ML14237A099

Millstone Power Station Unit 3 - License Amendment Request, Proposed Technical Specifications Change to the Auxiliary Feedwater System

ML14237A099
Person / Time
Site:	Millstone
Issue date:	08/19/2014
From:	Sartain M D Dominion, Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
14-301
Download: ML14237A099 (25)

Text

Dominion Nuclear Connecticut, Inc. " *5000 Dominion Boulevard, Glen Allen, VA 23060 m inionWeb Address: www.dom.comAugust 19, 2014U.S. Nuclear Regulatory Commission Serial No. 14-301Attention: Document Control Desk NSSL/MAE ROWashington, DC 20555 Docket No. 50-423License No. NPF-49DOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 3LICENSE AMENDMENT REQUEST, PROPOSED TECHNICAL SPECIFICATIONSCHANGE TO THE AUXILIARY FEEDWATER SYSTEMPursuant to 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC) requests anamendment to Operating License NPF-49 for Millstone Power Station Unit 3 (MPS3).The proposed change will revise Technical Specification (TS) 3/4.7.1.2, "AuxiliaryFeedwater System," Surveillance Requirement (SR) 4.7.1.2.1.b. The proposed TSchange is consistent with the Standard Technical Specifications for WestinghousePlants (NUREG-1431, Revision 4).The proposed change has been reviewed and approved by the Facility Safety ReviewCommittee.Information provided in the attachments to this letter is summarized below:" Attachment 1 provides the Description, Technical Evaluation, RegulatoryEvaluation and Environmental Consideration for the proposed TechnicalSpecifications change. As discussed in this attachment, the proposedamendment does not involve a significant hazards consideration pursuant to theprovisions of 10 CFR 50.92." Attachment 2 contains marked-up pages to reflect the proposed change to theTechnical Specifications." Attachment 3 contains marked-up pages to reflect the proposed change to theTechnical Specifications Bases for information only and will be implemented inaccordance with the Technical Specification Bases Control Program.DNC requests approval of the proposed amendment by August 19, 2015. Onceapproved, the amendment will be implemented within 120 days.In accordance with 10 CFR 50.91(b), a copy of this license amendment request is beingprovided to the State of Connecticut.

Serial No. 14-301Docket No. 50-423Page 2 of 3If you have any questions regarding this submittal, please contact Wanda Craft at (804)273-4687.Sincerely,Mark D. SartainVice President -Nuclear EngineeringCOMMONWEALTH OF VIRGINIA ))COUNTY OF HENRICO )The foregoing document was acknowledged before me, in and for the County andCommonwealth aforesaid, today by Mark D. Sartain, who is Vice President -NuclearEngineering of Dominion Nuclear Connecticut, Inc. He has affirmed before me that he is dulyauthorized to execute and file the foregoing document in behalf of that Company, and that thestatements in the document are true to the best of his knowledge and belief.Acknowledged before me this /ý --'-day of AL //4/L(o4..., 2014.My Commission Expires: '5 31-loO -JINotary PublicReg.

• 140542[M4y Co~mm1@iuon Eupires My 31, 2016>Commitments made in this letter: NoneAttachments:1. Evaluation of the Proposed Change2. Marked-up Technical Specifications Pages3. Marked-up Technical Specifications Bases Pages for information only Serial No. 14-301Docket No. 50-423Page 3 of 3cc: U.S. Nuclear Regulatory CommissionRegion I2100 Renaissance BlvdSuite 100King of Prussia, PA 19406-2713Mohan ThadaniSenior Project ManagerU.S. Nuclear Regulatory CommissionOne White Flint North, Mail Stop 08 B 111555 Rockville PikeRockville, MD 20852-2738NRC Senior Resident InspectorMillstone Power StationDirector, Radiation DivisionDepartment of Energy and Environmental Protection79 Elm StreetHartford, CT 06106-5127 Serial No. 14-301Docket No. 50-423Attachment IEvaluation of the Proposed ChangeDOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 3 Serial No. 14-301Docket No. 50-423Attachment 1, Page 1 of 101.0 DESCRIPTION2.0 PROPOSED CHANGE3.0 TECHNICAL ANALYSIS4.0 REGULATORY SAFETY ANALYSIS4.1 Applicable Regulatory Requirements/Criteria4.2 No Significant Hazards Consideration

