April 18, 2014Mr. David A. HeacockPresident and Chief Nuclear OfficerDominion NuclearInnsbrook Technical Center5000 Dominion BoulevardGlen Allen, VA 23060-6711SUBJECT:MILLSTONE POWER STATION UNIT NO. 2 - ISSUANCE OF AMENDMENTRE: REVISE TECHNICAL SPECIFICATION 3/4.7.11 ULTIMATE HEAT SINK(TAC NO. MF1779)Dear Mr. Heacock:The Commission has issued the enclosed Amendment No. 318 to Renewed Facility OperatingLicense No. DPR-65 for the Millstone Power Station, Unit No. 2, in response to your applicationdated May 3, 2013, as supplemented by letters dated June 27, 2013, July 19, July 30, August 1,and October 2, 2013.The amendment would revise Technical Specification (TS) 3/4.7.11, “Ultimate Heat Sink”, toincrease the current ultimate heat sink water temperature limit from 75 °F to 80 °F and changethe TS Action to state, "With the ultimate heat sink water temperature greater than 80 °F, be inHOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours."A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included inthe Commission’s biweekly Federal Register notice.Sincerely,/RA/James Kim, Project ManagerPlant Licensing Branch 1-1Division of Operating Reactor LicensingOffice of Nuclear Reactor RegulationDocket No. 50-336Enclosures:1. Amendment No. 318 to DPR-652. Safety Evaluationcc w/encls: Distribution via ListservApril 18, 2014Mr. David A. HeacockPresident and Chief Nuclear OfficerDominion NuclearInnsbrook Technical Center5000 Dominion BoulevardGlen Allen, VA 23060-6711SUBJECT:MILLSTONE POWER STATION UNIT NO. 2 - ISSUANCE OF AMENDMENTRE: REVISE TECHNICAL SPECIFICATION 3/4.7.11 ULTIMATE HEAT SINK(TAC NO. MF1779)Dear Mr. Heacock:The Commission has issued the enclosed Amendment No. 318 to Renewed Facility OperatingLicense No. DPR-65 for the Millstone Power Station, Unit No. 2, in response to your applicationdated May 3, 2013, as supplemented by letters dated June 27, 2013, July 19, July 30, August 1,and October 2. 2013.The amendment would revise Technical Specification (TS) 3/4.7.11, “Ultimate Heat Sink”, toincrease the current ultimate heat sink water temperature limit from 75 °F to 80 °F and changethe TS Action to state, "With the ultimate heat sink water temperature greater than 80 °F, be inHOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours."A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included inthe Commission’s biweekly Federal Register notice.Sincerely,/RA/James Kim, Project ManagerPlant Licensing Branch 1-1Division of Operating Reactor LicensingOffice of Nuclear Reactor RegulationDocket No. 50-336Enclosures:1. Amendment No. 318 to DPR-652. Safety Evaluationcc w/encls: Distribution via ListservDISTRIBUTION:PUBLICRidsNrrLAKGoldsteinRidsNrrPMMillstoneRidsNrrDraAadb ResourceRidsOgcMailCenter ResourceRidsRgn1MailCenterRidsNrrDorlLpl1-1 ResourceRidsNrrDorlDpr ResourceRidsNrrDssSrxb ResourceRidsNrrDssSbpb ResourceRidsNrrDssStsb ResourceS. Jones, NRRR. Bellamy, RIRidsAcrsAcnw_MailCenter ResourceAccession No.: ML14037A408OFFICELPL1-1/PMLPL1-1/LANAMEJKimKGoldsteinDATE2/12/142/12/14OFFICESRXBB/BCSTSB/BCSBPB/BC*GCasto8/20/13AFPB/BC*See memo dated August 20, 2013**See memo dated February 4, 2014SCVB/BC**EPNB/BCRDennigTLupold2/4/142/12/14OGC/NLO withLPL1-1/BCcommentsNAMEDATECJackson2/24/14RElliott2/14/14AKlein3/4/14AGhosh3/13/14OFFICIAL RECORD COPYBBeasley4/14/14DOMINION NUCLEAR CONNECTICUT, INC.DOCKET NO. 50-336MILLSTONE POWER STATION, UNIT NO. 2AMENDMENT TO RENEWED FACILITY OPERATING LICENSEAmendment No. 318Renewed License No. DPR-651.The Nuclear Regulatory Commission (the Commission) has found that:A.The application for amendment by the applicant dated May 3, 2013, assupplemented by letters dated June 27, 2013, July 19, July 30, August 1, andOctober 2, 2013, complies with the standards and requirements of the AtomicEnergy Act of 1954, as amended (the Act), and the Commission's rules andregulations set forth in 10 CFR Chapter I;B.The facility will operate in conformity with the application, the provisions of theAct, and the rules and regulations of the Commission;C.There is reasonable assurance (i) that the activities authorized by thisamendment can be conducted without endangering the health and safety of thepublic, and (ii) that such activities will be conductzed in compliance with theCommission's regulations;D.The issuance of this amendment will not be inimical to the common defense andsecurity or to the health and safety of the public; andE.The issuance of this amendment is in accordance with 10 CFR Part 51 of theCommission's regulations and all applicable requirements have been satisfied.Enclosure 1-22.Accordingly, the license is amended by changes to the Technical Specifications asindicated in the attachment to this license amendment, and paragraph 2.C.