSS-a Required Action and associated Completion Time of CD LTRCondition F is not met. Section 2.4 ofAttachments 8 and 10TS 3.3. 1.1 New Condition J is added to address the action to take Complies with DSS-in the event a Required Action and associated CD LTRCompletion Time of Condition F is not met. Section 2.4 ofAttachments 8 and 1017 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGETable 2 -Changes to NMIP2 Technical SpecificationsNMP2 TS Description of the Change SupportingAttachmentTS 3.3.1.1 New Condition K is added to address the action to take Complies with DSS-in the event a Required Action and associated CD LTRCompletion Time of Condition J is not met. Section 2.4 ofAttachments 8 and 10TS SR Correct an editorial error in Note 3 (i.e., ORRM is Editorial correction3.3.1.1.13 changed to OPRM)TS SR Eliminate TS SR 3.3.1.1.16 and references to it in TS Complies with DSS-3.3.1.1.16 Table 3.3.1.1-1 CD LTRand TS Section 2.4 ofTable Attachments 8 and 103.3.1.1-1TS Table Change the AV for APRM -Flow Biased STP -Section 5.3.1 of3.3.1.1-1, Upscale from ":50.55W+60.5% RTP and < 115.5% Attachments 8 and 10Function 2.b RTP" to "< 0.61W + 63.4% RTP and < 115.5% RTP"TS Table Add a new note that requires the Flow Biased Complies with DSS-3.3.1.1-1, Simulated Thermal Power -Upscale scram setpoint to CD LTRFunction 2.b be reset to the values defined by the COLR to Section 2.4 ofimplement the Automated BSP Scram Region in Attachments 8 and 10accordance with Required Action F--.4F.2 of TS3.3.1.1TS Table Add a new note for Function 2.e, OPRM -Upscale, to Complies with DSS-3.3.1.1-1, denote that following implementation of DSS-CD, CD LTRFunction 2.e DSS-CD is not required to be armed while in the DSS- Section 2.4 ofCD Armed Region during the first reactor startup and Attachments 8 and 10during the first controlled shutdown that passescompletely through the DSS-CD Armed Region.However, DSS-CD is considered operable and capableof automatically arming for operation at recirculationdrive flow rates above the DSS-CD Armed RegionTS Table Change the mode of applicability for TS Table 3.3.1.1- Complies with DSS-3.3.1.1-1, 1, Function 2.e, OPRM-Upscale from Mode 1 to > 18% CD LTRFunction 2.e RTP. Section 2.4 ofAttachments 8 and 10TS Table Change the allowable value from "As specified in the Complies with DSS-3.3.1.1-1, COLR" to "NA" CD LTRFunction 2.e Section 2.4 ofAttachments 8 and 1018 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGETable 2 -Changes to NMP2 Technical SpecificationsNMP2 TS Description of the Change SupportingAttachmentTS.LCO Add a new requirement that pr...i..bModify the LCO Complies with DSS-3.4.1 to prohibit operation in the MELLLA domain or CD LTRMELLLA+ expanded operating domain as defined inthe COLR when in operation with a single recirculation Sections 1.2.4 andloop 3.6.3 2.4 ofAttachments 8 and 10address thatMELLLA+ is notanalyzed for singleloop operationIn addition, NMP2does not currentlypermit single loopoperation while in theMELLLA domain,because it is notanalyzed.TS 3.4.1, Add Required Action B.2 to identify that intentional Complies with DSS-Condition B operation in the MELLLA domain or MELLLA+ CD LTRdomain as defined in the COLR is prohibited when arecirculation loop is declared "not in operation" due to Sections 1.2.4 anda recirculation loop flow mismatch not within limits 3.6.3 2.4 ofAttachments 8 and 10address thatMELLLA+ is notanalyzed for singleloop operationIn addition, NMP2does not currentlypermit single loopoperation while in theMELLLA domain,because it is notanalyzed.19 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGETable 2 -Changes to NMP2 Technical SpecificationsNMP2 TS Description of the Change SupportingAttachmentTS 5.6.5 Replace the reference to "Reactor Protection System Complies with DSS-Instrumentation Setpoint for the OPRM -Upscale CD LTRFunction Allowable Value for Specification 3.3.1.1" Section 2.4 ofwith a reference to "The Manual Backup Stability Attachments 8 and 10Protection (BSP) Scram Region (Region I), the ManualBSP Controlled Entry Region (Region II), the modifiedAPRM Simulated Thermal Power -High setpointsused in the OPRM (Function 2.e), Automated BSPScram Region, and the BSP Boundary for Specification3.3.1.1."TS 5.6.8 Add a new TS section (i.e., TS 5.6.8) to define the Complies with DSS-contents of the report required by new Required Action CD LTRF-.2-.F.3 of TS 3.3.1.1 Section 2.4 ofAttachments 8 and 103.6 TSTF-493There are no effects on the current TS or their licensing bases relative to TSTF-493. Two TSReactor Protection System (RPS) functions are changing in this amendment: (1) the OPRM -Upscale function; and (2) the APRM -Flow Biased Simulated Thermal Power (STP) -Upscalefunction. The OPRM setpoints are unique to a particular core design for a particular fuel cycle.The OPRM function setpoints do not have specific TS allowable values (AVs). The APRM STP -High AVs are specified in TS Table 3.3.1.1-1.MELLLA+ changes the OPRM setpoints in that they are now derived from DSS-CD algorithmsversus Option III algorithms; however, their protective function remains the same. The revisedBases for TS 3.3.1.1 provided in Attachment 2 of this Enclosure states: "The OPRM Upscalefunction settings are not traditional instrumentation setpoints determined under an instrumentsetpoint methodology. There is no Allowable Value for this Function, and the OPRM UpscaleFunction is not [Limiting Safety System Setting (LSSS) Safety Limit (SL)]-related and [the DSS-CD Licensing Topical Report, NEDC-33075P-A] confirms that the OPRM Upscale Functionsettings based on DSS-CD also do not have traditional instrumentation setpoints determinedunder an instrument setpoint methodology."MELLLA+ also changes the APRM -Flow Biased Simulated Thermal Power -Upscale AV fortwo loop operations in the MELLLA+ domain and the APRM -Flow Biased Simulated ThermalPower -Upscale function is used for the Automated Backup Stability Protection (ABSP) if theOPRM becomes inoperable. The APRM STP-High AV and setpoint do have setpointmethodology applied as described in TSTF-493. In addition, the TSTF-493 footnotes werepreviously added to this function in Amendment 140 to the NMP2 Renewed Operating LicenseNPF-69 issued on December 22, 2011 (Reference 12).20 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE3.7 Topics Discussed During NRC Pre-MeetingsOn February 27, 2013, representatives from NMPNS met with the NRC to discuss theMELLLA+ LAR. During this meeting, the NRC sought clarification regarding several topics.Table 3 summarizes those topics, and provides