5.0 ENVIRONMENTAL CONSIDERATION

Serial No. 14-301Docket No. 50-423Attachment 1, Page 2 of 101.0 DESCRIPTIONThe proposed change will revise Millstone Power Station Unit 3 (MPS3) TechnicalSpecification (TS) 3/4.7.1.2, "Auxiliary Feedwater System," SurveillanceRequirement (SR) 4.7.1.2.1.b. The proposed change replaces the surveillancefrequency and acceptance criteria for the AFW pumps with a reference to theInservice Testing (IST) program (TS 4.0.5) for the specific pump testing acceptancecriteria and the surveillance frequency, which is consistent with the other pump andvalve surveillance requirements in the TS. The proposed change also addsinformation on suitable plant conditions for performance of the steam turbine drivenAFW pump surveillance. The proposed TS change is consistent with the StandardTechnical Specifications (STS) for Westinghouse Plants (NUREG-1431, Revision 4).2.0 PROPOSED CHANGESR 4.7.1.2.1 .b., "Auxiliary Feedwater System"The current SR states:"At least once per 92 days on a STAGGERED TEST BASIS, tested pursuant toSpecification 4.0.5, by:1) Verifying that on recirculation flow each motor-driven pump develops a total headof greater than or equal to 3385 feet;2) Verifying that on recirculation flow the steam turbine-driven pump develops atotal head of greater than or equal to 3780 feet when the secondary steamsupply pressure is greater than 800 psig. The provisions of Specification 4.0.4are not applicable for entry into MODE 3."The SR will be changed to read as follows:----------------------- NOTE --------------------------Not required to be performed for the steam turbine driven auxiliary feedwater pumpuntil 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 800 psig in the steam generators.Verify the developed head of each auxiliary feedwater pump at the flow test point isgreater than or equal to the required developed head when tested pursuant toSpecification 4.0.5. The provisions of Specification 4.0.4 are not applicable to thesteam turbine driven auxiliary feedwater pump for entry into MODE 3."

Serial No. 14-301Docket No. 50-423Attachment 1, Page 3 of 10TS Bases 3/4.7.1.2, "Auxiliary Feedwater System":TS Bases 3/4.7.1.2 has been modified to: 1) include a discussion addressingchanges to SR 4.7.1.2.1.b, 2) add additional discussion addressing SR 4.7.1.2.1.c,and 3) add additional discussion addressing SR 4.7.1.2.2. The proposed change tothe TS Bases is provided for information only and will be implemented in accordancewith the TS Bases Control Program.3.0 TECHNICAL ANALYSIS3.1 System DescriptionThe AFW system is described in MPS3 Final Safety Analysis Report (FSAR)Section 10.4.9. The AFW system consists of two motor-driven auxiliaryfeedwater pumps, one turbine-driven auxiliary feedwater pump, and theassociated piping and valves necessary to connect the demineralized waterstorage tank (DWST) to the pump suctions, and the pump discharges to thefeedwater system.Two half-size motor-driven auxiliary feedwater pumps and one full-size turbine-driven auxiliary feedwater pump are provided. Sufficient auxiliary feedwater forplant cooldown can be supplied by either the turbine-driven pump or the twomotor-driven pumps. A single motor-driven auxiliary feedwater pump is alsocapable of supplying auxiliary feedwater flow to two intact steam generators.This capacity is provided to protect against multiple failures as well as to providepower source diversity.The steam generator auxiliary feedwater pumps are used as an emergencysource of feedwater supply to the steam generators. The pumps are required toensure safe shutdown in the event of loss of power or function as an EngineeredSafeguards System to remove core decay heat. The pumps are on standbyservice during normal plant operation.3.2 Current Licensing BasesThe MPS3 design was reviewed in accordance with NUREG-0800, "StandardReview Plan for the Review of Safety Analysis Report for Nuclear Power Plants,"SRP 6.2.1.1.A, Rev. 2, July 1981.As noted in the Final Safety Analysis Report (FSAR) Section 3.1, the designbases for MPS3 was reviewed against the NRC General Design Criteria forNuclear Power Plants, 10 CFR 50, Appendix A, as amended through October 27,1978. The adequacy of the MPS3 design relative to the design criteria isdiscussed in the FSAR Sections 3.1.1 and 3.1.2.

Serial No. 14-301Docket No. 50-423Attachment 1, Page 4 of 10The auxiliary feedwater system is designed in accordance with the followingcriteria.1. General Design Criterion 2, for structures housing the system and thesystem itself being capable of withstanding the effects of naturalphenomena such as earthquakes, tornadoes, hurricanes, and floods.2. General Design Criterion 4, with respect to structures housing the systemand the system itself being capable of withstanding the effects of externalmissiles and internally generated missiles, pipe whip, and jet impingementforces associated with pipe breaks.3. General Design Criterion 5, for shared systems and components importantto safety being capable to perform required safety functions.4. General Design Criterion 19, for the design capability of systeminstrumentation and controls for prompt hot shutdown of the reactor andpotential capability for subsequent cold shutdown.5. General Design Criterion 34, to ensure:a. The capability of the auxiliary feedwater system to sufficiently transferfission product decay heat and other residual heat from the reactorcore at a rate such that specified acceptable fuel design limits and thedesign conditions of the reactor coolant pressure boundary are notexceeded.b. Suitable redundancy in components, features, interconnections, leakdetection, and isolation capabilities is provided to assure, underassumption of a single failure, the continued safety function regardlessof the loss of either onsite, offsite, or the generating capability of bothpower systems.6. General Design Criterion 44, to ensure:a. The capability to transfer heat loads from the reactor system to a heatsink under both normal operating and accident conditions.b. Redundancy of components so that under accident conditions thesafety function can be performed assuming a single active componentfailure (This may be coincident with the loss of offsite power for certainevents).c. The capability to isolate components, subsystems, or piping, ifrequired, so that the system safety function is maintained.