(2) ofRenewed Facility Operating License No. DPR-65 is hereby amended to read as follows:(2)Technical SpecificationsThe Technical Specifications contained in Appendix A, as revised throughAmendment No. 318, are hereby incorporated in the renewed license. Thelicensee shall operate the facility in accordance with the Technical Specifications.3.This license amendment is effective as of the date of issuance, and shall beimplemented within 60 days of issuance.FOR THE NUCLEAR REGULATORY COMMISSION/RA/Benjamin G. Beasley, ChiefPlant Licensing Branch I-1Division of Operating Reactor LicensingOffice of Nuclear Reactor RegulationAttachment: Changes to the Licenseand Technical SpecificationsDate of Issuance: April 18, 2014 ATTACHMENT TO LICENSE AMENDMENT NO. 318RENEWED FACILITY OPERATING LICENSE NO. DPR-65DOCKET NO. 50-336Replace the following page of the Renewed Facility Operating License with the attached revisedpage. The revised page is identified by amendment number and contains marginal linesindicating the areas of change.Remove3Insert3Replace the following pages of the Appendix A Technical Specifications, with the attachedrevised page. The revised page is identified by amendment number and contain marginal linesindicating the areas of change.Remove3/4 7-34Insert3/4 7-3411SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIONRELATED TO AMENDMENT NO. 318TO RENEWED FACILITY OPERATING LICENSE NO. DPR-65DOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION, UNIT NO. 2DOCKET NO. 50-3361.0INTRODUCTIONBy letter dated May 3, 2013 (Agencywide Documents Access and Management System(ADAMS) Accession No. ML13133A033), as supplemented by letters dated June 27, 2013(ADAMS Accession No. ML13198A271), July 19, 2013 (ADAMS Accession No. ML13204A035),July 30, 2013 (ADAMS Accession No. ML13213A024), August 1, 2013 (ADAMS Accession No.ML13219A109), and October 2, 2013 (ADAMS Accession No. ML13281A809), DominionNuclear Connecticut, Inc. (the licensee) proposed changes to the Technical Specifications (TSs)for Millstone Power Station Unit 2 (MPS2).The proposed change revises TS 3/4.7.11, “Ultimate Heat Sink” (UHS), to increase the currentUHS water temperature limit from 75 °F to 80 °F and change the TS Action to state, "With theultimate heat sink water temperature greater than 80 °F, be in HOT STANDBY within 6 hoursand in COLD SHUTDOWN within the following 30 hours."The supplemental letters dated June 27, 2013, July 19, July 30, August 1, and October 2, 2013,provided additional information that clarified the application, did not expand the scope of theapplication as originally noticed, and did not change the Nuclear Regulatory Commission (NRC)staff’s original proposed no significant hazards consideration determination as published in theFederal Register on August 20, 2013 (78 FR 51225).2.0REGULATORY EVALUATIONThe regulatory requirements and the guidance upon which the staff based its review of theeffects on containment analyses due to the proposed change are based on the following10 CFR 50 Appendix A General Design Criteria (GDC):GDC-16 as it relates to the containment and associated systems establishing aleak-tight barrier against the uncontrolled release of radioactivity to the environment andassuring that the containment design conditions important to safety are not exceeded foras long as the postulated accident conditions require.GDC-38 as it relates to the containment heat removal system safety function which shallbe to reduce rapidly, consistent with the functioning of other associated12systems, the containment pressure and temperature following any Loss-Of-CoolantAccident (LOCA) and to maintain them at acceptably low levels.GDC-50 as it relates to the containment heat removal system which shall be designedso that the containment structure and its internal compartments can accommodatewithout exceeding the design leakage rate and with sufficient margin, the calculatedpressure and temperature conditions resulting from any LOCA.The MPS2 Final Safety Analysis Report (FSAR) Section 1A lists the extent of MPS compliancewith 10 CFR Part 50 Appendix A of which Criterion 44, Cooling Water, requires a system totransfer the combined heat from Systems Structures and Components (SSC) important to safetyto an UHS under normal and accident conditions with suitable redundancy and specifiedelectrical system availabilities and assuming a single failure. FSAR Section 1A, regardingCriterion 44, requires the RBCCW and SW systems to transfer the combined heat from SSCsunder normal and accident conditions. FSAR Section1A further states, “The RBCCW and SWsystems are provided with suitable redundancy in components and suitable interconnections toassure heat removal capability. The systems are designed to enable isolation of systemcomponents or headers and to detect system mal-operation. The RBCCW and SW systems aredesigned to operate with onsite power (assuming offsite power is not available) and with offsitepower (assuming onsite power is not available). The systems are designed such that a singlefailure in either system will not adversely affect safe operation, accident mitigation, or safeshutdown of the