a cross reference to the location in Attachments 8and 10 of this Enclosure that addresses the topic. The NRC issued a summary of this meeting onMarch 13, 2013 (Reference 13).Table 3 -Topics Discussed During NRC Pre-Meeting on February 27, 2013Topic as Summarized in NRC Meeting SummaryIssued on March 13, 2013 (Reference 13)MELLLA+ SARReference (Attachments 8and 10 of this Enclosure)Automated Backup Stability ProtectionThe NMP2 submittal is based on Revision 6 of NEDC-33075P-A. The NMP2 is planning to take exception to Rev 6 relative tothe Automatic Backup Stability Protection (ABSP) set points byusing a simplified method that is consistent with the ABSP setpoint methodology described in Revision 7 of NEDC-33075P.Since the NRC staff has not approved Revision 7 of the LicensingTopical Report (LTR) NEDE-33075P, Re: Detect and SuppressSolution-Confirmation Density (DSS-CD) for Automatic BackupStability Protection (ABSP), the License Amendment Request(LAR) should not refer to revision 7 of NEDE-33075P, butprovide the justifications, consistent with revision 7, for anyexceptions taken in the LAR.The NMPNS submittal isbased on Revision 7 ofNEDC-33075P-A.Since the February 27, 2013meeting, the NRC approvedRevision 7 of NEDC-33075P-AJustification provided inSection 2.4.3Emergency Core Cooling System NPSH Information provided inThe NMP2 does not take credit for Containment Accident Section 4.2.6Pressure (CAP) to assure adequate net positive suction head(NPSH). In response to NRC staff, the licensee stated that a re-analysis of CAP is not required as a result of MELLLA+. Basedon feedback from the NRC staff, the NMP2 MELLLA+ submittalwill reference the NMP2 Extended Power Uprate (EPU)submittal Requests for Additional Information (RAr's) related toCAP and describe that the NPSH margins in the NMP2 EPUresponses remain bounding for MELLLA+.DSS-CD Implementation Information provided inImplementation of DSS-CD Stability Solution in Place of Option Section 2.4.1III. The NMP2 MELLLA+ submittal will address theimplementation strategy for DSS-CD, including the need formonitoring the timing for arming the protection associated withDSS-CD and the Oscillation Power Range Monitor (OPRM) dataanalysis already completed.21 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGETable 3 -Topics Discussed During NRC Pre-Meeting on February 27,2013Topic as Summarized in NRC Meeting SummaryIssued on March 13, 2013 (Reference 13)MELLLA+ SARReference (Attachments 8and 10 of this Enclosure)TRACG ATWS with Core Instability (ATWSI)The NMP2 submittal will include anticipated transients withoutSCRAM with instability (ATWSI) sensitivity analysis resultsusing a modified T-min correlation similar to what GeneralElectric Hitachi Nuclear Energy (GEH) provided in response toanother licensee's RAI. Additional information on the model wasrequested if and when it becomes available. However, GEH notedthat there is no additional testing at this time.Information provided inSection 9.3.3Operator TrainingProvide the implementation plan outlining the simulator upgradeand operator training plan to support implementation of the LAR.Information provided inSection 10.6NMPNS has requested thatthe NRC approve this LARby October 2014. Tosupport this schedule,NMPNS plans to upgradethe simulator by the secondquarter of 2014 to supportoperator training in thesecond and third quarters of2014.Reference Core versus Actual Cycle Specific Core See Notes I through 3Cycle Specific Core Design and Associated Safety Analyses, andReload Analysis using PRIME Code. The NMP2 submittal will Information provided indescribe the potential differences in the analytical inputs and Sections 2.1, 2.2, and 2.6.3results between the reference core and the actual reload analysis and Footnote 4 of Appendixthat will be submitted as a supplement to the MELLLA+ Asubmittal.GESTR-M versus PRIME Following the NRCSubsequent to the meeting the NRC staff noted that the licensee's discussions, the MELLLA+presentation stated that the licensee's LAR submission is going to SAR was revised to utilizeinclude the analyses based in GESTR-M Code and it is planning PRIME Thermal-to supplement its LAR with the Analyses based on PRIME Code, Mechanical (T-M)The LAR submission based on GESTR-M Code would not be methodology. In addition,"acceptable," This staff concern has been communicated to the PRIME fuel parameterslicensee on March 12, 2013. have been used in theanalyses requiring fuelIn an email dated March 12, 2013, the NRC staff noted that a performance parameters.LAR submission based on GESTR-M Code would not be"acceptable". A follow-up meeting with the NRC was conducted Information provided inon March 29, 2013. Table 1-1, Sections 2.6.3and 4.322 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGETable 3 -Topics Discussed During NRC Pre-Meeting on February 27, 2013Topic as Summarized in NRC Meeting Summary MELLLA+ SARIssued on March 13, 2013 (Reference 13) Reference (Attachments 8and 10 of this Enclosure)Notes:1. The fuel and cycle-dependent analyses, including the plant-specific thermal limitsassessment, will be submitted for NRC staff confirmation by supplementing the initialMELLLA+ Safety Analysis Report (SAR) in accordance with Limitation and Condition12.4 of the MELLLA+ Licensing Topical Report (LTR) Safety Evaluation Report (SER).Specifically, CENG will provide the cycle specific Supplemental Reload Licensing Report(SRLR) and Fuel Bundle Information Report (FBIR), which includes the supplementalinformation to satisfy MELLLA+ LTR SER Limitation and Condition 12.4. CENG willsubmit this information by February 28, 2014.2. Nine Mile Point Nuclear Station, LLC (NMPNS) will provide a cycle-specific core designloading map along with a summary of differences between the reference design described inthe M+SAR and the cycle-specific core design. This summary will include differences inthe energy requirements, average enrichment, and analytical inputs, a cycle-specific thermallimits assessment, and the actual reload analysis results. Additionally, the SupplementalReload Licensing Report, which includes the cycle specific core map, will be provided.Submittal of the cycle-specific design will satisfy the NRC request made at the MELLLA+LAR pre-meeting on March 13, 2013.3. The NMP2 Cycle 15 specific reload analysis will utilize TRACG rather than ODYN forAOO. Section 9.1.1 of the MELLLA+ SAR (Attachments 8 and 10 of this Enclosure) states:"In the event that the cycle-specific reload analysis is based on TRACG rather ODYN forAOO, no 0.01 added to the OLMCPR is required."