Serial No. 14-301Docket No. 50-423Attachment 1, Page 5 of 107. General Design Criterion 45, for design provisions to permit periodicinservice inspection of system components and equipment.8. General Design Criterion 46, for design provisions to permit appropriatefunctional testing of the system and components to ensure structuralintegrity and leak tightness, operability and performance of activecomponents, and capability of the integrated system to function asintended during normal, shutdown, and accident conditions.9. General Design Criterion 57, for design provisions to ensure that each linethat penetrates primary reactor containment and is neither part of thereactor coolant pressure boundary nor connected to the containmentatmosphere shall have at least one containment isolation valve which shallbe either automatic, or locked closed, or capable of remote manualoperation.10.The following Regulatory Guides (subject to exceptions as specified inMPS3 FSAR Section 1.8, "Conformance with NRC Regulatory Guides"):" Regulatory Guide 1.26, for the quality group classification of systemcomponents." Regulatory Guide 1.29, for seismic design classification of systemcomponents." Regulatory Guide 1.62, for design provisions made for manualinitiation of each protective action." Regulatory Guide 1.102, for the protection of structures, systems,and components important to safety from the effects of flooding.* Regulatory Guide 1.117, for the protection of structures, systems,and components important to safety from the effects of tornadomissiles.11. Branch Technical Positions APCSB 3-1 and MEB 3-1, for breaks in highand moderate energy piping systems outside containment.12. Branch Technical Position ASB 10-1, for auxiliary feedwater pump driveand power supply diversity.13. Branch Technical Position RSB 5-1 for safety grade cold shutdown.

Serial No. 14-301Docket No. 50-423Attachment 1, Page 6 of 103.3 Analysis of the Proposed ChangesThe proposed change would revise TS SR 4.7.1.2.1.b to remove the surveillancefrequency and acceptance criteria for the AFW pumps and reference the ISTProgram as a basis for the surveillance requirement. The IST Program (TS 4.0.5)will verify the component acceptance criteria, consistent with design basisrequirements, and control the frequency of test performance. The acceptancecriteria (e.g., developed head, pump flowrate) are based on the design basisrequirements. Performance of the required testing will verify proper componentoperation, and will detect component degradation. The frequency of testperformance may change based on equipment performance.The use of the IST Program to control pump and valve testing is consistent withcurrent industry practices and published guidelines. Many of the surveillancerequirements contained in NUREG-1431, Rev. 4, illustrate the use of the ISTProgram to verify the acceptance criteria and control the frequency of testperformance. The surveillance requirements contained in NUREG-1431 refer to theIST Program; however, because MPS3 has custom TS, the surveillancerequirements in MPS3 TS refer to Specification 4.0.5 (Inservice Testing Program).Individual NUREG-1431 surveillance requirements refer to the IST Program which iscontained in Section 5 (Technical Specification 5.5.8, "Inservice Testing Program").However, since the MPS3 TS still contains TS 4.0.5, the proposed surveillancerequirements will refer to "Specification 4.0.5."The proposed changes in SR 4.7.1.2.b are justified as follows:1. The surveillance frequency in SR 4.7.1.2.b would be replaced with areference to the IST Program.The IST Program specifies a minimum test performance interval of 92 days.Therefore, the surveillance frequency when tested pursuant to Specification4.0.5 will remain unaffected.2. The pump acceptance criteria in SRs 4.7.1.2.1.b.1 and 4.7.1.2.1.b.2 would bereplaced with a reference to the IST Program.The pump acceptance criteria specified by design basis requirements isverified by the IST Program, which is referenced (Specification 4.0.5) in theproposed SR 4.7.1.2.1.b. It is not necessary to specify the acceptancecriteria in the surveillance requirement. The IST Program provides sufficientcontrol of this value to ensure the associated pumps will perform as assumedin the accident analysis. Removal of this specific value will not adverselyimpact test performance. This approach to allow the IST Program to specifythe acceptance criteria based on design basis requirements is consistent withNUREG-1431 Rev. 4 (SR 3.7.5.2).

Serial No. 14-301Docket No. 50-423Attachment 1, Page 7 of 10The IST program requires verification of the acceptance criteria of the AFWpumps at the flow test point rather than at recirculation flow conditions, amethod acceptable by the ASME OM Code. The proposed change conformswith the STS in testing of the AFW pumps and does not reduce the testingrequirements for the AFW pumps.3. The addition of the Note to the proposed SR 4.7.1.2.b that the test does nothave to be performed for the steam turbine driven AFW pump until 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />safter reaching 800 psig in the steam generators.This change will provide additional guidance to the plant operators and allowsdeferral of the surveillance until suitable plant conditions are established andthe plant is stable. This deferral is required because there may be insufficientsteam pressure to perform the test prior to reaching 800 psig in the steamgenerators. The use of a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time limit is consistent with the guidancecontained in Generic Letter (GL) 87-09. This approach to address theperformance of surveillance requirements that cannot be performed untilcertain plant conditions are established is consistent with NUREG-1431, Rev.4 (SR 3.7.5.2).