plant.”FSAR Section 9.9.16, “Vital Switchgear Ventilation System,” specifies the maximum allowedroom temperature limits for the vital AC and DC switchgear rooms. The upper and lower4160/6190 volt switchgear rooms, the west 480 volt switchgear room, and the east and westvital DC switchgear rooms are the vital switchgear rooms cooled by the UHS. The SW systemfrom the UHS supplies the cooling water to the ventilation systems and refrigerant system thatcool these vital switchgear rooms. The upper and lower 4160/6190 volt switchgear rooms havea room temperature limit of 122 °F. The west 480 volt switchgear room and the east and westvital DC switchgear rooms have room temperature limits of 104°F.FSAR Section 9.9.8 states that the Engineered Safety Features Room Air Recirculation System(ESFRARS) is designed to limit the maximum ambient temperature to 145 °F except for a brieftransient temperature excursion following an accident. ESFRARS is cooled by RBCCW whichis cooled by SW and the UHS.FSAR Section 6.5.2 states that each containment air recirculation cooling unit is designed forremoving 80 x 106 Btu/hr under Main Steam Line Break (MSLB) accident or LOCA conditionsprior to recirculation with air flow of 34,800 cfm and a fouling factor of 0.0005 for the RBCCWside of the coil. The containment air recirculation and cooling units are cooled by RBCCW whichis cooled by SW and the UHS.FSAR Section 6.3.2 states that the high pressure and low pressure safety injection pumps havemechanical seals. The seals are designed for operation with a pumped fluid temperature of350°F. To permit extended operation under these conditions, a portion of the pump fluid isexternally cooled by the RBCCW system and re-circulated to the seals. The containment spraypumps also have mechanical seals cooled by RBCCW. The seal coolers for these pumps arecooled by RBCCW which is cooled by SW and the UHS.FSAR, Section 6.1.2.1 defines engineered safety features to include safety injection,13containment air recirculation and cooling, reactor building closed cooling system (RBCCW),emergency electrical power (diesel generators) and SW among other safety features. The SWsystem supports safety injection, Emergency Diesel Generators (EDG), and RBCCW.FSAR, Section 9.7.2, states that the SW system shall be designed with suitable redundancy thatin event of a LOCA and a concurrent loss of offsite power and single active failure that the SWsystem can perform its safety functions.3.0TECHNICAL EVALUATIONThe UHS consists of the Service Water (SW) system supplying cooling water to safety relatedloads, i.e. the Reactor Building Closed Cooling Water (RBCCW) system, diesel engine heatexchangers, and the vital AC and DC switchgear ventilation system. The SW system alsosupplies cooling water to non-safety related loads, i.e. Turbine Building Closed Cooling water,chilled water heat exchangers and chlorination system. The water source for the UHS is LongIsland Sound which is connected to the Atlantic Ocean. The SW system has three half capacitypumps taking suction downstream of the traveling screens in the intake structure. Two SWpumps are in continuous operation with a spare pump provided. One pump supplies sufficientheat removal capability for the RBCCW heat exchangers to safely shut down the plant and foraccident mitigation.TS LCO 3.7.11, "Ultimate Heat Sink," currently requires the UHS to have a temperature of lessthan or equal to 75°F in order to be OPERABLE. If 75 °F is exceeded, the ACTION is:a.)With temperature > 75 °F and ≤ 77 °F, operation may continue provided the watertemperature averaged over the previous 24 hour period is verified ≤ 75 °F at least TS LCO3.7.11, "Ultimate Heat Sink," currently requires the UHS to have a temperature of less than orequal to 75°F in order to be OPERABLE. If 75 °F is exceeded, the ACTION is:a.)With temperature > 75 °F and ≤ 77 °F, operation may continue provided thewater temperature averaged over the previous 24 hour period is verified ≤ 75 °Fat least once per hour. Otherwise be in HOT STANDBY within the next 6 hoursand in COLD SHUTDOWN within the following 30 hours.b.)With the UHS water temperature > 77 °F, be in HOT STANDBY within 6 hoursand in COLD SHUTDOWN within the following 30 hours.”The proposed amendment revises TS LCO 3.7.11, by raising the maximum temperature limit ofthe UHS from 75 °F, as stated above, to 80 °F. Also TS 3.7.11 ACTIONS (a) and (b) that arestated above are replaced by new TS 3.7.11 ACTION: “With the UHS water temperaturegreater than 80 °F, be in HOT STANDBY within 6 hours and in COLD SHUTDOWN within thefollowing 30 hours.”COLD SHUTDOWN within the following 30 hours.b.)With the UHS water temperature > 77 °F, be in HOT STANDBY within 6 hoursand in COLD SHUTDOWN within the following 30 hours.”The proposed amendment revises TS LCO 3.7.11, by raising the maximum temperature limit ofthe UHS from 75 °F, as stated above, to 80 °F. Also TS 3.7.11 ACTIONS (a) and (b) that