Text

Monticello Nuclear Generating Plant - Operating License and Technical Specifications
	ML15110A280
Person / Time
Site:	Monticello
Issue date:	04/15/2015
From:	; Xcel Energy
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	QF0212, Rev. 5
	Download: ML15110A280 (99)
	v • d • e

QF0212, Revision 5 (FP-SC-RSI-04)Page 1 of 1Xcel EnergySHIPPING DOCUMENTNORTHERN STATES POWER -MND/B/A Xcel EnergyMonticello Nuclear Plant, 2807 W Hwy. 75, Monticello, MN 55362Date: 4-14-15Shipping DocumentTracking Number:Ship To:USNRC11555 Rockville PikeRockville, MD 20852-2738Attention Of: Document Control Desk (103)Carrier: UPS -Standard Overnight RMA No:Pro I Tracking No: PO / Contract No:Packaging: Number of Packages: 1 Weight:Dangerous Goods/ UN/NA No: Insurance Est. ValueHazardous Materials? Required?Reason for Shipment: Overnight Shipment to USNRCMelody Imholte -Please ensure tracking number is communicated to me -melody.imholte@xenuclear.comItem No. Qt.. Unit j Description Catalog iDQ1 Box Tech Spec UpdatesRequest arBy signin s s pin d me t you re dec aring, to the best of your knowledge, that the matena b in Wshipped is incompliance with Xcel Energy Corporate Policies. Please print and sign your name legibly.SWIP Making Shipment: Date:Received By: Date:For will-call use onlyUse of this form as a procedural aid does not require retention as a quality record.'4ool&f

~OPERATING LICENSE AND TECHNICAL SPECIFICATIONSUPDATING INSTRUCTIONSTECHNICAL SPECIFICATIONSREMOVE ___SERPage Amendment Document Type N'a Am-en'dment3 186 Operating .... .1871 -License 'Table 1 186 Operating Table I 1ýý7License andTS LEFPTable 2 186 Record of TS -Table.2 187iChanges and .j.,OL Amend.5.5-11 160 Specification ...1'. .1. 7 5.5.11:::, :r ..:, / (Do not insert in TS binder, only an updating aid.)MonticelloAm 187 2. Pursuant to the Act and 10 CFR Part 70, NSPM to receive, possess, anduse at any time special nuclear material as reactor fuel, in accordancewith the limitations for storage and amounts required for reactoroperations, as described in the Final Safety Analysis Report, assupplemented and amended, and the licensee's filings dated August 16,1974 (those portions dealing with handling of reactor fuel);3. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM to receive,possess, and use at any time any byproduct, source and special nuclearmaterial as sealed neutron sources for reactor startup, sealed sources forreactor instrumentation and radiation monitoring equipment calibration,and as fission detectors in amounts as required;4. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM to receive,possess, and use in amounts as required any byproduct, source orspecial nuclear material without restriction to chemical or physical form,for sample analysis or instrument calibration or associated withradioactive apparatus or components; and5. Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to possess, butnot separate, such byproduct and special nuclear material as may beproduced by operation of the facility.C. This renewed operating license shall be deemed to contain and is subject to theconditions specified in the Commission's regulations in 10 CFR Chapter I and issubject to all applicable provisions of the Act and to the rules, regulations, andorders of the Commission, now or hereafter in effect; and is subject to theadditional conditions specified or incorporated below:1. Maximum Power LevelNSPM is authorized to operate the facility at steady state reactor corepower levels not in excess of 2004 megawatts (thermal).2. Technical SpecificationsThe Technical Specifications contained in Appendix A, as revised throughAmendment No. 187, are hereby incorporated in the license. NSPM shalloperate the facility in accordance with the Technical Specifications.3. Physical ProtectionNSPM shall implement and maintain in effect all provisions of theCommission--approved physical security, guard training and qualification,and safeguards contingency plans including amendments made pursuantto provisions of the Miscellaneous Amendments and SearchRenewed License No. DPR-22Amendment No. 1AhF 187 TABLE I (Page I of 3)MONTICELLO NUCLEAR GENERATING PLANTOPERATING LICENSE AND TECHNICAL SPECIFICATIONS LIST OF EFFECTIVE PAGESPage1.1-31.1-21.1-31.1-61.21.1.1-21.2-11.2-21.3-11.3-21.3-31.3-1.3-51.3-81.3-91.3-801.3-911.3-101.3-111.4-11.4-21.4-31.4-6Am. No.(2)(2)(2)146148146176175146146146146146146146146146146146146146146146146146158146146Paae3.1.4-23.1.4-33.1.5-13.1.5-23.1.5-33.1.6-13.1.6-23.1.7-13.1.7-23.1.7-33.1.7-43.1.7-53.1.7-63.1.8-13.1.8-23.2.1-13.2.2-13.2.2-23.2.3-13.3.1.1-13.3.1.1-23.3.1.1-33.3.1.1-43. 3.1.1-53. 3.1.1-63. 3.1.1-73. 3.1.1-83. 3. 1.1-93. 3. 1.1-103.3.1.2-13.3.1.2-23.3.1.2-33.3.1.2-43.3.1.2-53.3.2.1-13.3.2.1-23.3.2.1-33.3.2.1-43.3.2.1-53.3.2.2-13.3.2.2-23.3. 3. 1-13.3.3.1-23.3.3.1-33.3.3.2-13.3.4.1-13.3.4.1-23.3.4.1-3Am. No.146158146146146146146148146146146146146146146146146146146171176180180180180180180180180146146146146146146146159159173146146146146146146146146146Paqge3. 3. 5.1-13.3. 5. 1-23.3.5.1-33.3.5.1-43.3. 5. 1-53.3.5.1-63.3. 5. 1-73.3.5.1-83.3.5.1-93. 3. 5.1-103.3.5.1-113.3.5.2-13.3.5.2-23.3.5.2-33.3.5.2-43. 3.6. 1-13.3.6.1-23.3.6.1-33.3.6.1-43.3.6.1-53.3.6.1-63.3.6.1-73.3.6.2-13.3.6.2-23.3.6.2-33.3.6.3-13.3.6.3-23.3.6.3-33. 3.7. 1-13.3.7.1-23.3.7.1-33.3.7.2-13.3.7.2-23. 3.8. 1-13.3.8.1-23.3.8.1-33.3.8.2-13.3.8.2-23.3.8.2-33.4.1-13.4.1-23.4.2-13.4.3-13.4.3-23.4.4-13.4.4-23.4.5-1Am. No.146146146146151176146176161146146146146146146146146146146176146164146146146146146146Paqe3.4.5-23.4.6-13.4.6-23.4.7-13.4.7-23.4.8-13.4.8-23.4.9-13.4.9-23.4.9-33.4.10-13.5.1-13.5.1-23.5.1-33.5.1-43.5.1-53.5.1-63.5.1-73.5.2-13.5.2-23.5.2-33.5.3-13.5.3-23.6.1.1-13.6.1.1-23.6.1.2-1Am. No.146148148146146146146172172172146146184184184162 (7)1671681461461671461461461461461.4-7 1462.0-1 1853.0-13.0-23.0-33.0-43.0-53.1.1-13.1.1-23.1.1-33.1.2-13.1.2-23.1.3-13.1.3-23.1.3-33.1.3-4.3.1.4-11571461571461461461461461461461461581581581461481481481481481461461471461461461591591461461681461461463.6.1.2-2 1463.6.1.2-33.6.1.2-43.6.1.3-13.6.1.3-23.6.1.3-33.6.1.3-43.6.1.3-53.6.1.3-63.6.1.3-73.6.1.3-83.6.1.4-13.6.1.5-13.6.1.5-23.6.1.6-13.6.1.6-2146146148146148146180148146176146146168146146Am. 187 TABLE 1 (Page 2 of 3)MONTICELLO NUCLEAR GENERATING PLANTOPERATING LICENSE AND TECHNICAL SPECIFICATIONS LIST OF EFFECTIVE PAGES361e3.6.1.7-13.6.1.7-23.6.1.8-13.6.1.8-23.6.2.1-13.6.2.1-23.6.2.1-33.6.2.2-13.6.2.3-13.6. 2. 