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/CriteriaIn Section 50.36, "Technical specifications," of Title 10 of the Code of FederalRegulations (10 CFR), the Commission established its regulatory requirementsrelated to the content of technical specifications. Pursuant to 10 CFR 50.36(d),technical specifications are required to include items in the following five specificcategories related to station operation: (1) safety limits, limiting safety systemsettings, and limiting control settings; (2) limiting conditions for operation; (3)surveillance requirements; (4) design features; and (5) administrative controls.This license amendment request deals with a proposed change to a surveillancerequirement.The design criteria for the AFW system are addressed in section 3.2 above.4.2 No Significant Hazards ConsiderationPursuant to 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC) requestsamendment to Operating License NPF-49 for Millstone Power Station Unit 3(MPS3). The proposed changes will revise Technical Specification (TS)3/4.7.1.2, "Auxiliary Feedwater System," Surveillance Requirement (SR)4.7.1.2.1.b. The proposed TS change is consistent with the Standard TechnicalSpecifications for Westinghouse Plants (NUREG-1431, Revision 4).

Serial No. 14-301Docket No. 50-423Attachment 1, Page 8 of 10According to 10 CFR 50.92(c), a proposed amendment to an operating licenseinvolves no significant hazards consideration if operation of the facility inaccordance with the proposed amendment would not:1. Involve a significant increase in the probability or consequences of anaccident previously evaluated; or2. Create the possibility of a new or different kind of accident from any accidentpreviously evaluated; or3. Involve a significant reduction in a margin of safety.In support of this determination, an evaluation of each of the three criteria setforth in 10 CFR 50.92 is provided below regarding the proposed licenseamendment.1. The proposed amendment does not involve a significant increase in theprobability or consequences of an accident previously evaluated.Response: NoThe proposed amendment associated with the modifications to the existingsurveillance requirement will not cause an accident to occur and will not result inany change in the operation of the associated accident mitigation equipment.The ability of the equipment associated with the proposed amendment to mitigatethe design basis accidents will not be affected. The proposed TechnicalSpecification surveillance requirement is sufficient to ensure the requiredaccident mitigation equipment will be available and function properly for designbasis accident mitigation. In addition, the design basis accidents will remain thesame postulated events described in the MPS3 Final Safety Analysis Report, andthe consequences of those events will not be affected.Therefore, the proposed amendment will not significantly increase the probabilityor consequences of an accident previously evaluated.2. The proposed amendment does not create the possibility of a new or differentkind of accident from any accident previously evaluated.Response: NoThe proposed amendment to the Technical Specifications surveillancerequirement does not impact any system or component that could cause anaccident. The proposed amendment does not involve a physical alteration of theplant. No new or different types of equipment will be installed and there are nophysical modifications to existing equipment associated with the proposedamendment. The proposed amendment will not alter the way any structure, Serial No. 14-301Docket No. 50-423Attachment 1, Page 9 of 10system, or component functions, and will not alter the manner in which the plantis operated or require any new operator actions. There will be no adverse effecton plant operation or accident mitigation equipment. The response of the plantand the operators following an accident will not be different. In addition, theproposed amendment does not create the possibility of a new failure modeassociated with any equipment or personnel failures.Therefore, the proposed amendment will not create the possibility of a new ordifferent kind of accident from any accident previously evaluated.3. The proposed amendment does not involve a significant reduction in a marginof safety?Response: NoThe proposed amendment to the Technical Specification surveillancerequirement will not cause an accident to occur and will not result in any changein the operation of the associated accident mitigation equipment. The equipmentassociated with the proposed Technical Specification surveillance requirementwill continue to be able to mitigate the design basis accidents as assumed in thesafety analysis. The proposed surveillance requirement is adequate to ensureproper operation of the affected accident mitigation equipment. In addition, theproposed amendment will not affect equipment design or operation, and thereare no changes being made to the Technical Specification required safety limitsor safety system settings. The proposed amendment, in conjunction with the ISTProgram, will provide adequate control measures to ensure the accidentmitigation functions are maintained.Therefore, the proposed amendment will not result in a significant reduction in amargin of safety.ConclusionBased upon this discussion, it is concluded that the proposed TechnicalSpecification change to revise TS 3/4.7.1.2, "Auxiliary Feedwater System," SR4.7.1.2.1.b, does not involve a significant hazards consideration.