arestated above are replaced by new TS 3.7.11 ACTION: “With the UHS water temperaturegreater than 80 °F, be in HOT STANDBY within 6 hours and in COLD SHUTDOWN within the14following 30 hours.”The Surveillance Requirement (SR) 4.7.11b currently requires that the UHS be determinedOPERABLE: “At least once per 6 hours by verifying the water temperature to be within limitswhen the water temperature exceeds 70°F.” The proposed amendment changes 70°F to 75°Fin SR 4.7.11.The NRC staff performed a review of Millstone Power Station Unit 2 License AmendmentRequest (LAR) and the licensee’s design inputs, assumptions and methodology, whereappropriate, for adherence to regulatory requirements and guidelines. In order to allow a 5 °Frise in UHS temperature and still maintain operability, the equipment important to safety that issupported by the UHS must remain operable. The UHS supports SSC cooled by the SW systemand the RBCCW system. The UHS directly supplies the SW system which in turn coolsRBCCW heat exchangers, the EDG, and supports the Vital Switchgear Ventilation Systems thatcool vital AC and DC switchgear rooms.The RBCCW system in turn cools the containment air recirculation and cooling units,containment spray pump mechanical seal coolers, high and low pressure safety injection pumpmechanical seal coolers, Engineering Safety Features (ESF) room air recirculation coolers, andthe Spent Fuel Pool (SFP) heat exchangers. The SW system and the RBCCW system serveother loads which are non-safety and thus are not evaluated in this safety evaluation.3.1Emergency Diesel GeneratorsThe EDGs are directly cooled by SW. The diesel generators are rated at 2750 KW and must becapable of running for a minimum of 30 days after a design basis accident. For each EDG, theSW system cools three heat exchangers, i.e. an air cooler, a lube oil cooler and a jacket watercooler. The three heat exchangers are in series on the SW side. The licensee stated in theirLAR that the minimum required SW flow at 80 °F inlet temperature is 637 gallons per minute(gpm) for each EDG and that a minimum of 672 gpm was determined available by calculation.However, the NRC staff noted that the SW flow rate and temperature corresponding to the heatexchange values given by the licensee did not agree with vendor data sheets for the EDG heatexchangers.Therefore, in an NRC letter dated July 18, 2013 [Reference 6], and email dated July 23, 2013[Reference 7], the NRC staff requested additional information and justification for the licensee’sstatements that there is sufficient margin in the SW flow to the EDGs to support operation withthe UHS at 80 °F, and to resolve the apparent disagreement with the vendor data sheets. Thelicensee responded in letters dated July 19, 2013 [Reference 3] and July 30, 2013 [Reference4], that the vendor data sheets listed heat exchanger parameters for 118 percent design load(i.e. the 30 minute rating), as explained in DNC calculation 94-DES-1111-M2 was used todetermine the EDG cooling parameters. This calculation determined that 507 gpm of SW wasrequired for continuous EDG loading (2750 KW) with a SW temperature of 77 °F as comparedto 700 gpm for 118 percent load with SW at 75 °F as specified on the vendor data sheet. Thelicensee provided calculation 12-328 (Attachment 9 of Reference 3), which used prevalentnuclear industry software, Proto-HX, to show that 637 gpm SW would be required to remove theEDG heat load at design rating (2750 KW) and SW at 80 °F. According to Reference 5, thelicensee stated that the analytical models of Proto-HX are benchmarked against the actualsystem by verifying results of the model match results obtained in the field. In Reference 3 thelicensee stated that the predicted SW flow of 672 gpm to the EDGs was determined where 637gpm is required when SW temperature is 80 °F. The licensee used Proto-FLO, a widely used15nuclear industry software to model the SW system to determine that the predicted flow of 672gpm would be available to each EDG. The Proto-FLO models are benchmarked to actual flowsobtained during flow testing. With pump degradation inputted as part of the model, predictedflows are below actual flows which add conservatism. The licensee reduced predicted flow rateby 10 percent before comparing it to the acceptance criteria. The SW flow rate tests performedduring 2R20 and 2R21 confirmed that SW flow to each EDG was meeting minimum predictedflow rates. Based on the licensee’s calculations and flow rate tests, the staff finds the licensee’sanalysis that a SW temperature of 80 °F is satisfactory for continuous loading of the EDGs.3.2Vital Switchgear Ventilation SystemsThe SW system supports the Vital Switchgear Ventilation System which cools the vital ACswitchgear rooms containing the west 480 volt, upper and lower 4160/6190 volt switchgearrooms, and the east and west vital DC volt switchgear rooms. The 4160/6190 vital ACswitchgear rooms have a maximum room temperature