3-23.6.3.1-13.6.4.1-13.6.4.1-23.6.4.2-13.6.4.2-23.6.4.2-33.6.4.3-13.6.4.3-23.7.1-13.7.1-23.7.2-13.7.2-23.7.3-13.7.4-13.7.4-23.7.4-33.7.5-13.7.5-23.7.6-13.7.6-23.7.7-13.7.8-13.8.1-13.8.1-23.8.1-33.8.1-43.8.1-53.8.1-63.8.1-73.8.1-83.8.1-93.8.1-103.8.2-1Am. No. Paae Am. No. Paae Am. No. OperatingLicense146 3.8.2-2 148 3.10.6-1 146 Cover p.146 3.8.2-3 146 3.10.6-2 146 1146 3.8.3-1 178 3.10.7-1 146 2146 3.8.3-2 178 3.10.7-2 146 3146 3.8.3-3 146 3.10.8-1 146 4146 3.8.4-1 146 3.10.8-2 146 5146 3.8.4-2 153(8) 3.10.8-3 146 61461461461461461461461461461461811461461461461461751601811541541461461461461461461461461461461461461461461483.8.4-33.8.5-13.8.5-23.8.6-13.8.6-23.8.6-33.8.6-43.8.7-13.8.7-23.8.8-13.8.8-23.9.1-13.9.1-23.9.2-13.9.3-13.9.4-13.9.4-23.9.5-13.9.6-13.9.7-13.9.7-23.9.8-13.9.8-23.10.1-13.10.1-23.10.1-33.10.2-13.10.2-23.10.3-13.10.3-23.10.3-33.10.4-13.10.4-23.10.4-33.10.5-13.10.5-2146148146146146146146146146148146146146146146146146146146146146146146174146146146146146146146146146146146146156156187186156176176(6)1761761761691604.0-14.0-25.1-15.2-15.2-25.3-15.4-15.5-15.5-25.5-35.5-45.5-55.5-65.5-75.5-85.5-95.5-105.5-115.5-125.5-135.6-15.6-25.6-35.7-15.7-25.7-35.7-4182182146146163146146146146146146181181181146146176(6)187182182146180180146146146146789101112OL App. ACover p.Op. LicenseAppendix CC-1C-2C-3C-4175102175110Am. 187 TABLE 1 (Page 3 of 3)MONTICELLO NUCLEAR GENERATING PLANTOPERATING LICENSE AND TECHNICAL SPECIFICATIONS LIST OF EFFECTIVE PAGESNOTESNOTE 1: (Removed)NOTE 2: Am. 152 removed Table of Contents (TOC) from NRC issued Tech. Specs. TOC at Revision 0.NOTE 3: (Removed)NOTE 4: (Removed)NOTE 5: (Removed)NOTE 6: NRC Correction Letter to Amendment 176, dated 12/16/2013, corrected spelling of "gage" to,.gauge" in OL License Condition 15(a). Added previously issued Amendment No. 175 (struckthrough) to bottom of page 5.5-10 and removed old revision bar.NOTE 7: SR 3.5.1.3.b (Alternate Nitrogen System supply pressure to ADS) is annotated and is beingtreated as a non-conservative TS. AN2 pressure is increased as specified within Ops. ManualB.08.04.03-05 as an interim measure.NOTE 8: The required 125 VDC charger amperage in SR 3.8.4.2 Option 1 of 50 amps is annotated. Theinterim required amperage is 75 amps. This condition is being treated as a non-conservative TS(see AR 01456839).Am. 187 TABLE 2 (Page 1 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNSP PageRevision(REV) No.LicenseDPR-22Amend No. & DateAEC Tech SpecChanqe IssuanceNo. & DateMaior SubiectOriginal1 1/19/71Note 212345.678910111213141516171819Note 2Note 22 2/20/73Note 2Note 2Note 2Note 2Note 2Note 2Note 2Note 234 6/17/746 8/20/74Note 3578910 7/8/7512111314Note 12 1/14/723 10/31/724 12/8/72Note 15 3/2/731 4/28/71 &6 4/3/737 5/4/738 7/2/739 8/24/7310 10/2/7311 11/27/73&12 11/15/7313 3/30/7414 5/14/74Note 1Note 1Note 3 10/24/7415 1/15/7516 2/3/7517 2/26/7518 4/10/75Note 120 9/15/7519 9/17/7521 10/6/7522 10/30/75Appendix A Technical Specificationsincorporated in DPR-22 on 9/8/70Removed 5 MWt restrictionMOGS Technical Specification changesissued by AEC but never distributed or putinto effect, superseded by TS Change 1211/15/73RHR service water pump capability changeTemporary surveillance test waiverIncrease in U-235 allowed in fission chambersMiscellaneous Technical Specification changes,Respiratory Protection, & AdministrativeControl ChangesRespiratory Protection ChangesRelief Valve and CRD Scram Time ChangesFuel Densification LimitsSafety Valve Setpoint ChangeOffgas Holdup System, RWM, andMiscellaneous Changes8x8 Fuel Load Authorization8x8 Full Power authorizationChanged byproduct material allowanceChanged byproduct material allowanceInverted Tube (CRD) LimitsREMP ChangesReactor Vessel Surveillance Program ChangesVacuum Breaker Test ChangesCorrects Errors & Provides ClarificationIncreased allowed quantity of U-235Snubber RequirementsRemoved byproduct material allowanceSuppression Pool Temperature LimitsAppendix K and GETAB LimitsAm. 187 TABLE 2 (Page 2 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNSP PageRevision(REV) No.2021222324252627282930313233343536373839NOTE 14041424344454647LicenseDPR-22Amend No. & Date15 1/22/76 NOTE 416 2/3/7617 3/16/76NOTE 3 4/13/7618 4/14/7621 5/20/7619 5/27/7620 6/18/7622 7/13/7623 9/27/7624 10/15/7625 10/27/7626 4/1/7727 5/24/7728 6/10/7729 9/16/7730 9/28/7731 10/14/7732 12/9/7733 1/25/7834 4/14/7835 9/15/7836 10/30/7837 11/6/78NOTE 3 11/24/7838 3/15/7939 5/15/7940 6/5/7941 8/29/79Reporting RequirementsCRD Collet Failure SurveillanceNSP Organization ChangesAdoption of GETABContainment Isolation Valve TestingInterim Appendix B, Section 2.4 Tech. Specs.Low Steamline Pressure Setpoint and MCPR ChangesAPLHGR, LHGR, MCPR LimitsCorrection of Errors and Environmental ReportingStandby Gas Treatment System SurveillanceCRD Test FrequencySnubber Testing ChangesAPRS Test MethodMAPLHGR Clamp at Reduced FlowRadiation Protection Supervisor QualificationREMP ChangesMore Restrictive MCPRInservice Inspection ChangesReporting RequirementsFire Protection RequirementsIncrease in spent fuel storage capacityRPT RequirementsSuppression Pool Surveillance8x8R Authorization, MCPR Limits & SRV SetpointsCorrected Downcomer SubmergenceIncorporation of Physical Security Plan into LicenseRevised LPCI Flow CapabilityRespiratory Protection Program ChangesFire Protection Safety Evaluation ReportManor SubjectAm. 187 TABLE 2 (Page 3 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNSP PageRevision(REV) No.LicenseDPR-22Amend No. & DateMaeor Subject4849424312/28/792/12/8050NOTE 1NOTE 15152NOTE 1NOTE 15354NOTE 144 2/29/80-8/29/80-9/19/80-10/24/80-1/9/81-1/9/81-1/13/811 2/12/812 3/2/81MAPLHGR vs. Exposure TableMCPR & MAPLHGR Changes for Cycle 8 and Extended CoreBurnupILRT RequirementsOrder for Modification of License-EnvironmentalQualificationRevised Order for Modification of License-EnvironmentalQualificationOrder for Modification of License-EnvironmentalQualification RecordsIssuance of Facility Operating License (FTOL)Order for Modification of License Concerning BWRScram Discharge Systems (License conditions removed perAmendment No. 11 dated 10/8/82)Order for Modification Mark I ContainmentRevision of License Conditions Relating to FireProtection ModificationsTMI Lessons Learned & Safety -Related Hydraulic Snubber AdditionsLow voltage protection, organization and miscellaneousIncorporation of Safeguards Contingency Plan and SecurityForce Qualification and Training Plan into LicenseCycle 9 -ODYN Changes, New MAPLHGR's, RPSResponse time changeInservice Inspection ProgramFire Protection Technical Specification ChangesMark I Containment Modifications343/27/813/27/81555657585 5/4/816786/3/816/30/8111/5/8159NOTE 19 12/28/81-1/19/82Inservice Surveillance Requirements for SnubbersRevised Order for ModificationMark I ContainmentScram Discharge VolumeNew Scram Discharge Volumes606110115/20/8210/8/82Am. 187 TABLE 2 (Page 4 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNSP PageRevision(REV) No.LicenseDPR-22Amend No. & DateMajor