5.0 ENVIRONMENTAL CONSIDERATION

DNC has evaluated this proposed license amendment consistent with the criteria foridentification of licensing and regulatory actions requiring environmental assessmentin accordance with 10 CFR 51.21, "Criteria for and identification of licensing andregulatory actions requiring environmental assessments." DNC has determined thatthis proposed change meets the criteria for categorical exclusion set forth inparagraph (c)(9) of 10 CFR 51.22, "Criterion for categorical exclusion; identificationof licensing and regulatory actions eligible for categorical exclusion or otherwise not Serial No. 14-301Docket No. 50-423Attachment 1, Page 10 of 10requiring environmental review," and has determined that no irreversibleconsequences exist in accordance with paragraph (b) of 10 CFR 50.92, "Issuance ofamendment." This determination is based on the fact that this proposed change isbeing processed as an amendment to the license issued pursuant to 10 CFR 50,"Domestic Licensing of Production and Utilization Facilities," which changes arequirement with respect to installation or use of a facility component located withinthe restricted area, as defined in 10 CFR 20, "Standards for Protection AgainstRadiation," or which changes an inspection or surveillance requirement and theamendment meets the following specific criteria :1. The amendment involves no significant hazards consideration.As demonstrated in Section 5.2 above, "No Significant Hazards Consideration,"the proposed change does not involve any significant hazards consideration.2. There is no significant change in the types or significant increase in the amountsof any effluent that may be released offsite.The proposed changes would revise TS 3/4.7.1.2, "Auxiliary Feedwater System,"SR 4.7.1.2.1.b. The proposed changes do not result in an increase in powerlevel, and do not increase the production nor alter the flow path or method ofdisposal of radioactive waste or byproducts; thus, there will be no significantchange in the amounts of radiological effluents released offsite.Based on the above evaluation, the proposed change will not result in asignificant change in the types or significant increase in the amounts of anyeffluent released offsite.3. There is no significant increase in individual or cumulative occupational radiationexposure.The proposed change would not result in any changes to the configuration of thefacility. The proposed changes will revise TS 3/4.7.1.2, "Auxiliary FeedwaterSystem," SR 4.7.1.2.1.b. The proposed change will not cause a change in thelevel of controls or methodology used for the processing of radioactive effluentsor handling of solid radioactive waste, nor will the proposed amendment result inany change in the normal radiation levels in the plant. Therefore, there will be noincrease in individual or cumulative occupational radiation exposure resultingfrom this change.

Serial No. 14-301Docket No. 50-423Attachment 2Attachment 2Marked-Up Technical Specifications PagesDOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 3 Serial No. 14-301Docket No. 50-423Attachment 2Febntary 28, 2007PLANT SYSTEMSAUXILIARY FEEl)WATER SYSTEMFor Information Only -No ChangeLIMITING CONDITION FOR OPERATION3.7. 1.2 At least three independent steam generator auxiliary feedwater pumps and associatedflow paths shall be OPERABLE with:a. Two motor-driven auxiliary feedwater pumps, each capable of being powered fromseparate emergency busses, andb. One steam turbine-driven auxiliary feedwater pump capable of being poweredfrom an OPERABLE steam supply system.APPLICABILITY: MODES 1, 2, and 3.ACTION:Inoperable Equipment Required ACTIONa. Turbine-driven auxiliary a. Restore affected equipment to OPERABLE statusfeedwater pump due to one of the within 7 days. If these ACTIONS are not met, be intwo required steam supplies being at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> andinoperable, in HOT SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.b. b. Restore affected equipment to OPERABLE statuswithin 7 days. If these ACTIONS are not met, be in-NOTE -at least HOT SHUTDOWN within the following 12Only applicable if MODE 2 has hours.not been entered followingREFUELINGOne turbine-driven auxiliaryfeedwater pump in MODE 3following REFUELINGc. One auxiliary feedwater pump in c. Restore the auxiliary feedwater pump toMODE 1, 2, or 3 for reasons other OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If thesethan a. or b. above. ACTIONS are not met, be in at least HOTSTANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOTSHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.d. Two auxiliary feedwater pumps in d. Be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and inMODE 1, 2, or 3. HOT SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.MILLSTONE -UNIT 33/4 7-4Amendment No. , O, 235 Serial No. 14-301Docket No. 50-423Attachment 2Febrma- -9244PLANT SYSTEMSAUXILIARY FEEDWATER SYSTEMLIMITING CONDITION FOR OPERATIONACTION: (Continued)Inoperable Equipment Required ACTIONe. Three auxiliary feedwater e.pumps in MODE 1, 2, or 3.-------NOTE --------LCO 3.0.3 and all other LCO required ACTIONSrequiring MODE changes are suspended until oneAFW pump is restored to OPERABLE status.Immediately initiate ACTION to restore one auxiliaryfeedwater pump to OPERABLE status.SURVEILLANCE REQUIREMENTS4.7.1.2.1 Each auxiliary feedwater pump shall be demonstrated OPERABLE:a. At the frequency specified in the Surveillance Frequency Control Program by:-- ------- --NOTE --------------Auxiliary feedwater pumps may be considered OPERABLE during alignment andoperation for steam generator level control, if they are capable of being manuallyrealigned to the auxiliary feedwater mode of operation.Verifying each auxiliary feedwater manual, power operated, and automatic valve ineach water flow path and in each required steam supply flow path to the steamturbine driven auxiliary feedwater pump, that is not locked, sealed, or otherwisesecured. in position, is in the correct position. Insert A -Page 3/4 7-5,!"D.4ý Veiifying ihmat i reiretlaiaflow efth meterda, -nip deelp -I ataial head of greair that ar eqa 3t3485 feet-,2ý eri6,itig that en reeiraultiaae flew the 9team tur-bina driven pum~pda;vzlapc a total head zf greatezr than or qual to 3780 feet wheit dieseen~my sefnt uppy resureisgreAta dtLm 8tJ0 pi.4te h' isosogrez~ fatfie 4.0. 4 are not m~~iel fgr. ent", into MODE MILLSTONE -UNIT 33/4 7-5Amendment No. 96. 4-O1, 4-2, 1-39,46 235825 Serial No. 14-301Docket No. 50-423Attachment 2Insert A -Page 3/4 75----- ------------- --- -------- ---NLJr ------------- ---------------- ---------------------Not required to be performed for the steam turbine driven auxiliary feedwater pump until24 hours after reaching 800 psig in the steam generators.Verify the developed head of each auxiliary feedwater pump at the flow test point isgreater than or equal to the required developed head when tested pursuant toSpecification 4.0.5. The provisions of Specification 4.0.4 are not applicable to the steamturbine driven auxiliary feedwater pump for entry into MODE 3.