limit of 122 °F. The west 480 voltswitchgear room has a maximum room temperature limit of 104 °F. In order for UHStemperature of 80 °F to be acceptable, the SW must be capable of keeping the roomtemperatures at or below the above stated temperatures.In the LAR dated May 3, 2013 [Reference 1], the licensee stated that their SW thermal-hydraulicflow analysis with SW/UHS temperature at 80 °F gives acceptable results provided the west 480volt load center switchgear room cooler is cleaned at 18-month intervals. In a Request forAdditional Information (RAI) email dated July 26, 2013 [Reference 8], the NRC staff asked thelicensee specifically whether the room temperature limits specified above and in the FSARwould not be exceeded and whether the thermal performance of the associated cooling coils asspecified on the vendor data sheets would be obtained with the proposed increase in UHStemperature. In the response dated August 1, 2013 [Reference 5], the licensee stated thatProto-HX models were developed for the vital AC switchgear room cooling coils (X-181, X-182,and X-183) and a Mathcad model was developed for the vital DC switchgear room coils(X-169A/B). The models are documented in calculations which determine the minimum SWflow rates to achieve the required heat transfer. To determine predicted SW flow, Proto-FLOmodel of the SW system benchmarked against testing flow data determine whether predictedflow is adequate to remove the required heat. The most recent flow testing data was used toassess flow model uncertainty. Predicted flows were reduced 10 percent for the RBCCW, EDGand west 480 volt cooling coil and by 15 percent for the other vital AC and DC switchgear roomcooling coils. Comparisons of predicted flow and required flow follow.The calculations show that 90 gpm is required for the west 480 volt switchgear room coils(X-181A and X-181B combined) with 80 °F UHS temperature, while the predicted flow is145 gpm. The calculations show that 17 gpm is required for the upper 4160/6900 voltswitchgear room coil (X-183) with an 80 °F UHS temperature, while the predicted flow is 23 gpmfor X-183. The calculations show that 15 gpm is required for lower 4160/6900 volt switchgearroom coil (X-182) with an 80 °F UHS temperature, while the predicted flow is 28 gpm for X-182.The calculations show that 26.9 gpm is required for the east and west vital DC, while thepredicted delivered flow is 30 gpm.Based on the licensee’s calculations and flow rate tests, the staff finds the licensee’s analysisfor a SW/UHS temperature of 80 °F is satisfactory because there are sufficient margins in flowrates for supporting the Vital Switchgear Ventilation Systems and meeting the maximum roomtemperature requirements specified in FSAR Section 9.9.16.163.3Reactor Building Component Cooling Water SystemThe RBCCW system has two redundant trains each with a RBCCW heat exchanger. Themaximum load on a RBCCW heat exchanger is 204 MBTU/hr during a design basis event. Asstated above, the RBCCW heat exchangers are directly cooled by the SW and the UHS. TheRBCCW system in turn cools containment air recirculation and cooling units, containment spraypump mechanical seal coolers, high and low pressure safety injection pump mechanical sealcoolers, ESF room air recirculation coolers, and the SFP heat exchangers.In the LAR the licensee stated that the RBCCW supply temperature of 85 °F in Modes 1,2,3 willbe a design requirement and will be obtained using modified operating procedures to minimizeRBCCW heat loads and maximize SW flow to the RBCCW heat exchangers whenever UHStemperature exceeds 75 °F. The licensee stated that the RBCCW system will be able toperform its intended functions with an UHS temperature of 80 °F. The licensee did notsufficiently justify the proposed increase in the temperature limit of the UHS. Therefore, theNRC staff submitted RAIs in a letter dated July 18, 2013 [Reference 6] and emails dated July 23and 26, 2013 [Reference 7 and 8] asking the licensee to provide the quantitative effects andacceptability of the increase in RBCCW cooling water on all safety related loads cooled byRBCCW. The staff asked the licensee to discuss the ability of the RBCCW system to meet thedesign requirements ensuring that the cooling requirements for the ESF room air recirculationcoolers, containment air recirculation and cooling units, high pressure and low pressure andcontainment spray pump mechanical seals and SFP are achieved.In responses dated July 19, 2013 [Reference 3], July 30, 2013 [Reference 4] and August 1,2013 [Reference 5], the licensee explained that in performing the design basis accident (DBA)analyses, it analyzed for both maximum effect on the containment and maximum effect on theRBCCW heat exchanger outlet temperatures considering the maximum heat input to RBCCW inboth the injection and recirculation modes of LOCA mitigation as well as the MSLB analysis.The licensee stated that with a UHS temperature of 