Subject6263646566676869707172737475767778798081828384858687888990121314151617181920212223242526272829303132333435363738394011/30/8212/6/8212/10/8212/17/824/18/834/17/8311/28/8312/30/831/16/841/23/842/2/844/3/845/1/848/15/849/24/8410/31/8411/2/8411/16/8411/16/8411/27/845/28/8510/7/8510/8/8512/3/8512/23/851/22/862/12/863/13/863/18/86RPS Power MonitorCycle 10Recirc Piping and Coolant Leak DetectionAppendix I Technical Specifications (removed App. B)Organizational ChangesMiscellaneous ChangesSteam Line Temperature Switch SetpointRadiation Protection ProgramSRM Count RateDefinition of OperabilityMiscellaneous Technical Specification ChangesRPS Electrical Protection Assembly Time DelayScram Discharge Volume Vent and Drain ValvesMiscellaneous Technical Specification ChangesCycle 11RHR Intertie Line AdditionHybrid I Control Rod AssemblyARTSLow Low Set LogicDegraded Voltage Protection LogicSurveillance RequirementsScreen Wash/Fire Pump (Partial)Fuel Enrichment LimitsCombustible Gas Control SystemVacuum Breaker CyclingNUREG-0737 Technical SpecificationsEnvironmental Technical SpecificationsAdministrative ChangesClarification of Radiation Monitor RequirementsAm. 187 TABLE 2 (Page 5 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNSP Page License Major SubjectRevision DPR-22(REV) No. Amend No. & Date91 41 3/24/86 250 Volt Battery92 42 3/27/86 Jet Pump Surveillance93 43 4/8/86 Simmer Margin Improvement94 44 5/27/86 Cycle 12 Operation95 45 7/1/86 Miscellaneous Changes96 46 7/1/86 LER Reporting and Miscellaneous Changes97 47 10/22/86 Single Loop Operation98 48 12/1/86 Offgas System Trip99 49 8/26/87 Rod Block Monitor100 50 8/26/87 APRM and IRM Scram Requirements101 51 10/16/87 2R Transformer102 52 11/18/87 Surveillance Intervals -ILRT Schedule103 53 11/19/87 Extension of Operating License104 54 11/25/87 Cycle 13 and Misc Changes105 55 11/25/87 Appendix J Testing106 56 12/11/87 ATWS -Enriched Boron107 57 9/23/88 Increased Boron Enrichment108 58 12/13/88 Physical Security Plan109 59 2/16/89 Miscellaneous Administrative Changes110 60 2/28/89 Miscellaneous Administrative Changes111 61 3/29/89 Fire Protection and Detection System112 62 3/31/89 ADS Logic and S/RV Discharge Pipe Pressure113 63 4/18/89 Miscellaneous Technical Specification Improvements114 64 5/10/89 Containment Vent and Purge Valves115 65 5/30/89 NUREG-0737 -Generic Letter 83-36116 66 5/30/89 Reactor Vessel Level Instrumentation117 67 6/19/89 Extension of MAPLHGR. Exposure for One Fuel Type118 68 7/14/89 SRO Requirements & Organization Chart RemovalAm. 187 TABLE 2 (Page 6 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNSP PageRevision(REV) No.LicenseDPR-22Amend No. & DateMaior Subiect119120121122123124125126127128125130131132133134135136137138139140141142143697071727374757677787980818283848586879/12/899/28/8910/19/8911/2/895/1/906/5/9010/12/9012/20/902/15/913/28/914/9/918/12/914/16/927/15/928/18/921/27/936/29/937/12/934/15/94Operations Committee Quorum RequirementsRelocation of Cycle-Specific Thermal-Hydraulic LimitsDeletion of Primary Containment Isolation Valve TableRG 1.99, Rev 2, ISI & ILRTCombined STA/LSO PositionRemoval of WRGM Automatic ESF ActuationDiesel Fuel Oil StorageMiscellaneous Administrative ChangesRedundant and IST TestingAlarming DosimetrySAFER/GESTRTorus Vacuum Breaker Test Switch/EDG Fuel Oil Tank LevelSurveillance Test Interval Extension -Part IAlternate Snubber Visual Inspection IntervalsRevisions to Reactor Protection System Tech SpecsMELLIA and Increase Core FlowRevision to Diesel Fire Pump Fuel Oil Sampling RequirementsRevisions to Control Rod Drive Testing RequirementsRevised Coolant Leakage Monitoring FrequencyAverage Planar Linear Heat Generation Rate (APLHGR)Specification & Minimum Critical Power Ratio Bases RevisionsRemoval of Chlorine Detection Requirements andChanges to Control Room Ventilation System RequirementsRevisions to Radiological Effluent SpecificationsSecondary Containment System and Standby GasTreatment System Water Level Setpoint ChangeChange in Safety Relief Valves Testing RequirementsRevised Core Spray Pump Flow88 6/30/9489 8/25/94909192939/7/949/9/949/15/947/12/95Am. 187 TABLE 2 (Page 7 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNSP PageRevision(REV) No.LicenseDPR-22Amend No. & DateManor Subject1441451461471489495969710/2/954/3/964/9/969/17/9698 7/25/9799 10/29/9711/25/97149NOTE 5150NOTE 61001 00a1014/20/984/30/9808/28/98102 09/16/98103 12/23/98Standby Gas Treatment and Secondary Containment SystemsMSIV Combined Leakrate, and Appendix J, Option BPurge and Vent Valve Seal Replacement IntervalImplementation of BRWOG Option I-D core Stability Solutionand re-issue of pages 11, 12, 82 and 231 to reflect pagesissued by NRC amendments.Bases changes on containment overpressure and numberof RHR pumps required to be operable. Reissue pages 207,209, 219, 229k, 229p, 230, 245 to reflect pages issued byNRC amendments.SLMCPR for Cycle 18 and reissue pages vi, 155, 202, 207,219, 229uReissue pages a, b, g, iii, vi, 14, 25a, 155, 198y, 198z, 202,207, 209, 219, 229k, 229p, 229r, 229u, 230, 245SLMCPR for Cycle 19Reissue all pages.Reactor Coolant Equivalent Radioiodine Concentration andControl Room HabitabilityMonticello Power RerateSurveillance Test Interval/Allowed Outage Time ExtensionProgram -Part 2Revision of Statement on Shift Length & other Misc ChangesCST Low Level HPCI/RCIC Suction TransferRevised RPV-PT Curves & remove SBLC RV setpointReactor Pressure Vessel Hydrostatic and Leakage TestingTesting Requirements for Control Room EFT FiltersSafety Limit Minimum Critical Power Ratio for Cycle 20Transfer of Operating Authority from NSP to NMCTransfer of Operating License from NSP to a New UtilityOperating CompanyEmergency Filtration Train Testing Exceptions andTechnical Specification RevisionsAlternate Shutdown System Operability RequirementsSafety/Relief Valve Bellows Leak Detection SystemTest Frequency10410510610710810911011112/24/9803/19/9910/12/9911/24/9912/8/9902/16/0008/07/0008/18/00112 08/18/00113 10/02/00114 11/30/00Am. 187 TABLE 2 (Page 8 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNSP PageRevision(REV) No.LicenseDPR-22Amend No. & DateMajor Subiect115115a12/21/0002/13/01116 03/01/01117 03/07/01118 03/09/01118a. 05/10/01119119a12004/05/0106/28/0107/24/01121 07/25/01122 08/01/01122a 10/22/01123 10/26/01123a 10/25/01124 10/30/01124a 12/05/01Administrative Controls and Other Miscellaneous ChangesBases Change to Reflect Modification 98Q145 InstalledControl Room Toxic Gas Air SupplyRelocation of Inservice Inspection Requirements to aLicensee ProgramReactor Water Cleanup (RWCU) System AutomaticIsolation and Miscellaneous Instrumentation System ChangesRevision of Standby Liquid Control System