4.0 REGULATORY EVALUATION

4.1 Evaluation of NMIP2 License Amendment Requests to Establish that They Are Not Linked4.1.1 Guidance from NRR Office Instruction LIC-109, "Acceptance Review Procedures"Revision 1 of Office of Nuclear Reactor Regulation (NRR) Office Instruction LIC-109,"Acceptance Review Procedures," (Reference 14) provides the NRR staff (and other NRC staffsupporting NRR licensing activities) a basic framework for performing an acceptance reviewupon receipt of a Requested Licensing Action (RLA) from a licensee. It defines that the NRCshould not accept for NRC review and approval an RLA that is linked to another RLA.Section 1.3.2 of LIC-109 states linked RLAs "are RLAs, where approval of one RLA iscontingent upon the approval of (an) other RLA(s) currently under review. This definitionevaluates the independence of an RLA with respect to all other RLAs currently under review."Section 3.1.1 of LIC-109 states: "Linked RLAs: Determine whether the approval of the RLA iscontingent upon the approval of other RLAs currently under review. It is important to note thatmultiple RLAs can affect the same systems or Technical Specifications (TSs) without being23 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGElinked. As such, it may be possible to issue them in any order and without regard to the results ofthe review of the others. An RLA should not be accepted for NRC review and approval until allprerequisite RLAs have been reviewed and approved by the NRC."In addition, Example 3 of LIC-109 provides the following example of linked RLAs."While the NRC staff is reviewing a licensee's request to change the accident analyses for aloss-of-coolant accident (LOCA), the licensee submits an application for an extended poweruprate (EPU). The analysis and supporting justification for the EPU are based, in part, on theproposed LOCA analysis currently under review."LIC-109 states that this example is not acceptable, "because the EPU should not begin until allprerequisite reviews have been completed (Linked RLAs). Additionally, the regulatory basis citedin the EPU application (i.e., the currently unapproved LOCA analysis) is not the current licensingbasis for the plant (Regulatory Basis)." It stated that it may be acceptable for review if "reviewand approval of the EPU was not contingent upon the outcome of the NRC Staff's review of theLOCA analysis."4.1.2 Evaluation of NMP2 RCS Pressure -Temperature and MELLLA+ License AmendmentRequestsOn November 21, 2012, NMP2 submitted a License Amendment Request (LAR) to create a newTS section that is numbered TS 5.6.7 for the Reactor Coolant System Pressure and TemperatureLimits Report (Reference 5). In the MELLLA+ LAR, a new TS is added that is number TS 5.6.8.The numbering of TS 5.6.8 is an administrative consideration. The MELLLA+ LAR isindependent of the LAR submitted on November 21, 2012. NRC approval or rejection ofReference 5 would have no technical impact on the MELLLA+ LAR.li *

  • 1"1i. #.a 11vaiuauon Of rtiir aLO one IVILLLA=Jk- Ueense~ Amendment KquestsOn july 5, 2013, NMPNS submitted a request to amend the NMP2 Renewed Operating License(OL) NPF 69 to incr-ease the isotopie cnrichmcnt of boron 10 in the sedium pentaborae selutlused to p..p..e the n.ut..n absorbcr solutioen in the SLS 15). This request in:ludes thesupporting ehanges to the NMP2 Teehnical Speeification (TS) 3.1.7., "Standby Liquid Control(SLC) System," to increase the boron 10 isotopic enriehmifent in the sodium pentaborate solutionutilized in the SLC System and to deer-ease the SLG System tank Yolume.The SLS LAR and the MELLLA+ LAR both affeet NMP2 TS 3.1.7, including the same changesto SR 3.1.7.10 and Figure 3.1.7 1 to inerease the isotopic enrichment of boroen 10 int the sodiumpcntaberate solution and the associated change in the SLS Tank Minimum volume. Section 3.1.1of LIC 109 establishes that multiple RLAs can affect the same systems or. TSs without binAs statedi in Seeiont 1.3. of LI 11 9, llllnked7V~li~iI RLA "af Rb]s where aproa R Iscntingent upon the approval of (an) other- RLA(s) curffenly under- review.'of boron 10 in the sodium pentaboa.e solution and reduce the SLS Tank Minimum Yolumerequirements, so that they eould be implemented during the spring r-efifeling outage in 2014 forNMP2. These changes will be justified utilizing the current licensing basis, and ae no24 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEdependent efn the analysis that will be submited in the MELLLA-- LAR. The NRC an reviewand approve the SLS LAR witheut referenec te the MELLLA+ LRThe MELLLA+- LAR propeses changes to the NMP2 T-8, inceluding ehanges to NMP2 T-S31.that would inrease the SLS pump discharge pressure aceptance ..-iterc-n, incrIease the isotpicenrichment of boroen 10 in the sodium pentaborate solution, and r-educe the SL's Tank Minimumifvolume requirements. These changes will be justified utilizing analyses that are m e"d inthe MELL.LA+- LAR, inceluding a MELLLA+i specifie boron equivalency analysis and AT-W8antalysis. Given that the justification for- thesc changes will be proevided int the MELLLA+ LAR,the NRCG can review and approvye the MELLLA+ LAR without refer-ence to the SLS LAR.if the SLS LAR is approved by the NRC and implemented prior to the NRC approyal of theMELLLA+ LAR, the onily impact to the hMELLLA+ LAR would be to r-emove the proeposedchanges to SR 3.1.7.10 and F-igure 3.1.7 1. The antalyses proevided int the MELLLA+ LAR justivf'that those values arc approeprafte for- operatfion in the MELLL6A+ domfain. Thus, those analysesremain valid, and NRC r-evie ireued to justif' operation int the MELLLA+- dom-alin with41those SLS porametefs.if the SLS LAR is not approeved by the NRC, this action would have no impact on the NRC-r-eview and approval of the MELLLA+- LAR.if the MELLLA+ L6AR is approved by the NRC and implemented pr~ior to the NRC approvl othe SLS LAR, then the SLS LAR would be retracted by NMPNS because the SLS [AR does notaddress oper-ation in the MELLLA+- oper-ating domain and the applicable changes wouldbif the N4BLLLA+/-- LAR is not approved by the NRC, this action would have no impact en theKlt ADv~ alit a1,1.,' .., u Q1t Cs tin L -t&J NI lllU I+! lill-FF4..44.1.3 NMP2 ConclusionGiven the above, the RCS Pressure -Temperature LAR (Reference 5), the SLS LAR (Refer-ence4--5) and the MELLLA+ LARs are separate and independent licensing actions that the NRC canreview and approve independently. Thus, they are not linked RLAs as defined in LIC- 109.4.2 Applicable Regulatory Requirements/Criteria4.2.1 MELLLA+ and DSS-CD10 CFR 50.4610 CFR 50.46(a)(1)(i) states: "Each boiling or pressurized light-water nuclear power reactorfueled with uranium oxide pellets within cylindrical zircaloy or ZIRLO cladding must beprovided with an emergency core cooling system (ECCS) that must be designed so that itscalculated cooling performance following postulated loss-of-coolant accidents conforms to thecriteria set forth in paragraph (b) of this section..."25 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEThe acceptance criteria of 10 CFR 50.46(b) are:"(1) Peak cladding temperature. The calculated maximum fuel element cladding temperatureshall not exceed 22000 F."(2) Maximum cladding oxidation. The calculated total oxidation of the cladding shallnowhere exceed 0.17 times the total cladding thickness before oxidation..."