Serial No. 14-301Docket No. 50-423Attachment 2February 25, 2014PLANT SYSTEMS For Information Only -No ChangeAUXILIARY FEEDWATER SYSTEMSURVEILLANCE REQUIREMENTS (Continued)c. At the frequency specified in the Surveillance Frequency Control Program byverifying that each auxiliary feedwater pump starts as designed automatically uponreceipt of an Auxiliary Feedwater Actuation test signal. For the steam turbine-driven auxiliary feedwater pump, the provisions of Specification 4.0.4 are notapplicable for entry into MODE 3.4.7.1.2.2 An auxiliary feedwater flow path to each steam generator shall be demonstratedOPERABLE following each COLD SHUTDOWN of greater than 30 days prior to enteringMODE 2 by verifyhig flow to each steam generator.MILLSTONE -UNIT 33/4 7-5a,Amendment No. 23-5, 258 Serial No. 14-301Docket No. 50-423Attachment 3Attachment 3Marked-Up Technical Specifications Bases Pages for Information OnlyDOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 3 Serial No. 14-301Docket No. 50-423Attachment 3LBDCR 07-MP3-037July 12, 2007PLANT SYSTEMS For Information Only -No ChangeBASES3/4.7.1 TURBINE CYCLE3/4.7. 1.1 SAFETY VALVES (Continuedl)restoration to +/- 1% of the specified lift setting is not required for valves that will not be used (e.g.,replaced) for the next operating cycle. While the lift settings are being restored to within the +/- 1%of the required setting, the MSSVs remain OPERABLE provided the actual lift setting is within+/- 3% of the required setting. The lift settings, according to Table 3.7-3, correspond to ambientconditions of the valve at nominal operating temperature and pressure.This SR is modified by a Note that allows entry into and operation in MODE 3 prior toperforming the SR. The MSSVs may be either bench tested or tested in situ at hot conditionsusing an assist device to sinmulate lift pressure. If the MSSVs are not tested at hot conditions, thelift setting pressure shall be corrected to ambient conditions of the valve at operating temperatureand pressure.REFERENCESI. FSAR, Section 10.3.1.2. ASME, Boiler and Pressure Vessel Code,Section III, 1971 edition.3. FSAR, Section 15.2.4. NRC Information Notice 94-60, "Potential Overpressurization of the Main SteamSystem," August 22, 1994.3/4.7.1.2 AUXILIARY FEEDWATER SYSTEMThe OPERABILITY of the Auxiliary Feedwater (AFW) System ensures a makeup water supplyto the steam generators (SGs) to support decay heat removal from the Reactor Coolant System(RCS) upon the loss of normal feedwater supply, assuming the worst case single failure. TheAFW System consists of two motor driven AFW pumps and one steam turbine driven A.FWpump. Each motor driven AFW pump provides at least 50% of the AFW flow capacity assumedin the accident analysis. After reactor shutdown, decay heat eventually decreases so that onemotor driven AFW pump can provide sufficient SG makeup flow. The steam driven AFW pumphas a rated capacity approximately double that of a motor driven AFW pump and is thus definedas a 100% capacity pump.Given the worst case single failure, the AFW System is designed to mitigate the consequences ofnumerous design basis accidents, including Feedwater Line Break, Loss of Normal Feedwater,Steam Generator Tube Rupture, Main Steam Line Break, and Small Break Loss of CoolantAccident.MILLSTONE -UNIT 3B 3/4 7-2Amendment No. +G20. 4-39, 440, Serial No. 14-301Docket No. 50-423Attachment 3LDC-,R N,. , 4 MP3 01l10, 20"'PLANT SYSTEMSBASESAUXILIARY FEEDWATER SYSTEM (Continued)In addition, given the worst case failure, the AFW is designed to supply sufficient makeupwater to replace SG inventory loss as the RCS is cooled to less than 350'F at which point theResidual Heat Removal System may be placed into operation..u....llL.R..