SurveillanceRequirementsBases Change -50°F Loop Temperature, Bus Transfer &Rerate CorrectionFire Protection Technical Specification ChangesBases Change -Added information on cooldown rateRelocation of Radiological Effluent Technical Specificationsto a Licensee-Controlled ProgramClarify air ejector offgas activity sample point and operabilityrequirementsRelocation of Inservice Testing Requirements to a Licensee-Controlled ProgramBases Change -Remove scram setpoints sentence andcorrect typoControl Rod Drive and Core Monitoring Technical SpecificationChangesBases Change -Change to reflect new operation ofdrywell to suppression chamber vacuum breaker valveposition indicating lightsRelocation of Technical Specification Administrative ControlsRelated to Quality Assurance PlanBases Change -Change to reflect revised TechnicalSpecification definition of a containment spray/coolingsubsystemSafety Limit Minimum Critical Power Ratio for Cycle 21Elimination of Local Suppression Pool Temperature LimitsBases Change -Change reflects relocation of samplepoint for the offgas radiation monitor125126126a12/06/0101/18/0202/15/02Am. 187 TABLE 2 (Page 9 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNSP PageRevision(REV) No.LicenseDPR-22Amend No. & Date127 05/31/02128 06/11/02128a 07/11/02129 08/27/02129a129b130130a09/12/0209/12/0209/23/0209/26/02Major SubiectMissed Surveillance Requirement TechnicalSpecification Changes (TSTF-358)Changes to the Technical Specifications RevisedReference Point for Reactor Vessel Level Setpoints,Simplification of Safety Limits, and Improvement to the BasesBases Change -Correct Drywell to SuppressionChamber Vacuum Breaker Indicating LightDescriptionRevise Technical Specifications and SurveillanceRequirements Relating to Standby Diesel GeneratorsBases Change -Change to Snubber Operability DescriptionBases Change -Remove Language That ImpliesTrip Settings Can Be Modified By Deviation ValuesContainment Systems Technical Specification ChangesBases Change -HPCI -Change Wording / HPCI & RCIC -Enhance with Wording Consistent with NUREG-1433-Rev 1Update the Multiplier Values for Single Loop OperationAverage Planar Linear Heat Generation Rate (APLHGR)Conversion to Option B for Containment Leak Rate TestingRevision to Pressure-Temperature CurvesBases Change -Adequate Reactor Steam Flow forHPCI/RCIC TestingOne-Time Extension of Containment Integrated Leak-Rate Test IntervalBoiling Water Reactor Vessel and Internals ProjectReactor Pressure Vessel Integrated SurveillanceProgramBases Change -Clarify description of head coolingGroup 2 valvesElimination of Requirements for Post Accident SamplingSystem (TSTF-413)Bases Change -Editorial change to define theabbreviation "EFCV."Drywell Leakage and Sump Monitoring Detection System131 10/02/02132133133a02/04/0302/24/0303/28/03134 03/31/03135 04/22/03135a 04/24/03136 06/17/03136a 09/25/03137 08/21/03Am. 187 TABLE 2 (Page 10 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNSP PageRevision(REV) No.LicenseDPR-22Amend No. & DateMajor Subject137a137b10/09/0310/14/03138 05/21/04138a 06/10/04139 06/02/04140140a11/02/0401/13/05141 01/28/05141 a 02/24/05141b 03/10/05141c 03/10/05142 02/01/05143 09/30/0510/20/05Bases Change -RCS Leakage Requirements for TS 3.6.4.DBases Change -Clarification of system controlboundary for ASDSElimination of Requirements for Hydrogen Recombiners andHydrogen and Oxygen Monitors (TSTF-447)Bases Change -Clarification of Tech Spec Table 4.1.1Manual ScramRevised Analysis of Long-Term Containment Responseand Net Positive Suction Head (Design Bases and USAR change)Revision to Technical Specification Tables 3.2.1 and 3.2.4Bases Change -Removal of Drywell Vent Coolers from3.6/4.6 BasesRevision to Technical Specifications Table 3.2.3 andSection 3.7/4.7Bases Change -Implement Improved BPWS asDescribed in NEDO-33091-ABases Change -Bases Changes for LicenseAmendments 138 and 140Bases Change -Removal of 3% Delta-K from StandbyLiquid Control Bases 3.4.A/4.4.ADeletion of Requirements for Submittal of OccupationalRadiation Reports, Monthly Operating Reports, and Report ofSafety/Relief Valve Failures and Challenges (TSTF-369)Implementation of 24-Month Fuel CyclesChange to Facility Operating License Pursuant toCommission Order EA-03-086 Regarding RevisedDesign Basis Treat (DBT); and Revisions to PhysicalSecurity Plan, Training and Qualification Plan, andSafeguards Contingency PlanSurveillance Test Intervals for various instruments(Second part of 24-month Fuel Cycle amendment.)TS Bases changes to conform with the implementation ofLicense Amendments 143 and 144 (24-Month Fuel Cycles).Alternate Source Term -Fuel Handling Accident (TSTF-51)NOTE 1:144 01/12/06144a 04/05/06145 04/24/06Am. 187 TABLE 2 (Page 11 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNSP Page LicenseRevision DPR-22(REV) No. Amend No. & Date146 06/05/06Correction Letter 10/12/06147 07/05/06NOTE 1: Renewed OL 11/08/06148 12/07/06Appendix J Exemptions149 01/18/07Correction Letter 02/23/07150 03/09/07151 07/20/07Maior SubiectImproved Technical Specifications (TSTF-359, 372, 439,455, 458, 460, 464, 479, 480, 485)Correction of typo in Amendment 146 (page 3.3.5.1-8)Degraded Voltage Allowable Value Change (Second part ofITS -Follow-on ITS Amendment)Renewed Facility Operating LicenseAlternate Source Term -Full ScopeIn conjunction with issuance of Amendment 148, exemptions to10 CFR 50.54(o) and to 10 CFR 50, Appendix J, Option B,Sections III.A and III.B were issued.One-Time Extension of LPCI Loop Select Logic Time DelayRelay Surveillance IntervalCorrection: Remove Am 148 from first page of OL and addedRenewed License No. DPR-22 to footer on page 2.Increase SFP allowed Heat Load and allow installation of 64 cellPaR Fuel Storage Rack (if required) to maintain Full Core Offloadcapability during ISFSI construction.Extend Surveillance Interval from 92-days to 24-months andincrease Allowable Values for LPCI Loop Select LogicTime Delay Relays. (Also, correct typo on page 3.3.5.1-6.)Conforming License Amendment to incorporate the MitigationStrategies Required by Section B.5.b of Commission OrderEA-02-026 and the Radiological Protection Mitigation StrategiesRequired by Commission Order EA-06-137Remove the Table of Contents (TOC) out of the TechnicalSpecifications and place under licensee control. TS TOC initialrevision is Revision 0.Revise Surveillance Requirement 3.8.4.2 to specify that theDivision 2 battery chargers are verified to supply greater than orequal to 110 amps.Add Action Statement for two inoperable Control RoomVentilation subsystems to Specification 3.7.5 (TSTF-477).Revise Surveillance Requirement 3.5.1.3.b to specify that theAlternate Nitrogen System supply pressure to the ADS valves isverified to be greater than or equal to 410 psig.Transfer of Operating License from NMC to NSP -MinnesotaAdd LCO 3.0.9 on unavailable Barriers (TSTF-427).Revise Control Rod notch testing frequency from once per 7 daysto every 31 days (TSTF-475).Power Range Neutron Monitoring SystemAm. 