(3) Maximum hydrogen generation. The calculated total amount of hydrogen generated fromthe chemical reaction of the cladding with water or steam shall not exceed 0.01 times thehypothetical amount that would be generated if all of the metal in the cladding cylinderssurrounding the fuel, excluding the cladding surrounding the plenum volume, were to react."(4) Coolable geometry. Calculated changes in core geometry shall be such that the coreremains amenable to cooling."(5) Long-term cooling. After any calculated successful initial operation of the ECCS, thecalculated core temperature shall be maintained at an acceptably low value and decay heatshall be removed for the extended period of time required by the long-lived radioactivityremaining in the core."Section 4.3 of the MELLLA+ SAR demonstrates that the requirements established in 10 CFR50.46(a)((1)(i) and the acceptance criteria of 10 CFR 50.46(b)(1) through (5) will be met duringoperation of NMP2 in the MELLLA+ operating domain.Appendix A to 10 CFR 50, General Design Criteria10 CFR 50.36(c)(2)(ii), Criterion 2 requires that TS limiting conditions for operation includeprocess variables, design features, and operating restrictions that are initial conditions of designbasis accident analysis. Compliance with the TS ensures that the NMP2 system performanceparameters are maintained within the values assumed in the safety analyses. The TS changes aresupported by the safety analyses and continue to provide a level of protection comparable to thecurrent TS. Applicable regulatory requirements and significant safety evaluations performed insupport of the proposed changes are described in Attachments 8 and 10 of this Enclosure.Information Notices 2009-23 and 2011-21NRC Information Notice 2009-23, "Nuclear Fuel Thermal Conductivity Degradation," notifiedlicensees that analyses performed using pre-1999 methods may be less conservative thanpreviously understood (References 16 and 17). In addition, NRC Information Notice 2011-21,"Realistic Emergency Core Cooling System Evaluation Model Effects Resulting from NuclearFuel Thermal Conductivity Degradation," notifies addresses that the impact of irradiation on fuelthermal conductivity has the potential to cause errors in ECCS evaluation models, specifically ahigher peak cladding temperature (Reference 18).This issue does not apply to this submittal, because the MELLLA+ SAR utilized to justifyoperation in the MELLLA+ operating domain includes PRIME T-M methodology as discussed inSection 2.6.3 of Attachments 8 and 10 of this Enclosure.26 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE4.2.2 Standby Liquid Control System Isotopic Enrichment of Boron 1Appendix A to 10 CFR 50, General Design CriteriaGeneral Design Criterion (GDC) 26, "Reactivity control system redundancy and capability,"states:"Two independent reactivity control systems of different design principles shall be provided.One of the systems shall use control rods, preferably including a positive means for insertingthe rods, and shall be capable of reliably controlling reactivity changes to assure that underconditions of normal operation, including anticipated operational occurrences, and withappropriate margin for malfunctions such as stuck rods, specified acceptable fuel designlimits are not exceeded. The second reactivity control system shall be capable of reliablycontrolling the rate of reactivity changes resulting from planned, normal power changes(including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of thesystems shall be capable of holding the reactor core subcritical under cold conditions."For BWRs, the provisions of 10 CFR 50.62 require that the second reactivity control system bethe SLS. Its function is, per the requirements, to inject into the reactor pressure vessel a boratedwater solution at a prescribed flow rate, concentration and boron-10 isotopic enrichment. Theboron in the solution absorbs neutrons, thus providing reactivity control to shut down the reactorin the event the control rods fail to insert into the core.GDC 27, "Combined reactivity control systems capability," states:"The reactivity control system shall be designed to have a combined capability, inconjunction with poison addition by the emergency core cooling system, of reliablycontrolling reactivity changes to assure that under postulated accident conditions and withappropriate margin for stuck rods the capability to cool the core is maintained.The SLS is the poison addition system described in GDC 27.10 CFR 50.62, "Requirements for reduction of risk from anticipated transients withoutscram (ATWS) events for light-water-cooled nuclear power plants"10 CFR 50.62 (c)(4) states:"Each boiling water reactor must have a standby liquid control system (SLCS) with thecapability of injecting, into the reactor pressure vessel a borated water solution at such a flowrate, level of boron concentration, and boron-10 isotope enrichment, and accounting forreactor pressure vessel volume, that the resulting reactivity control is at least equivalent tothat resulting from injection of 86 gallons per minute of 13 weight percent sodiumpentaborate decahydrate solution at the natural boron-10 isotope abundance into a 251-inchinside diameter reactor pressure vessel for a given core design. The SLCS and its injectionlocation must be designed to perform its function in a reliable manner..."In the NRC-approved licensing topical report, NEDE-31096P-A, "Anticipated Transients WithoutScram: Response to NRC ATWS Rule, 10 CFR 50.62," General Electric provides guidance onmodifications to the SLC system to ensure licensee compliance with the ATWS rule. The NRCapproved the methods presented in NEDE-31096P-A for use by Boiling Water Reactor licensees27 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEto demonstrate compliance with the ATWS Rule. The application of this guidance demonstratesthat the equivalency requirement of 10 CFR 50.62 is met.10 CFR 50.67, "Accident source term"10 CFR 50.67.b(1) provides guidance to licensees with respect to revision of the licensee'scurrent accident source term in design basis radiological consequence analyses. Specifically, theregulation states that in order to revise the accident source term, a licensee shall apply for alicense amendment under 10 CFR 50.90 and that the application shall contain an evaluation of theconsequences of applicable design basis accidents previously analyzed in the safety analysisreport.The radiological consequences of certain design basis accidents (DBA) have been reevaluatedusing a full implementation of an Alternate Source Term as described in Regulatory Guide (RG)1.183 (Reference 19) and NRC Standard Review Plan (SRP) 15.0.1 (Reference 20). Theevaluation was performed at 120 percent of the original licensed power to bound the effects offuture power uprates. The evaluation demonstrates that the calculated offsite exposures andcontrol room doses meet the criteria of 10 CFR 50.67.The supporting analyses for Alternate Source Tenn assume the pH of the suppression pool iscontrolled to prevent the re-evolution of iodine following a DBA LOCA. This is accomplished byinjecting the SLS solution (i.e., boron solution) following a DBA LOCA to ensure pH iscontrolled to a value greater than 7.0. Analysis has confirmed that the SLS will continue tomaintain