• q..... 1.7.1.2.1 .. rifi.... .th.t coo ramp's t tI I ft..c.rc.t... l~....tost poi.t is gr. a .or. than or q'ual to tho roquirord total hood. Tphi Ar:- oill.Aiadp.ot11a fteo ,FWA ptillp~ pef"tneeohees .tdgraded duri. :.theo8orftxngoeyolo. Uooouso :4....... hiole to colda.FVint o the...... ... t.... ,n 1 gono.tr .1... they Opeatingtesting is p -t ............ ...... ..latio flow Thi , to.t cord........... A et (dp v .u. nd;...ndieftiN't of .....rol .......o. This t t eoont .f...- oompoAt OPEB.. ILiTY u...sd to tr.nre:f-opm:ttee Tandto d-teatct izciricnt failuros by a pt ornts. oo ootal. hoodspefe inrsoimelallee Requironfont ;71.. doos.,ot Xmoludooz.ta.gin fes ntfLA rfinunoortolintv. TPhis 641 14 tR Add shall ho Ao d at th4A 4 I AFA.0 H.@ SF@6u.F !Vol.Motor driven auxiliary feedwater pumps and associated flow paths are OPERABLE in thefollowing alignment during normal operation below 10%,o RATED THERMAL POWER.* Motor operated isolation valves (3FWA*MOV35A/B/C/D) are open in MODE 1, 2 and 3,* Control valves (3FWA*HV31 A/B/C/D) may be throttled or closed during alignment,operation and restoration of the associated motor driven ATFW pump for steam generatorinventory control.The motor operated isolation valves must remain fully open due to single failure criteria(the valves and associated pump are powered from the opposite electrical trains).The Turbine Driven Auxiliary Feedwater (TDAFW) pump and associated flow paths areOPERABLE with all control and isolation valves hilly open in MODE 1, 2 and 3. Due to HighEnergy Line Break analysis, the TI)AFW pump cannot be used for steam generator inventorycontrol during normal operation below 10% RATED THERMAL POWER.At MPS 3, only two of the three available steam supplies are required to establish anOPERABLE steanm supply system. With one of the two required steam supplies inoperable,nonnally the third steam supply will be used to satisfy the requirement for two OPERABLEsteam supplies. If the third steam supply is also inoperable (i.e., only one steam supply to theturbine-driven auxiliary feedwater pump is OPERABLE), then ACTION a. is entered.If the turbine-driven auxiliary feedwater pump is inoperable due to one required steamsupply being inoperable in MODES 1, 2, and 3, or if a turbine-driven auxiliary feedwater pump isinoperablewhile in MODE 3 immediately following REFUELINQ action must be taken torestore the inoperable equipment to an OPERABLE status within 7 days. The 7 day allowedoutage time is reasonable, based on the following reasons:MILLSTONE -UNIT 3B 3/4 7-2aAmendment No. +0-2. 4-39, 440.