187NOTE 1:08/23/07152 11/08/07153 01/30/08154 01/23/08155 02/21/08NOTE 1: 15615715809/22/0810/22/0811/19/08159 1/30/09 TABLE 2 (Page 12 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNSP PageRevision(REV) No.LicenseDPR-22Amend No. & DateCorrection Letter 3/9/09160 3/17/09161 4/7/09162 7/10/09163 8/19/09164 9/28/09165 5/4/11166 7/29/11167 1/11/12168 7/27/12169 8/27/12Maior SubjectCorrect footer of Specification 3.3.7.2 -AST Amendment 148.Control Room Envelope Habitability (TSTF-448).Revise the Allowable Value and channel calibration frequency forTable 3.3.5.1-1, Function 2.j, Recirculation Riser DifferentialPressure -High (Break Detection).Add new Conditions to Specification 3.5.1 for restoration ofvarious low-pressure ECCS subsystems out-of-service (OOS)combinations (e.g., one low-pressure ECCS division OOS).Deleted paragraph d concerning work hour limitations underSpecification 5.2.2 to align with rule changes under 10 CFR 26,Subpart I (TSTF-511).Corrected Modes in Table 3.3.6.1-1, Function 5.d, RWCU Systemisolation on a SLC System initiation, to match SLC Systemmodes (Specification 3.1.7) after adoption of full-scope AST, i.e.,added Mode 3 to Function 5.dRevise MCPR Safety Limit to 1.15 to reflect reload analyses(which include EPU and MELLLA+ considerations).Add Cyber Security Plan license condition under OL Section 3,Physical Protection. (Operating License change only.)Revise Core Spray flowrate from 2800 gpm to 2835 gpm.Revise surveillance requirements in Specifications 3.4.3, 3.5.1and 3.6.1.5 to remove requirements to lift-test SRVs during plantstartup.Revise licensing basis to reflect removal of the capability toautomatically transfer to the 1AR Transformer as a source ofpower to the essential buses on degraded voltage and insteaddirectly transfer to the EDGs. (Operating License change only.)Revise Table 3.3.5.1-1, Functions 1.e and 2.e, "Reactor SteamDome Pressure Permissive -Bypass Timer (Pump Permissive)",(i.e., 20 minute ADS bypass timer), to remove the lower limit ofthe allowable value.Revise Required Actions table for LCO 3.3.1.1 to provide arestoration period before entering the required actions when theAPRMs are inoperable due to SR 3.3.1.1.2 not met.Revise Specification 3.4.9, add new Specification 5.6.5, and adda new definition to specify the adoption of a PTLR.170 9/7/12171 1/25/13172 1/20/13Am. 187 TABLE 2 (Page 13 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNSP PageRevision(REV) No.LicenseDPR-22Amend No. & Date173 7/15/13174 8/9/13175 8/28/13176 12/9/13Correction Letter 12/16/13(to Amendment 176)177 1/28/14178 1/28/14179 2/28/14180 3/28/14181 5/2/14182 10/24/14183 10/31/14184 11/3/14185 11/25/14186 11/28/14187 1/8/15AEC Tech Spec Maior SubiectChanqe IssuanceNo. & DateAdd footnote reflecting that RWM can be bypassed when theimproved BPWS is used for reactor shutdown (TSTF-476).Add allowance to Specification 3.10.1 LCO to allow for scramtime testing during inservice leak testing and hydrostatic testing(TSTF-484).Miscellaneous Operating License and TS editorial andadministrative changes.Extended Power UprateCorrected spelling of "gage" to "gauge" in OL LicenseCondition 15(a). Added previously issued Amendment No. 175(struckthrough) to bottom of page 5.5-10 and removed oldrevision bar.Modifies the Emergency Plan to revise the EAL for the TurbineBuilding Normal Waste Sump Monitor. (Emergency Plan changeonly.)Revise specification to reflect relocating fuel and lube oil requiredvolumes to the TS Bases and replacing them with duration basedrequirements (TSTF-501).Revises Shutdown Margin (SDM) definition to address advancedfuel designs (TSTF-535).Maximum Extended Load Line Limit Analysis, Plus (MELLLA+)Reduces SBGT and CREF Systems runtime from 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months to15 minutes every 31 days. Removes surveillance requirementsto have heaters operating and TS 5.5.6.e electric heater outputtest requirement from the Ventilation Filter Testing Program(TSTF-522).Revise TS to Support Fuel Storage System ChangesRemove Radwaste Operator as 60-Minute Responder(Emergency Plan change only.)Revise TS 3.5.1 to Remove Condition F allowing two Core Spraysubsystems to be OOS.Reduce the Reactor Steam Dome Pressure Specified in theReactor Core Safety Limits, resolves GE PROF analysis Part 21issue.Revise Cyber Security Implementation Schedule MilestonesRevise TS 5.5.11 to remove not used reduced pressure testingcapability from the TS for the Drywell Airlock doors due to thedoor design.Am. 187 TABLE 2 (Page 14 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNOTES1. License Amendment or Order for Modification of License not affecting Technical Specifications.2. Technical Specification change issued prior to 10 CFR revisions which require issuance of TechnicalSpecification changes as License Amendments.3. Modification to Bases. No Technical Specification change or License Amendment issued.4. Technical Specification change numbers no longer assigned beginning with Amendment 15.5. Pages reissued 11/25/97 to conform with NRC version. Revision number of affected pages not changed.6. All pages reissued using INTERLEAF in different font. Using NRC Amendment Nos. and issue date. ForBases and Table of Contents, spelling errors corrected and editorial corrections made and all AmendmentNos. changed to 100a. For remaining Tech Spec pages, no other changes made and current AmendmentNos. used.Am. 187 Programs and Manuals5.55.5 Programs and Manuals5.5.11 Primary Containment Leakage Rate Testing Program (continued)d. Leakage rate acceptance criteria are:1. Containment leakage rate acceptance criterion is < 1.0 La. During thefirst unit startup following testing in accordance with this program, theleakage rate acceptance criteria are < 0.60 La for the Type B and Ctests and < 0.75 La for Type A tests.2. Air lock testing acceptance criterion is an overall air lock leakage rateof _< 0.05 La when tested at _> Pa.e. The resilient seals of each 18 inch primary containment purge and ventvalve shall be replaced at least once every 9 years. The provisions ofSR 3.0.2 are applicable to this requirement. If a common mode failureattributable to the resilient seals is identified based on the results ofSR 3.6.1.3.11, the resilient seals of all 18 inch primary containment purgeand vent valves shall be replaced.f. The provisions of SR 3.0.3 are applicable to the Primary ContainmentLeakage Rate Testing Program.5.5.12 Battery