suppression pool pH level above 7.0 following a LOCA which involves significantfission product releases.Information Notice 2001-13In response to potential non-conservatisms in pressure calculations related to SLS dischargepressure during ATWS scenarios, the NRC issued Information Notice (IN) 2001-13(Reference 21). IN 2001-13 requested licensees to evaluate relief valve pressure margins on theSLS and confirm to the NRC that the systems remained in compliance with NRC regulations.NMPNS determined that the concerns identified in IN 2001-13 were applicable to NMP2.This LAR is proposing to revise NMP2 TS SR 3.1.7.7 by increasing the minimum requiredNMP2 SLS pump test discharge pressure from 1,327 psig to 1,335 psig, while maintainingadequate margin for relief valve lift.4.2.3 Safety Limit Minimum Critical Power Ratio10 CFR 50.3610 CFR 50.36(c)(1), requires that power reactor facility TS include safety limits for processvariables that protect the integrity of certain physical barriers that guard against the uncontrolledrelease of radioactivity. The fuel cladding is one of the physical barriers that separate theradioactive materials from the environment. The purpose of the SLMCPR is to ensure thatspecified acceptable fuel design limits (SAFDLs) are not exceeded during steady state operationand analyzed transients. The integrity of this cladding barrier is related to its relative freedomfrom perforations or cracking. Fuel cladding perforations can result from thermal stresses, whichcan occur from reactor operation significantly above design conditions. Since the parameters that28 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEresult in fuel damage are not directly observable during reactor operation, the thermal andhydraulic conditions that result in the onset of transition boiling have been used to mark thebeginning of the region in which fuel cladding damage could occur.The GNF analysis presented in Attachments 9 and 11 of this Enclosure established that NMP2continues to meet the requirement of 10 CFR 50.36(c)(1) with the increased acceptance criteriafor the SLMCPR for two recirculation loops in operation.Appendix A to 10 CFR 50, General Design Criteria10 CFR 50.36(c)(2)(ii), Criterion 10 requires:"The reactor core and associated coolant, control, and protection systems shall be designedwith appropriate margin to assure that specified acceptable fuel design limits are notexceeded during any condition of normal operation, including the effects of anticipatedoperational occurrences."The fuel cladding must not sustain damage as a result of normal operation and abnormaloperational transients. The reactor core safety limits are established to preclude violation of thefuel design criterion such that at least 99.9% of the fuel rods in the core would not be expected toexperience the onset of transition boiling.The GNF analysis presented in Attachments 9 and 11 of this Enclosure established that NMP2continues to meet 10 CFR 50, Appendix A, Criterion 10 with the increased acceptance criteria forthe SLMCPR for two recirculation loops in operation.4.3 Precedent4.3.1 MELLLA+ and DSS-CDThis license amendment request is based on approved GEH license topical reports (References 1through 4) and their associated Safety Evaluation Reports. The NMP2 application follows themethodologies and limitations of those LTRs and their respective SERs. On January 21, 2010,Monticllo Nuelear Generating Plant submitted a liccmse amendment request to adoptexpanded MELLLA+- operating domain; this license amendment request remains under review*bthe NRC. (ADAMS I A ession No. ML 100280558),The NRC approved a similar request on March 28, 2014 for the Monticello Nuclear GeneratorPlant to permit operating in the MELLLA+ operating domain (ADAMS AccessionNo. ML14035A248 (Reference 22).4.3.2 Standby Liquid Contro! System 19otepic Enriehment of Boren 10The NRC has approved a number- of r-equests to inerease the isotopic enrichment of boroen10ithe sedium pentaborate utilized to prepar-e the solutiont that is utilized in the SLS. These incelude*Columbia Gcnerating Station issuance of Amendment Re: Increased Boron Concentrationin Standby Liquid Control System (TAC No. ME4789), dated May 18, 2011, (ADAMAccession No. MWI 11170370) (Refcr-enc 23)29 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEThis amendment is similar- to the NMP2 proeposed change with r-espeet to the iner-ease in theIsetepie ber-en lenr-ilhment in the s.dium pentabor.e so.ution. utilized in the SLS. ForColumbia Generating Station, the bero.n 10 enrichment in.r.ease was from 22 at.m per.ent t4 .;em pe..eefit, .Qta1lJ~f~l -h --JJ C+-- Q+tI I f-+ L1ftt~t -JLL IUA S.I 33~lI -V A -I~lA -IZ Dfl ClS .AvLiquid Conitroel System (TAC Nos. MDI 424 and MD)1425), dated Februar-Y 28, 2007-,(ADAMS8 Accession No. ML0703902 15) (Reference 2-4)This amendment is Sim.. ilr to the NM2 proepesed ehage with r.espe. t to the in.rease in th,.... berf -efr-nriehitment in the sediuA p.ntabor-at÷ solution utilized in the SL'S and theSLS volum, e de.rease. The am.endment alsc rIedu.ed the sodium pentab.fate ho w eve .. ... ....r-r .....S is net.. a e to the. ...... .. ....' ........... .........ifieconcentr-ation.4.3.3 Safety Limit Minimum Critical Power RatioThe NRC has approved a number of requests to increase the SLMCPR for two recirculation loopsin operation that utilized GNF analysis to support the change. These include:" LaSalle County Station, Unit 2 -Issuance of Amendment No. 192 Regarding TechnicalSpecification Change For Safety Limit Minimum Critical Power Ratio (TAC No. ME9769),dated February 27, 2013 (ADAMS Accession No. ML 13050A637) (Reference 25)* Cooper Nuclear Station -Issuance of Amendment Re: Revision of Technical Specifications -Safety Limit Minimum Critical Power Ratio (TAC No. ME8853), dated November 9, 2012(ADAMS Accession No. ML12299A092) (Reference 26)4.4 Significant Hazards ConsiderationNine Mile Point Nuclear Station LLC (NMPNS) is requesting an amendment to Renewed FacilityOperating License NPF-69 for Nine Mile Point Unit 2 (NMP2). The proposed amendmentincludes supporting changes to the NMP2 Technical Specifications (TSs) necessary to: 1)implement the Maximum Extended Load Line Limit Analysis Plus (MELLLA+) expandedoperating domain; 2) change the stability solution to the Detect and Suppress Solution -Confirmation Density (DSS-CD); 3) use the TRACG04 analysis code; and 4) ietIease-this.topic enr-i.hment of boron 10 in the Standby Liquid Contol System (SLS); and 5) increase theSafety Limit Minimum Critical Power Ratio (SLMCPR) for two recirculation loops in operation.The proposed changes to the NMP2 TSs:The following is a list of the proposed changes to the NMP2 TSs:* Revise Safety Limit (SL) 2.1.1.2 by increasing the SLMCPR for two recirculation loops inoperation from > 1.07 to > 1.09* Revise the acceptance criterion in TS 3.1.7, "Standby Liquid Control (SLC) System,"Surveillance Requirement (SR) 3.1.7.7 by increasing the discharge pressure from> 1,327 pounds per square inch gauge (psig) to > 1,335 psig.Revise the aeceptance cr.iterion in TS SR 3.1.7.10 by increasing the sodium pentab.- , ateboron 10 enrichment r-equir-ement from ! 