80 °F, the RBCCW cooled ESF room airrecirculation coolers and still maintained the ESF room maximum ambient temperature below145 °F and was within the Equipment Environmental Qualification (EEQ) limits. Using theDominion Gothic methodology for containment response following a LOCA and MSLB with aUHS temperature of 80 °F, the licensee determined that safety related equipment incontainment will be within their EEQ limits. Analyses also showed that the increase in UHS to80 °F was within acceptance criteria for seal performance of the mechanical seals of the safetyinjection and containment spray pumps because they within their EEQ limits. The staff finds thelicensee’s evaluation acceptable because the licensee used NRC approved methodology andthe RBCCW cooled ESF room air recirculation coolers and still maintained the ESF roommaximum ambient temperature within the Equipment Environmental Qualification (EEQ) limits.For SFP cooling with the UHS at 80 °F and the limiting RBCCW temperature profile, thelicensee stated that the maximum SFP temperature has been analytically determined to bebelow 200 °F. This is based on an SFP maximum initial temperature of 150 °F and heat up dueto suspension of cooling for four hours when RBCCW is isolated to the SFP at the start of aLOCA and manually restored fours after the LOCA. The analytically determined limit of below200 °F is less than 212 °F which is the assumed temperature during accident conditions asspecified in FSAR Section 5.4.3.1.3, “Thermal Loads.” Therefore, the staff finds the increase inUHS limit to 80 °F acceptable for SFP cooling.173.4Containment3.4.1Mass and Energy (M&E) Release DataThe licensee stated that the revised containment analysis incorporated corrected M&E releasedata for use as input to the LOCA and Main Steamline Break (MSLB) containment responseanalysis. The NRC staff requested in a RAI that the licensee describe the basis and reasons forthe revision in the M&E release data. In response to SCVB-RAI-1 (Reference 9), the licenseestated that Westinghouse identified that the computer code CEFLASH-4A and erroneouslyunder predicted the M&E release data for the LOCA and MSLB blowdown phase.Westinghouse provided the revised data for the reactor coolant system hot leg, pump suctionand pump discharge legs double ended guillotine large breaks for the LOCA blowdown andreflood phases. Westinghouse also provided corrected M&E release data using the NRCapproved CE methodology using SGN-III computer program (FSAR Section 14.8.2.1.3) forMSLB initiated at 0-, 25-, 50-, 75-, and 102-percent power levels assuming several cases of asingle failure. The staff finds the licensee’s evaluation acceptable because the licensee usedNRC approved methodology and the corrected M&E data.3.4.2Containment Response AnalysisThe current containment analysis was performed by the licensee using the maximum value ofUHS temperature of 77 °F (Reference 11).3.4.2.1 LOCA Pressure and Temperature ResponseThe licensee re-analyzed the LOCA containment response using the GOTHIC methodologyapproved, for use by DNC only, by NRC in Reference 10, with the corrected WestinghouseM&E data for the LOCA blowdown and reflood phases. For evaluating the LOCA post-refloodand long term phases, the licensee used NRC approved GOTHIC methodology both fordetermining M&E release data and the containment response (Reference 10). The licensee’sevaluated cases with and without concurrent loss of offsite power while assuming a UHStemperature of 80 °F. The analysis resulted in a peak containment pressure of 52.5 psig, and apeak containment gas temperature of 279.2 °F. The calculated peak containment pressure andgas temperature are bounded by the containment design pressure 54 psig and the containmentliner and structural design temperature 289 °F. The effect of the increased UHS temperaturefrom the current 77 °F to the proposed TS limit of 80 °F increases the calculated containmentpeak pressure by 0.03 psi, because at the time of peak pressure, the safety related containmentair recirculation and cooling system is in operation. The licensee stated calculation of the peakcontainment liner temperature is not required because the peak containment gas temperaturefor EQ is less than the containment structural and liner design temperature. The staff finds thelicensee’s evaluation acceptable because the licensee used NRC approved methodology andthe calculated peak containment pressure and gas temperature are bounded by thecontainment design pressure and the containment liner and structural design temperature.3.4.2.2 MSLB Pressure and Temperature ResponseThe licensee re-analyzed the MSLB containment response using the NRC approved GOTHICmethodology (Reference 10), with the corrected Westinghouse M&E data, and assuming a UHStemperature of 80 °F. The analysis resulted in a peak containment pressure of 53.8 psig and apeak containment gas temperature of 360.9 °F. The licensee stated that the