Serial No. 14-301Docket No. 50-423Attachment 3LBDCR No. 04-MP3-011November 10, 2005PLANT SYSTEMS For Information Only -No ChangeBASESAUXILIARY FEEDWATER SYSTEM (Continued)a. For the inoperability of the turbine-driven auxiliary feedwater pump due to onerequired steam supply to the turbine-driven auxiliary feedwater pump beinginoperable (i.e., only one steam supply to the turbine-driven auxiliary feedwaterpump is operable), the 7 day allowed outage time is reasonable since the auxiliaryfeedwater system design affords adequate redundancy for the steam supply line forthe turbine-driven pump.b. For the inoperability of aturbine-driven auxiliary feedwater pump while in MODE3 immediately subsequent to a refueling, the 7 day allowed outage time isreasonable due to the minimal decay heat levels in this situation.c. For both the inoperability of the turbine-driven auxiliary feedwater pump due toone required steam supply to the turbine-driven auxiliary feedwater pump beinginoperable (i.e., only one steam supply to the turbine-driven auxiliary feedwaterpump is operable), and an inoperable turbine-driven auxiliary feedwater pumpwhile in MODE 3 immediately following a refueling outage, the 7 day allowedoutage time is reasonable due to the availability of redundant OPERABLE motordriven auxiliary feedwater pumps, and due to the low probability of an eventrequiring tile use of the turbine-driven auxiliary feedwater pump.The required ACTION dictates that if either tile 7 day allowed outage time is reached theunit must be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWNwithin the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.The allowed time is reasonable, based on operating experience, to reach the requiredconditions from full power conditions in an orderly manner and without challenging plantsystems.A Note limits the applicability of the inoperable equipment condition b. to when the unithas not entered MODE 2 following a REFUELING Required ACTION b. allows one auxiliaryfeedwater pump to be inoperable for 7 days vice the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed outage time in requiredACTION c. This longer allowed outage time is based on the reduced decay heat followingREFUELING and prior to the reactor being critical.With one of the auxiliary feedwater pumps inoperable in MODE 1, 2, or 3 for reasonsother than ACTION a. or b., ACTION must be taken to restore OPERABLE status within 72hours. This includes the loss of three steanm supply lines to the turbine-driven auxiliary feedwaterpump. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed outage time is reasonable, based on redundant capabilities affordedby the auxiliary feedwater system, time needed for repairs, and the low probability of a DBAoccurring during this time period. Two auxiliary feedwater pumps and flow paths remain tosupply feedwater to the steam generators.MILLSTONE -UNIT 3 B 3/4 7-2b Amendment No. 4 -1 4-39, 4--5, Serial No. 14-301Docket No. 50-423Attachment 3LBDC-QR 12 M-P3 gigS.ptem.xcr 20, 2012PLANT SYSTEMSBASESAUXILIARY FEEDWATER SYSTEM (Continued)If all three AFW pumps are inoperable in MODE 1, 2, or 3. the unit is in a seriouslydegraded condition with no safety related means for conducthig a cooldown, and only limitedmeans for conducting a cooldown with non safety related equipment. In such a condition, the unitshould not be perturbed by any action, including a power change, that might result in a trip. Theseriousness of this condition requires that action be started immediately to restore one AFW pumpto OPERABLE status. Required ACTION e. is modified by a Note indicating that all requiredMODE changes or power reductions are suspended until one AFW pump is restored toOPERABLE status. In this case., LCO 3.0.3 is not applicable because it could force the unit into aless safe condition.SR 4.7.1.2.1 a. verifies the correct alignment for manual, power operated, and automaticvalves in the auxiliary feedwater water and steani supply flow paths to provide assurance that theproper flow paths exist for auxiliary feedwater operation, This SR does not apply to valves thatare locked, sealed, or otherwise secured in position, since these valves are verified to be in thecorrect position prior to locking, sealing, or securing. This SR also does not apply to valves thatcannot be inadvertently misaligned, such as check valves. This Surveillance does not require anytesting or valve manipulations; rather, it involves verification that those valves capable ofpotentially being mispositioned are in the correct position. The surveillance frequency iscontrolled under the Surveillance Frequency Control Program.The SR is modified by a Note that states one or more auxiliary feedwater pumps may beconsidered OPERABLE during aligunent and operation for steanm generator level control, if it iscapable of being manually (i.e., remotely or locally, as appropriate) realigned to the auxiliaryfeedwater mode of operation, provided it is not otherwise inoperable. This exception to pumpOPERABILITY allows the pumnp(s) and associated valves to be out of their normal standbyalignment and temporarily incapable of automatic initiation without declaring the pump(s)inoperable. Since auxiliary feedwater may be used during STARTUP, SHUTDOWN, HOTSTANDBY operations, and HOT SHUTDOWN operations for steam generator level control, andthese manual operations are an accepted function of the auxiliary feedwater system,OPERABILITY (i.e., the intended safety function) continues to be maintained.-Insert B -Page B 3/4 7-2cjMILLSTONE -UNIT 3B 3/4 7-2c,ddilm cr, t Ntj.

Serial No. 14-301Docket No. 50-423Attachment 3Insert B -Page B 3/4 7-2cSurveillance Requirement 4.7.1.2.1 .b, which addresses periodic surveillance testing ofthe AFW pumps to detect gross degradation caused by impeller structural damage orother hydraulic component problems, is required by the ASME OM Code. This type oftesting may be accomplished by measuring the pump developed head at only one pointon the pump characteristic curve. This verifies both that the measured performance iswithin an acceptable tolerance of the original pumps baseline performance and that theperformance at the test flow is greater than or equal to the performance assumed in theunit safety analysis. The surveillance requirements are specified in the Inservice TestingProgram, which encompasses the ASME OM Code. The ASME OM Code provides theactivities and frequencies necessary to satisfy the requirements.This surveillance is modified by a note to indicate that the test can be deferred for thesteam driven AFW pump until suitable plant conditions are established. This deferral isrequired because steam pressure is not sufficient to perform the test until after MODE 3is entered. However, the test, if required, must be performed prior to entering MODE 2.Surveillance Requirement 4.7.1.2.1 .c demonstrates that each AFW pump starts onreceipt of an actual or simulated actuation signal. The surveillance frequency iscontrolled under the Surveillance Frequency Control Program. The actuation logic istested as part of the Engineered Safety Feature Actuation System (ESFAS) testing, andequipment performance is monitored as part of the Inservice Testing Program.Surveillance Requirement 4.7.1.2.2 demonstrates the AFW System is properly alignedby verifying the flow path to each steam generator prior to entering MODE 2 after morethan 30 days in any combination of MODE 5 or 6 or defueled. OPERABILITY of theAFW flow paths must be verified before sufficient core heat is generated that wouldrequire operation of the AFW System during a subsequent shutdown. To further ensureAFW System alignment, the OPERABILITY of the flow paths is verified followingextended outages to determine that no misalignment of valves has occurred. Thefrequency is reasonable, based on engineering judgement, and other administrativecontrols to ensure the flow paths are OPERABLE.