Monitoring and Maintenance ProgramThis Program provides for battery restoration and maintenance, based on therecommendations of IEEE Standard 450-1995, "IEEE Recommended Practicefor Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries forStationary Applications," or of the battery manufacturer of the following:a. Actions to restore battery cells with float voltage < 2.13 V; andb. Actions to equalize and test battery cells that had been discovered withelectrolyte level below the minimum established design limit.Monticello5.5-11Amendment No. 446,,48, 60, 187 TECHNICAL SPECIFICATION BASESUPDATING INSTRUCTIONSMNGP TECHNICAL SPECIFICATION BASESREMOVE !IN'SERTPage(s) Revision Document Type ýPaqe(s) RevisionTable 1 33 List of Effective Table 1 34Sections /SpecificationsTable 2 33 Record of Table 2 34RevisionsB 2.1.1-1 through 4 Bases B 2.1 .1-1 thirough 34B 2.1.1-4 Spec. 2.1.1 13 2.1.1-4B 3.3.6.1-1 through 4 Bases :B 3.3.6.1-1 through 34B 3.3.6.1-23 Spec. 3.3.6.1 B 3.3.6.1-24B 3.5.1-1 through 29 Bases B 3.5.1-1 through 34B 3.5.1-19 Spec. 3.5.1 ;B 3.5.1-20Destroy removed pages.Revision 34 TABLE 1 (Page 1 of 1)MONTICELLO NUCLEAR GENERATING PLANTBASES LIST OF EFFECTIVE SECTIONS/SPECIFICATIONSSection/Specification Revision No.B 2.1.1 34B 2.1.2 6B 3.0 27B 3.1.1 0B 3.1.2 0B 3.1.3 11B 3.1.4 19B 3.1.5 0B 3.1.6 29B 3.1.7 4B 3.1.8 4B 3.2.1 29B 3.2.2 0B 3.2.3 0B 3.3.1.1 32B 3.3.1.2 0B 3.3.2.1 26B 3.3.2.2 29B 3.3.3.1 3B 3.3.3.2 3B 3.3.4.1 0B 3.3.5.1 29B 3.3.5.2 0B 3.3.6.1 34B 3.3.6.2 0B 3.3.6.3 3B 3.3.7.1 4B 3.3.7.2 4B 3.3.8.1 22B 3.3.8.2 0B 3.4.1 32B 3.4.2 0B 3.4.3 25B 3.4.4 0B 3.4.5 0B 3.4.6 4B 3.4.7 0B 3.4.8 0B 3.4.9 25B 3.4.10 0B 3.5.1 34B 3.5.2 0B 3.5.3 0B 3.6.1.1 29B 3.6.1.2 29B 3.6.1.3 29Section/SpecificationB 3.6.1.4B 3.6.1.5B 3.6.1.6B 3.6.1.7B 3.6.1.8B 3.6.2.1B 3.6.2.2B 3.6.2.3B 3.6.3.1B 3.6.4.1B 3.6.4.2B 3.6.4.3B 3.7.1B 3.7.2B 3.7.3B 3.7.4B 3.7.5B 3.7.6B 3.7.7B 3.7.8B 3.8.1B 3.8.2B 3.8.3B 3.8.4B 3.8.5B 3.8.6B 3.8.7B 3.8.8Revision No.025002930000003129017318429033273374004B 3.9.1B 3.9.2B 3.9.3B 3.9.4B 3.9.5B 3.9.6B 3.9.7B 3.9.8B 3.10.1B 3.10.2B 3.10.3B 3.10.4B 3.10.5B 3.10.6B 3.10.7B 3.10.8000000210280300000Rev. 34 TABLE 2 (Page 1 of 6)TECHNICAL SPECIFICATION BASES RECORD OF REVISIONSRevision AffectedNumber Section/Specification0123AllB 3.8.3B 3.5.14B 3.3.3.1,B 3.3.3.2,B 3.3.6.3,B 3.10.3B 2.1.1,B 2.1.2,B 3.1.6,B 3.1.7,B 3.1.8,B 3.3.6.1,B 3.3.7.1,B 3.3.7.2,B 3.4.6,B 3.6.1.3,B 3.7.4,B 3.7.5,B 3.7.6,B 3.8.2,B 3.8.5,B 3.8.8B 3.3.5.1B 3.8.1(1)B 2.1.2B 3.3.2.1B.3.8.1,B 3.8.2Description of RevisionAmendment 146 -Original ITS RevisionSR 3.8.3.3, Diesel Fuel Oil Testing DescriptionLCO 3.5.1, ACTION D, changed description ofLPCI injection pathway.Miscellaneous ITS BasesClarifications/CorrectionsAmendment 148 -Bases Changesimplementing Full Scope AST.Amendment 151 -Extend Surveillance Intervaland AV for the LPCI Loop Select TD Relays.Clarify RCS Safety Limit values.Correct that initial MCPR values are specified inthe COLR.Clarify that the 2R and 1AR transformers areconsidered as a single off-site source when1AR is supplied from 345 kV Bus 1.5671. Replaces page B 3.8.1-25 in Sharepoint version of the TS. Page inadvertentlydeleted during implementation of Amendment 148 (CAP 01095053).Rev. 34 TABLE 2 (Page 2 of 6)TECHNICAL SPECIFICATION BASES RECORD OF REVISIONSRevision Affected Description of RevisionNumber Section/Specification7 (con't) B.3.8.4 Correct the float voltage for the 125 VDCbatteries in SR 3.8.4.1.B.3.8.4 Amendment 153 -Specify in SR 3.8.4.2 thatthe Division 2 battery charger supplies-> 110 amps.8 B.3.5.1 Amendment 155 -Revise SR 3.5.1.3 to correctAlternate Nitrogen System supply pressureto ADS and clarify OPERABILITY during bottlechangeout.B.3.7.4, Amendment 154 -Revise Bases forB.3.7.5 Specification 3.7.5 to reflect adoption ofTSTF-477, which allows both CRV subsystemsto be inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Clarify theOPERABILITY requirements of certain CRVfans currently required to support CREFsubsystem operation.9 B 3.5.1 Clarify RHR intertie discussion.B 3.5.1 Clarify Action M to indicate that the plant maynot be in a condition outside the accidentanalysis but is in a condition not specificallyjustified for continued operation.B 3.6.1.3 Add Action E to describe actions for when theMSIVs are not within leakage limits and re-labelsubsequent actions.10 B 3.5.1 Add HPCI "Keep-fill" discussion.B 3.9.7 Clarify Action A.1 for what is meant byinoperable.11 B 3.1.3, Amendment 158- Change control rod notchB 3.1.4 testing frequency from every 7 days to onlyonce per 31 days in accordance with TSTF-475,Revision 1.Rev. 34 TABLE 2 (Page 3 of 6)TECHNICAL SPECIFICATION BASES RECORD OF REVISIONSRevision Affected Description of RevisionNumber Section/Specification12 B 3.3.1.1, Amendment 159 -PRNMS.B 3.3.2.1,B 3.4.1B 3.3.5.1 Amendment 161 -LPCI Recirculation RiserDifferential Pressure -High (Break Size)allowable value and channel calibration intervalchange.13 B 3.0 Amendment 157 -Add Bases for new LCO3.0.9 for the unavailability of barriers, reflectingadoption of TSTF-427.B 3.4.9 Clarify that the shift in Figure 3.4.9-1 includesboth delta RTNDT and margin.B 3.5.1 Amendment 162 -Add new Conditions toSpecification 3.5.1 for restoration of variouslow-pressure ECCS subsystem out-of-servicecombinations.14 B 3.0 Section B 3.0 reissued in entirety. Pagenumbers at end of LCO Applicabilityover-lapped SR Applicability page numbers(CAP 01192534).15 B 3.7.4 Amendment 160 -Revise Bases for thespecification reflecting adoption of a ControlRoom Envelope Habitability program inaccordance with TSTF-448.16 B 3.3.1.1 Replace IRM -Neutron Flux -High High (1 .a)for calibrating IRMs by a heat balance byreferring to IRM/APRM overlap and APRMSetdown Scram meeting reactivityrequirements.B 3.7.7 Correct Turbine Bypass Valve capacity.17 B 3.7.3 Correct EDG-ESW Background description.Rev. 34 TABLE 2 (Page 4 of 6)TECHNICAL SPECIFICATION BASES RECORD OF REVISIONSRevision Affected Description of RevisionNumber Section/Specification18 B 3.6.1.3 Clarify PCIV definition.19 B 3.1.4 Clarify spacing requirements of adjacent slowcontrol rods.B 3.8.1, Revise Bases to reflect separation of 1ARSB 3.8.2 Transformer and Bus 1.20 B 3.5.1 Reissue section, missing text on HPCI keep-filldiscussion. (CAP 01328422)21 B 3.9.7 Correct prior clarification to Action A.1 for whatis meant by inoperable. (CAP 01257096)22 B 3.3.8.1 Amendment 169 -Revise licensing basis toreflect removal of the capability to automaticallytransfer to the 1AR Transformer as a source ofpower to the essential buses on degradedvoltage and instead directly transfer to theEDGs.23 B 3.3.5.1 Amendment 170 -Revised to reflect ancillarychange related to ADS 20-minute BypassTimer.24 B 3.3.1.1 