effect of theincreased UHS temperature from the current 77 °F to the proposed TS limit of 80 °F increases18the calculated containment peak pressure by 0.1 psi, and the predicted peak containment gastemperature increases by less than 0.1 °F. Since the calculated peak drywell gas temperatureis greater than the containment liner and structural design temperature, the licensee determinedthe liner temperature response using the methodology approved by NRC in Reference 10. Thecalculated peak containment liner temperature was 259.7 °F, which is less than its designtemperature of 289 °F. The effect of the increased UHS temperature from the current 77 °F tothe proposed TS limit of 80 °F increases the calculated containment peak liner temperature byless than 0.1 °F. The staff finds the licensee’s evaluation acceptable because the licensee usedNRC approved methodology and the calculated peak containment liner temperature was lessthan its design temperature.3.4.2.3 Net Positive Suction Head AnalysisThe NRC staff requested in a RAI that the licensee describe the impact of increasing the UHStemperature to the proposed TS limit of 80 °F on the post-accident maximum sump watertemperature and the available Net Positive Suction Head (NPSH) for the Emergency CoreCooling System (ECCS) and containment spray pumps following the Sump RecirculationActuation Signal (SRAS). The staff also requested to describe if the NPSH analysis complieswith Regulatory Guide (RG) 1.1 (Safety Guide 1). In response to SCVB-RAI-2 (Reference 9),the licensee stated that the analysis was based on the NRC approved methodology inReference 10 and a UHS temperature of 80 °F which resulted in a peak sump temperature of233.5 °F following the SRAS. The licensee stated that the impact of the increased UHStemperature to 80 °F increased the peak value of the sump water temperature following SRASby 1.2 °F. Following RG 1.1, assuming a sump water temperature of 212 °F at 14.7 psiapressure, the licensee conservatively calculated the minimum available NPSH without creditingthe Containment Accident Pressure (CAP) for the ECCS and containment spray pumps duringthe sump recirculation mode. The licensee stated that the conservatively calculated availableNPSH following RG 1.1 after SRAS exceeded the required NPSH for the ECCS andcontainment spray pumps. The staff finds the licensee’s evaluation acceptable because thelicensee followed GR 1.1 and conservatively calculated the minimum available NPSH withoutcrediting the CAP for the ECCS and containment spray pumps during the sump recirculationmode.3.4.2.4 Minimum Containment Pressure Analysis for ECCS PerformanceNUREG-0800, Standard Review Plan (SRP) 6.2.1.5 describes the minimum containmentpressure analysis for ECCS performance capability. RG 1.157, Section 3.12.1 providesguidance for calculating the containment pressure response used for evaluating coolingeffectiveness during the post-blowdown phase of a LOCA. The RG states that the containmentpressure should be calculated by including the effects of containment heat sinks and operationof all pressure reducing equipment assumed to be available. The NRC staff, requested in aRAI, that the licensee describe the impact of increasing UHS temperature from 75 °F to 80 °Fand the changes in M&E release on the minimum containment pressure analyses for ECCSperformance as described in the above NRC documents. In response to SCVB-RAI-3(Reference 9), the licensee stated that the M&E release does not impact the minimumcontainment pressure analyses for ECCS performance because the M&E input to the 10 CFR50.46 ECCS performance analysis is developed by AREVA, the current fuel supplier, using the10 CFR 50.46 large break LOCA analysis methodology. The licensee stated that the AREVAmethodology was not affected by the Westinghouse LOCA M&E release error. The licenseestated that conservative assumptions were used in the AREVA methodology which maximizes19containment heat removal for calculating the minimum containment pressure response followinga large break LOCA. The NRC staff finds the licensee response acceptable.MPS2 uses the design pressure of 54 psig as the value of Pa in the TS. Since the calculatedcontainment peak pressure is less than the value of Pa, the NRC staff concludes that thecontainment leakage testing program is not adversely affected.3.5Revisions to the UFSAR Concurrent with the License AmendmentThe licensee’s thermal-hydraulic analysis of the SW demonstrated acceptable results with thefollowing restrictions:The maximum allowable Differential Pressure (DP) for the 'A' RBCCW heat exchangerhas been reduced from 10-psid to 8-psid.RBCCW heat exchangers must be cleaned at a 3-month interval.Switchgear room coolers, X-181 A/B, must be cleaned at an 18-month interval.For SW temperatures above 75 °F, it may be necessary to also open the RBCCW wintertemperature control valves (2-SW-245, 2-SW-246, and 2-SW-247) to maximize flow tothe RBCCW heat