25 atom perceent to ý! 92 atem perceent, and make-a30 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGEeCorsponding ehange in TS Figure 3.1.7 1, "Sodium PFR cntorPtet SyltienVolum(/P)nsentation R"quircmcnts"*Revise TS Figure 3.1.7 1 to accouint for- the deer-ease in the minimuem volumoe of the SLS tan~kfrhm 1,558.6 gallons and 1,288 gallons at sodium pentaborat concentrations of 130/6 and14.4%a, ospedtively, to 1,600 gallons and 1,530 galTSns at sodium p3ntaborate con1entrationsof 13.6%r and 14.4ti, rrespeRtivel*Change the Required Actions for Condition F of TS 3.3.1.1, "Reactor Protection System(RPS) Instrumentation"* Change Condition G of TS 3.3. 1.1Add new Conditions lRand K to TS 3.3. 1.1" Correct an editorial error in Note 3 to TS SR 3.3.1.1.13 (i.e., "ORRM" is changed to"6OPRM"5)* Eliminate TS SR 3.3.1.1.16 and references to it in TS Table 3.3.1.1-1, "Reactor ProtectionSystem Instrumentation"" Change the allowable value (AV) for TS Table 3.3.1.1-1, Function 2.b, Average PowerRange Monitor (APRM) -Flow Biased Simulated Thermal Power (STP) -Upscale from"<5 0.55W +60.5% [Rated Thermal Power] RTP and < 115.5% RTP" to"< 0.61W + 63.4% RTP and < 115.5% RTP"" Add a new note to TS Table 3.3.1.1-1, Function 2.b that requires the Flow Biased SimulatedThermal Power -Upscale scram setpoint to be reset to the values defined by the CoreOperating Limits Report (COLR) to implement the Automated Backup Stability Protection(BSP) Scram Region in accordance with Required Action F--24F.2 of TS 3.3.1.1* Add a new note to TS Table 3.3.1.1-1, Function 2.e, Oscillation Power Range Monitor(OPRM) -Upscale to denote that following implementation of DSS-CD, DSS-CD is notrequired to be armed while in the DSS-CD Armed Region during the first reactor startup andduring the first controlled shutdown that passes completely through the DSS-CD ArmedRegion. However, DSS-CD is considered operable and capable of automatically arming foroperation at recirculation drive flow rates above the DSS-CD Armed Region* Change the mode of applicability for TS Table 3.3.1. 1-1, Function 2.e, OPRM-Upscale fromMode 1 to_> 18% RTP* Change the allowable value for TS Table 3.3.1.1-1, Function 2.e from "As specified in theCOLR" to "NA", Add a prohibition tModify TS Limiting Condition for Operation (LCO) 3.4.1,"Recirculation Loops Operating," the- peibitsto prohibit operation in the MaximumExtended Load Line Limit Analysis (MELLLA) domain or MELLLA+ expanded operatingdomain as defined in the COLR when in operation with a single recirculation loop* Add Required Action B.2 to TS 3.4.1 to identify that intentional operation in the MELLLAdomain or MELLLA+ domain as defined in the COLR is prohibited when a recirculationloop is declared "not in operation" due to a recirculation loop flow mismatch not within limits" Revise TS 5.6.5.a.4 to replace "Reactor Protection System Instrumentation Setpoint for theOPRM -Upscale Function Allowable Value for Specification 3.3.1.1" with "The ManualBackup Stability Protection (BSP) Scram Region (Region I), the Manual BSP ControlledEntry Region (Region II), the modified APRM Simulated Thermal Power -High setpointsused in the OPRM (Function 2.e), Automated BSP Scram Region, and the BSP Boundary forSpecification 3.3.1.1"* Add TS 5.6.8, "OPRM Report," to define the contents of the report required by new RequiredAction R.-F.__3 of TS 3.3.1.131 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGENMPNS has evaluated whether or not a significant hazards consideration is involved with theproposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance ofAmendment," as discussed below:1) Will the change involve a significant increase in the probability or consequences of anaccident previously evaluated?Response: No.The probability (frequency of occurrence) of Design Basis Accidents occurring is not affected byimplementing the MELLLA+ operating domain and DSS-CD stability solution, because NMP2continues to comply with the regulatory and design basis criteria established for plant equipment.A SLS failure is not a precursor of any previously evaluated accident in the NMP2 USAR. Theincrease to the SLMCPR for two recirculation loops in operation does not increase the probabilityof an evaluated accident. Consequently there is no change in the probability of an accidentpreviously evaluated accident.The spectrum of postulated transients was investigated and shown to remain within the NRCapproved acceptance limits. Fuel integrity is maintained by meeting existing design andregulatory limits. Further, a probabilistic risk assessment demonstrates that the calculated coredamage frequency and the large early release frequency do not significantly change due tooperation in the MELLLA+ domain.Challenges to the reactor coolant pressure boundary were evaluated for the MELLLA+ operatingdomain conditions (pressure, temperature, flow, and radiation) and were found to meet theiracceptance criteria for allowable stresses and overpressure margin.Challenges to the containment were evaluated and the containment and its associated coolingsystems continue to meet the current licensing basis. The calculated post LOCA suppression pooltemperature remains acceptable.The SLS is used to mitigate the consequences of an Anticipated Transient Without SCRAM(ATWS) special event and is used to limit the radiological dose during a Loss of CoolantAccident (LOCA). The proposed changes do not affect the capability of the SLS to perform thesetwo functions in accordance with the assumptions of the associated analyses. The ATWSevaluation with the proposed changes incorporated demonstrated that all the ATWS acceptancecriteria are met. The ability of the SLS to mitigate radiological dose in the event of a LOCA bymaintaining suppression pool pH > 7.0 is not affected by these changes.This proposed change to the SLMCPR for two recirculation loops in operation does not result inany modification to the design or operation of the systems that are used in mitigation of accidents.Limits have been established, consistent with NRC approved methods, to ensure that fuelperformance during normal, transient, and accident conditions is acceptable. The proposedchange to the SLMCPR for two recirculation loops in operation continues to conservativelyestablish this safety limit such that the fuel is protected during normal operation and during anyplant transients or anticipated operational occurrences.Therefore, the proposed change does not involve a significant increase in the probability orconsequences of an accident previously evaluated.32 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE2) Will the change create the possibility of a new or different kind of accident from anyaccident previously evaluated?Response: No.Equipment that could be affected by implementing the MELLLA+ operating domain andDSS-CD stability solution was evaluated. No new operating mode, safety-related equipmentlineup, accident scenario, or equipment failure mode was identified. The full spectrum of accidentconsiderations was evaluated and no new or different kind of accident was identified. TheMELLLA+ operating domain and DSS-CD stability solution use developed technology and applyit within the capabilities of existing plant safety-related equipment in accordance with theregulatory criteria (including NRC approved codes, standards and methods). No new accident orevent precursor was identified.The long-term stability solution is being changed from the currently approved Option III solutionto DSS-CD. DSS-CD