exchangers to support normal operation.An RBCCW supply temperature of 85 °F in Modes 1, 2, and 3 will be included in theUFSAR as a design requirement for the RBCCW system and the design basis reasonwith respect to Generic letter 96-06 shall be stated.Implementation of the amendment to TS 3/4.7.11 shall also include a revision to the UpdatedFinal Safety Analysis Report (UFSAR) to include the above stated restrictions. The NRC staffhas reviewed the licensee’s analysis provided in References 1 through 5 and finds that raisingthe temperature limit of LCO 3/4.7.11 to 80 °F while remaining in Modes 1, 2, or 3 is acceptable.Based on these findings, the NRC staff concludes that there is reasonable assurance that therequirements of 10 CFR Part 50 Appendix A Criterion 44 and the UFSAR will continue to bemet.3.6Program ReviewsThe licensee submitted program information for Fire Protection/Appendix R, In-ServiceInspection, and In-Service Testing, and the NRC staff determined based on the informationsubmitted that the impact on these programs was bounded by the SR information and istherefore acceptable.4.0STATE CONSULTATIONIn accordance with the Commission's regulations, the Connecticut State official was notified ofthe proposed issuance of the amendment. The State official had no comments.5.0ENVIRONMENTAL CONSIDERATIONThe amendment changes a requirement with respect to installation or use of a facilitycomponent located within the restricted area as defined in 10 CFR Part 20. The NRC staff hasdetermined that the amendment involves no significant increase in amounts, and no significantchange in the types of any effluents that may be released offsite, and that there is no significantincrease in individual or cumulative occupational radiation exposure. The Commission has20previously issued a proposed finding that the amendment involves no significant hazardsconsideration, and there has been no public comment on such finding (78 FR 51225).Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement orenvironmental assessment need be prepared in connection with the issuance of the amendment.6.0CONCLUSIONThe Commission has concluded, based on the considerations discussed above, that (1) there isreasonable assurance that the health and safety of the public will not be endangered byoperation in the proposed manner, (2) there is reasonable assurance that such activities will beconducted in compliance with the Commission's regulations, and (3) the issuance of theamendment will not be inimical to the common defense and security or to the health and safetyof the public.7.0REFERENCES1. Dominion Letter 13-227, dated May 3, 2013, Millstone Power Station Unit 2 LAR forChanges to TS 3/4.7.11 UHS (Agencywide Documents Access and ManagementSystem (ADAMS) Accession No. ML13133A033).2. Dominion Letter 13-227A, dated June 27, 2013, Millstone Power Station Unit 2Supplement to LAR for Changes to TS 3/4.7.11 UHS (ADAMS AccessionNo. ML13198A271).3. Dominion Letter 13-419, dated July 19, 2013, Millstone Power Station Unit 2 Responseto Request for Additional Information Regarding LAR for Changes to TS 3/4.7.11 UHS.(ADAMS Accession No. ML13204A035).4. Dominion Letter 13-438, dated July 30, 2013, Millstone Power Station Unit 2 Responseto Request for Additional Information Regarding LAR for Changes to TS 3/4.7.11 UHS.(ADAMS Accession No. ML13213A024).5. Dominion Letter 13-450, dated August 1, 2013, Millstone Power Station Unit 2 Responseto Request for Additional Information Regarding LAR for Changes to TS 3/4.7.11 UHS.(ADAMS Accession No. ML13219A109).6. NRC Letter to Mr. David A. Heacock, Dominion Nuclear Connecticut, Inc. dated July 18,2013. (ADAMS Accession No. ML13197A401).7. NRC Email to William Bartron dated July 23, 2013 (ADAMS AccessionNo. ML13206A021).8. NRC Email to William Bartron dated July 26, 2013 (ADAMS AccessionNo. ML13213A067).9. DNC Letter to NRC dated October 2, 2013, “Dominion Nuclear Connecticut. Inc.Millstone Power Station Unit 2 Response to Request for Additional InformationRegarding License Amendment Request for Changes to Technical Specification3/4.7.11, "Ultimate Heat Sink" (ADAMS Accession No. ML13281A809).10. Letter from NRC to DNC dated August 30, 2006, “Kewaunee Power Station(Kewaunee), Millstone Power Station, Unit Nos. 2 And 3 (Millstone 2 And 3), North AnnaPower Station, Unit Nos. 1 And 2 (North Anna 1 And 2), and Surry Power Station, Unit21Nos. 1 and 2 (Surry 1 And 2) - Approval of Dominion’s Topical Report DOM-NAF-3,“GOTHIC Methodology for Analyzing the Response to Postulated Pipe Ruptures InsideContainment” (TAC NOS. MC8831, MC8832, MC8833, MC8834, MC8835, andMC8836)”, (ADAMS Accession No. ML062420511).11. Letter from NRC to DNC dated May 31, 2001, “Millstone Nuclear Power Station, UnitNo.2 - Issuance of Amendment Re: Ultimate Heat Sink Action Requirements (TAC No.MB0867)”, (ADAMS Accession No. ML011410153).Principal Contributors:Date: April 18, 2014 G. PurciarelloA. Sallman