is designed to identify the power oscillation upon inception and initiatecontrol rod insertion (scram) to terminate the oscillations prior to any significant amplitudegrowth exceeding the applicable safety limits. DSS-CD is based on the same hardware design asOption III. However, it introduces an enhanced detection algorithm that detects the inception ofpower oscillations and generates an earlier power suppression trip signal .."lusively based .u..e...... p .ei.d .nfirmation rec.gnition. The existing Option III algorithms are retained (withgeneric setpoints) to provide defense-in-depth protection for unanticipated reactor instabilityevents.Structures, systems and components (SSCs) previously required for the mitigation of a transientremain capable of fulfilling their intended design functions. The proposed changes do notadversely affect safety-related systems or components and do not challenge the performance orintegrity of any safety-related system. The physical changes to the SLS a&e-islimited to theincrcase in the boron 10 enrichment of the sodium pentaboratc solution int the SLS8 storage tank,the correspending deerease in the net sodium pentabefmte selution volume roequir-ement in the SLSSstorag tank,- the increase in the SLS pump discharge pressure acceptance criterion-aii4,4heass.ciated inst:um.ntation .hangcs. The proposed changes do not otherwise affect the design oroperation of the SLS.This proposed change to the SLMCPR for two recirculation loops in operation does not result inany modification to the design or operation of the systems that are used in the mitigation ofaccidents. The proposed change to the SLMCPR for two recirculation loops in operation assuresthat safety criteria are maintained.The proposed changes do not adversely affect any current system interfaces or create any newinterfaces that could result in an accident or malfunction of a different kind than was previouslyevaluated.Therefore, the proposed changes do not create the possibility of a new or different kind ofaccident from any accident previously evaluated.33 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE3) Will the change involve a significant reduction in a margin of safety?Response: No.The MELLLA+ operating domain affects only design and operational margins. Challenges to thefuel, reactor coolant pressure boundary, and containment were evaluated for the MELLLA+operating domain conditions. Fuel integrity is maintained by meeting existing design andregulatory limits. The calculated loads on affected SSCs, including the reactor coolant pressureboundary, will remain within their design specifications for design basis event categories. NoNRC acceptance criterion is exceeded.Comprehensive analyses of the proposed changes have concluded that relevant design and safetyacceptance criteria will be met without a significant reduction in margins of safety. The analyseshave demonstrated that the NMP2 SSCs are capable of safely performing at MELLLA+conditions. The analyses identified and defined the major input parameters to the Nuclear SteamSupply System (NSSS), analyzed NSSS design transients, and evaluated the capabilities of theNSSS fluid systems, NSSS/Balance of Plant (BOP) interfaces, NSSS control systems, and NSSSand BOP components, as appropriate. Radiological consequences of design basis events remainwithin regulatory limits and are not increased significantly. The analyses confirmed that NSSSand BOP SSCs are capable of achieving MELLLA+ conditions without significant reduction inmargins of safety.Analyses have shown that the integrity of primary fission product barriers will not besignificantly affected as a result of change in the operating domain. Calculated loads on SSCsimportant to safety have been shown to remain within design allowables with MELLLA+conditions for all design basis event categories. Plant response to transients and accidents do notresult in exceeding acceptance criteria. As appropriate, the evaluations that demonstrateacceptability of MELLLA+ have been performed using methods that have either been reviewedand approved by the NRC staff, or that are in compliance with regulatory review guidance andstandards established for maintaining adequate margins of safety. These evaluations demonstratethat there are no significant reductions in the margins of safety.The SLS is used to mitigate the consequences of an ATWS event and is used to limit theradiological dose during a LOCA. The proposed changes do not affect the capability of the SLSto perform these two functions in accordance with the assumptions of the associated analyses.The ATWS evaluation with the proposed changes incorporated demonstrated that all the ATWSacceptance criteria are met. The ability of the SLS to mitigate radiological dose in the event of aLOCA by maintaining suppression pool pH > 7.0 is not affected by these changes.This proposed change to the SLMCPR for two recirculation loops in operation provides a marginof safety by ensuring that no more than 0.1% of fuel rods are expected to be in boiling transitionif the MCPR limit is not violated. The proposed change will ensure the appropriate level of fuelprotection is maintained. Additionally, operational limits are established based on the proposedSLMCPR to ensure that the SLMCPR is not violated during all modes of operation. This willensure that the fuel design safety criteria are met (i.e., that at least 99.9% of the fuel rods do notexperience transition boiling during normal operation as well as anticipated operationaloccurrences).Therefore, the proposed changes do not involve a significant reduction in a margin of safety.34 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE4.5 ConclusionsBased on the considerations discussed above, (1) there is reasonable assurance that the health andsafety of the public will not be endangered by operation in the proposed manner, (2) suchactivities will be conducted in compliance with the Commission's regulations, and (3) theissuance of the amendment will not be inimical to the common defense and security or to thehealth and safety of the public. Therefore, NMPNS concludes that the proposed amendmentpresents no significant hazards considerations under the standards set forth in 10 CFR 50.92, and,accordingly, a finding of "no significant hazards consideration" is justified.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respectto installation or use of a facility component located within the restricted area, as defined in10 CFR 20, or would change an inspection or surveillance requirement. However, the proposedamendment does not involve: (i) a significant hazards consideration; (ii) a significant change inthe types or significant increase in the amounts of any effluent that may be released offsite; or(iii) a significant increase in individual or cumulative occupational radiation exposure.Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion setforth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impactstatement or environmental assessment need be prepared in connection with the proposedamendment.

6.0 REFERENCES

1. GE Hitachi Nuclear Energy, "General Electric Boiling Water Reactor Maximum ExtendedLoad Line Limit Analysis Plus Licensing Topical Report," NEDC-33006P-A, Revision 3,June 2009 and NEDO-33006-A, Revision 3, June 2009.2. GE Hitachi Nuclear Energy, "GE Hitachi Boiling Water Reactor, Detect And SuppressSolution -Confirmation Density," NEDC-33075P, Revision 7, June 2011; and Anthony J.Mendiola (NRC) to Jerald G. Head (GEH), "Revised Draft Safety Evaluation for GE-HitachiNuclear Energy Americas, LLC Topical Report NEDC-33075P, Revision 7, 'GE HitachiBoiling Water Reactor Detect and Suppress Solution -Confirmation Density' (TAC No.ME6577)," dated August 6, 2013.3. GE Hitachi Nuclear Energy, "DSS-CD TRACG Application," NEDE-33147P-A, Revision 4,August 2013.35 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE4. a. GE Hitachi Nuclear Energy, "Applicability of GE Methods to Expanded OperatingDomains," NEDC-33173P-A, Revision 4, November 2012.b. Letter from R. Kingston (GEH) to NRC, "Clarification of Stability Evaluations -NEDC-33173P," MFN 08-541, June 25, 2008.c. Letter from J. Harrison (GEH) to NRC, "Implementation of Methods Limitations -NEDC-33173," MFN 08-693, September 18, 2008.d. Letter from J. Harrison (GEH) to NRC, "NEDC-33173P -Implementation of Limitation12," MFN 09-143, February 27, 2009.e. GE Hitachi Nuclear Energy, "Implementation of PRIME Models and Data inDownstream Methods," NEDO-33173, Supplement 4-A, Revision 1, November 2012.5. Letter from K. Langdon (NMPNS) to the Document Control Desk (NRC), LicenseAmendment Request Pursuant to 10 CFR 50.90: Relocation of Pressure and TemperatureLimit Curves to the Pressure and Temperature Limits Report, dated November 21, 2012(ADAMS Accession Number ML123380336).6. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," U.S. NRC,Revision 2, May 2011.7. NEDE 31096 A, "Anticipated Transients Without Senm Respense to NRC ATWS Rl..CFR5..62," Fcbrua, y ..87..Not used.8. GE Hitachi Nuclear Energy, "General Electric Standard Application for Reactor Fuel,"NEDE-2401 1-P-A, and NEDE-2401 1-P-A-US, Revision 19 April 2012.9 NEDC-32601P-A, "Methodology and Uncertainties for Safety Limit MCPR Evaluations,"August 1999.10 NEDC-32694P-A, "Power Distribution Uncertainties for Safety Limit MCPR Evaluations"August 1999.11 NEDC-32505P-A, "R-Factor Calculation Method for GEl 1, GE12 and GE13 Fuel,"Revision 1, July 1999.12. Letter from (NRC) to K. Langdon (NMPNS), Nine Mile Point Nuclear Station, Unit No. 2 -Issuance of Amendment Re: Extended Power Uprate (TAC No. ME1476), dated December22, 2011 (ADAMS Accession Number ML1 133000041).13. B. Vaidya (NRC), Summary of February 27,_2013, Meeting with Nine Mile Point NuclearStation, Unit 2, to discuss Planned Amendment Request on Implementation of MELLLA+(TAC No. MF0587), dated March 13, 2013 (ADAMS Accession Number ML13059A374)14. NRR Office Instruction LIC-109, "Acceptance Review Procedures," Revision 1.36 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGE1 1; I at..... fxs 12A U..C [Muit !NWJC tgA DA~n4m~nt IRequest Pursuant te 10 CFR 50.90! Standby Liquid Control System Inerease in lsctepieEnrichment of Beroen 10, dated July 1C, 2013 (ADAMS Accessiont NumbefMb 1 31l9-7A224-l-)Not used.16. Information Notice 2009-23, "Nuclear Fuel Thermal Conductivity Degradation," U.S.Nuclear Regulatory Commission, dated October 8, 2009.17. Information Notice 2009-23, Supplement 1, "Nuclear Fuel Thermal ConductivityDegradation," U.S. Nuclear Regulatory Commission, dated October 26, 2012.18. Information Notice 2011-21, "Realistic Emergency Core Cooling System Evaluation ModelEffects Resulting from Nuclear Fuel Thermal Conductivity Degradation," U.S. NuclearRegulatory Commission, dated December 13, 2011.19. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating DesignBasis Accidents at Nuclear Power Reactors," July 2000 (ADAMS Accession No.ML003716792).20. Standard Review Plan (SRP) 15.0.1, "Radiological Consequence Analyses Using AlternativeSource Terms," Revision 0, July 2000 (ADAMS Accession No. ML003721661).21. Information Notice 2001-13, "Inadequate Standby Liquid Control System Relief ValveMargin," U.S. Nuclear Regulatory Commission, August 10, 2001.22. Letter from T. Beltz (NRC) to K. Fili (Monticello Nuclear Generating Plant), MoniticelloNuclear Generating Plant -Issuance of Amendment No. 180 to Renewed Facility OperatingLicense repardine Maximum Extended Load Line Limit Anavsis Plus (TAC No. ME3145),23.24.dated March 28, 2014 (ADAMS Accession No. ML14035A248).lcttcr from T. J. O'Connor(Monticcfll Nuclear- Gencrating Plant) to U. S. Nuelear Regulatory Commission, -1iensAmendment Request.: M.. imum Etended Load Line Limit Analysis Pluts," dated Ja"ualry,21, 2010 (ADAMS Aeecssiefn No. ML100280558).Lcter- from M. C. Thadani (NRC) te M. E. Reddemann (Energy Ncrthwest), "CcltimbiaGencr-ating Station issuanee of Amendment Re. iner-easd Ber-en ConeentrMien in StandbyLiquid Control System (TAC No. ME14789)," dated May 18, 2011 (ADAMS Aeeession No.Mbin 1 tl-703 70)7Not used.Letter- from R. V. Guizmani (NRC) to 13. T. McKintney (PPL Susquehanna, LLC),"Susquehanna Steam Ekcetric Staion, units I and 2 Issuance Cf Amnendment -c StandbLiquid Control System (TAG Nos. MD1424 and MD!425)," dated Febrdar-y 28, 200-7I A+.. +r.,A SC A .T L "-, k ... uT DIV % IV, D "T. ' .. kJ T'-- +. D l .+ ..I.- .I 'Aeees;Ste" e. r"bt0-t196Y1d:&.t:1"r.1NVt USU,25. Letter from N. DiFrancesco (NRC) to M. J. Pacillo (Exelon Nuclear), LaSalle CountyStation, Unit 2 -Issuance of Amendment No. 192 Regarding Technical Specification Changefor Safety Limit Minimum Critical Power Ratio (TAC No. ME9769), dated February27, 2013 (ADAMS Accession No. ML13050A637).26. Letter from L. E. Wilkins (NRC) to B. J. O'Grady (Nebraska Public Power District), CooperNuclear Station -Issuance of Amendment Re: Revision of Technical Specifications -Safety37 of 38 ENCLOSUREREVISION 1 -EVALUATION OF THE PROPOSED CHANGELimit Minimum Critical Power Ratio (TAC No. ME8853), dated November 9, 2012(ADAMS Accession No. ML12299A092).38 of 38 ATTACHMENT 1NINE MILE POINT UNIT 2PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS(MARK-UPS)The current version of the following Technical Specification (TS) page had been marked-up to reflect theproposed changes and replaces the corresponding page previously submitted on November 1, 2013(ADAMS Accession No. ML 13316B 107):3.1.7-3TS page 3.1.7-4 and associated TS Insert 1 -Figure 3.1.7-1 were implemented with approval of NMP2Amendment 143, and are to be removed from the original Attachment 1 submitted on November 1, 2013(ADAMS Accession No. ML 13316B 107).The remaining pages in Attachment 1, submitted on NovemberML13316B107) and modified by the submittal on May 14,ML14139A416), are not changed with this submittal.1, 2013 (ADAMS Accession No.2014 (ADAMS Accession No.Nine Mile Point Nuclear Station, LLCJune 13, 2014 SLC System3.1.7SURVEILLANCE REQUIREMENTS (continued)SURVEILLANCE FREQUENCYSR 3.1.7.7 Verify each pump develops a flow rate In accordance> 41.2 gpm at a discharge pressure with the> 4-3H psig. InserviceTestingProgramSR 3.1.7.8 Verify flow through one SLC subsystem 24 months on afrom pump into reactor pressure vessel. STAGGERED TESTBASISSR 3.1.7.9 Verify all heat traced piping between 24 monthsstorage tank and pump suction valve isunblocked. ANDOnce within24 hours afterpipingtemperature isrestored to> 70°FSR 3.1.7.10 Verify sodium pentaborate enrichment Prior tois > 92 atom percent B-10. addition toSLC tankINMP23.1.7-3Amendment W, 111, 117, 123, 140,44-37
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