Amendment 171 -Revised to providerestoration period before declaring the APRMs inoperable when SR 3.3.1.1.2 is not met.25 B 3.4.3, Amendment 168 -Revised surveillanceB 3.5.1, requirements within these specifications toB 3.6.1.5 allow crediting overlapping testing rather thanrequiring a lift-test during plant startup.B 3.4.9 Amendment 172 -Revised specification toadopt PTLR.26 B 3.1.6 Amendment 173 -Revised to reflectB 3.3.2.1 incorporation of TSTF-476 for improved BPWS.Rev. 34 TABLE 2 (Page 5 of 6)TECHNICAL SPECIFICATION BASES RECORD OF REVISIONSRevisionNumber272829AffectedSection/SpecificationB 3.0B 3.8.2B 3.10.1B 3.1.6B 3.2.1B 3.3.1.1B 3.3.2.2B 3.3.5.1B 3.4.1B 3.5.1B 3.6.1.1B 3.6.1.2B 3.6.1.3B 3.6.1.8B 3.7.1B 3.7.7B 3.6.2.1B 3.8.3B 3.6.4.3B 3.7.4Description of RevisionClarify application of SR 3.0.2 and SR 3.0.3 toIST tests to clarify compliance to EnforcementGuidance Memorandum (EGM)12-001(CAP 01389604).Correct referenced SR number from SR 3.8.1.8to SR 3.8.1.6.Amendment 174 -Revised specification toincorporate TSTF-484, to allow scram timetesting in conjunction with hydrostatic testing.Amendment 177- Implement Extended PowerUprate.Revised TS Bases to indicate local SuppressionPool temperature limits were eliminated withAmendment 126.Amendment 178 -Revised specification torelocate stored fuel and lube oil volumes to TSbases and replace with duration requirements inaccordance with TSTF-501.Amendment 181 -Revised specifications toreflect reduced runtime for SBGT and CREFfrom 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months to 15 minutes in accordance withTSTF-522.3031Rev. 34 TABLE 2 (Page 6 of 6)TECHNICAL SPECIFICATION BASES RECORD OF REVISIONSRevision Affected Description of RevisionNumber Section/Specification32 B 3.3.1.1, Amendment 180 -Implement MaximumB 3.4.1 Extended Load Line Limit Analysis, Plus(MELLLA+)33 B 3.8.1, Add EDG Fuel Oil Transfer System trainB 3.8.3 description for each EDG.34 B 2.1.1, Amendment 185 -Reduce the Reactor SteamB 3.3.6.1 Dome Pressure specified in the Reactor CoreSafety Limits, resolved GE Part 21 condition fora potential to violate the safety limit during aPressure Regulator Failure Downscale event.B 3.5.1 Amendment 184 -Removes former Condition Fwhich allowed both Core Spray subsystems tobe inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .Rev. 34 Reactor Core SLsB 2.1.1B 2.0 SAFETY LIMITS (SLs)B 2.1.1 Reactor Core SLsBASESBACKGROUND USAR Section 1.2.2 (Ref. 1) requires the reactor core and associatedsystems to be designed to accommodate plant operational transients ormaneuvers that might be expected without compromising safety andwithout fuel damage. Therefore, SLs ensure that specified acceptablefuel design limits are not exceeded during steady state operation, normaloperational transients, and anticipated operational occurrences (AOOs).The fuel cladding integrity SL is set such that no significant fuel damageis calculated to occur if the limit is not violated. Because fuel damage isnot directly observable, a stepback approach is used to establish an SL,such that the MCPR is not less than the limit specified inSpecification 2.1.1.2. MCPR greater than the specified limit represents aconservative margin relative to the conditions required to maintain fuelcladding integrity.The fuel cladding is one of the physical barriers that separate theradioactive materials from the environs. The integrity of this claddingbarrier is related to its relative freedom from perforations or cracking.Although some corrosion or use related cracking may occur during the lifeof the cladding, fission product migration from this source is incrementallycumulative and continuously measurable. Fuel cladding perforations,however, can result from thermal stresses, which occur from reactoroperation significantly above design conditions.While fission product migration from cladding perforation is just asmeasurable as that from use related cracking, the thermally causedcladding perforations signal a threshold beyond which still greater thermalstresses may cause gross, rather than incremental, claddingdeterioration. Therefore, the fuel cladding SL is defined with a margin tothe conditions that would produce onset of transition boiling (i.e.,MCPR = 1.00). These conditions represent a significant departure fromthe condition intended by design for planned operation. The MCPR fuelcladding integrity SL ensures that during normal operation and duringAOOs, at least 99.9% of the fuel rods in the core do not experiencetransition boiling.Operation above the boundary of the nucleate boiling regime could resultin excessive cladding temperature because of the onset of transitionboiling and the resultant sharp reduction in heat transfer coefficient.Inside the steam film, high cladding temperatures are reached, and acladding water (zirconium water) reaction may take place. This chemicalMonticello B 2.1.1-1 Revision No. 34MonticelloB 2.1.1-1Revision No. 34 Reactor Core SLsB 2.1.1BASESBACKGROUND (continued)reaction results in oxidation of the fuel cladding to a structurally weakerform. This weaker form may lose its integrity, resulting in an uncontrolledrelease of activity to the reactor coolant.The reactor vessel water level SL ensures that adequate core coolingcapability is maintained during all MODES of reactor operation.Establishment of Emergency Core Cooling System initiation setpointshigher than this SL provides margin such that the SL will not be reachedor exceeded.APPLICABLE The fuel cladding must not sustain damage as a result of normalSAFETY operation and AOOs. The reactor core SLs are established to precludeANALYSES violation of the fuel design criterion that a MCPR limit is to be established,such that at least 99.9% of the fuel rods in the core would not beexpected to experience the onset of transition boiling.The Reactor Protection System setpoints (LCO 3.3.1.1, "ReactorProtection System (RPS) Instrumentation"), in combination with the otherLCOs, are designed to prevent any anticipated combination of transientconditions for Reactor Coolant System water level, pressure, andTHERMAL POWER level that would result in reaching the MCPR SafetyLimit.The approved pressure range (700 to 1400 psia) of the GEXL 14 criticalpower correlation is applied to resolve a 10 CFR Part 21 conditionconcerning a potential to violate Reactor Core Safety Limit 2.1.1.1 duringa Pressure Regulator Failure Maximum Demand (Open) transient(Reference 5). Application of this correlation, which applies to the GE14fuel in the core, allows reduction of the reactor steam dome pressure from785 to 686 psig, precluding violation of the safety limit for this event. Thischange in reactor steam dome pressure was approved in Amendment185 (Reference 7).2.1.1.1 Fuel Cladding Inte.grityThe GEXL14 critical power correlation is applicable for all critical powercalculations at pressures > 686 psig and core flows > 10% of rated flow(Reference 6). For operation at low pressures or low flows, another basisis used, as follows:Since the pressure drop in the bypass region is essentially allelevation head, the core pressure drop at low power and flows willalways be > 4.56 psi. Analyses (Ref. 2) show that with a bundle flowof 28 x 103 lb/hr, bundle pressure drop is nearly independent ofbundle power and has a value of 3.5 psi. Thus, the bundle flow witha 4.56 psi driving head will be > 28 x 103 lb/hr. Full scale ATLAS testMonticelloB 2.1.1-2Revision No. 34 Reactor Core SLsB 2.1.1BASESAPPLICABLE SAFETY ANALYSES (continued)data taken at pressures from 0 psig to 785 psig indicate that the fuelassembly critical power at this flow is approximately 3.35 MWt. Withthe design peaking factors, this corresponds to a THERMAL POWER> 50 % RTP. Thus, a THERMAL POWER limit of 25% RTP forreactor pressure < 686 psig or < 10% core flow is conservative.2.1.1.2 MCPRThe fuel cladding integrity SL is set such that no significant fuel damageis calculated to occur if the limit is not violated. Since the parameters thatresult in fuel damage are not directly observable during reactor operation,the thermal and hydraulic conditions that result in the onset of transitionboiling have been used to mark the beginning of the region in which fueldamage could occur. Although it is recognized that the onset of transitionboiling would not result in damage to BWR fuel rods, the critical power atwhich boiling transition is calculated to occur has been adopted as aconvenient limit. However, the uncertainties in monitoring the coreoperating state and in the procedures used to calculate the critical powerresult in an uncertainty in the value of the critical power. Therefore, thefuel cladding integrity SL is defined as the critical power ratio in thelimiting fuel assembly for which more than 99.9% of the fuel rods in thecore are expected to avoid boiling transition, considering the powerdistribution within the core and all uncertainties.The MCPR SL is determined using a statistical model that combines allthe uncertainties in operating parameters and the procedures used tocalculate critical power. The probability of the occurrence of boilingtransition is determined using the approved General Electric CriticalPower correlations. Details of the fuel cladding integrity SL calculationare given in Reference 2. Reference 3 includes a tabulation of theuncertainties used in the determination of the MCPR SL and of thenominal values of the parameters used in the MCPR SL statisticalanalysis.2.1.1.3 Reactor Vessel Water LevelDuring MODES 1 and 2 the reactor vessel water level is required to beabove the top of the active irradiated fuel to provide core coolingcapability. With fuel in the reactor vessel during periods when the reactoris shut down, consideration must be given to water level requirementsdue to the effect of decay heat. If the water level should drop below thetop of the active irradiated fuel during this period, the ability to removedecay heat is reduced. This reduction in cooling capability could lead toelevated cladding temperatures and clad perforation in the event that thewater level becomes < 2/3 of the core height. The reactor vessel waterlevel SL has been established at the top of the active irradiated fuel toMonticelloB 2.1.1-3Revision No. 34 Reactor Core SLsB 2.1.1BASESAPPLICABLE SAFETY ANALYSES (continued)provide a point that can be monitored and to also provide adequatemargin for effective action.SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuelclad barrier to prevent the release of radioactive materials to the environs.SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fueldesign criteria. SL 2.1.1.3 ensures that the reactor vessel water level isgreater than the top of the active irradiated fuel in order to preventelevated clad temperatures and resultant clad perforations.APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.41.3 are applicable in all MODES.SAFETY LIMIT Exceeding an SL may cause fuel damage and create a potential forVIOLATIONS radioactive releases in excess of 10 CFR 50.67, "Accident source term,"limits (Ref. 4). Therefore, it is required to insert all insertable control rodsand restore compliance with the SLs within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The 2 hourCompletion Time ensures that the operators take prompt remedial actionand also ensures that the probability of an accident occurring during thisperiod is minimal.REFERENCES 1. USAR, Section 1.2.2.2. NEDE-2401 1-P-A, "General Electric Standard Application for ReactorFuel" (revision specified in Specification 5.6.3).3. NEDE-31152P, "General Electric Fuel Bundle Designs," Revision 8,April 2001.4. 10 CFR 50.67.5. GE Part 21 Notification SC05-03, "Potential to Exceed Low PressureTechnical Specification Safety Limit," dated March 29, 2005.6. NRC Letter to A. Lingenfelter (GNF), 'Final Safety Evaluation forGlobal Nuclear Fuel (GNF) Topical Report (TR) NEDC-32851P,Revision 2, "GEXL14 Correlation for GE14 Fuel," (TAC No. MD5486)'dated August 3, 2007.7. Amendment No. 185, "Issuance of Amendment to Reduce theReactor Steam Dome Pressure Specified in the Reactor Core SafetyLimits," dated November 25, 2014. (ADAMS AccessionNo. ML14281A318)MonticelloB 2.1.1-4 -LastRevision No. 34 Primary Containment Isolation InstrumentationB 3.3.6.1B 3.3 INSTRUMENTATIONB 3.3.6.1 Primary Containment Isolation InstrumentationBASESBACKGROUND The primary containment isolation instrumentation automatically initiatesclosure of appropriate primary containment isolation valves (PCIVs). Thefunction of the PCIVs, in combination with other accident mitigationsystems, is to limit fission product release during and following postulatedDesign Basis Accidents (DBAs). Primary containment isolation within thetime limits specified for those isolation valves designed to closeautomatically ensures that the release of radioactive material to theenvironment will be consistent with the assumptions used in the analysesfor a DBA.The isolation instrumentation includes the sensors, relays, and switchesthat are necessary to cause initiation of primary containment and reactorcoolant pressure boundary (RCPB) isolation. Most channels includeelectronic equipment (e.g., trip units) that compares measured inputsignals with pre-established setpoints. When the setpoint is exceeded,the channel output relay actuates, which then outputs a primarycontainment isolation signal to the isolation logic. Functional diversity isprovided by monitoring a wide range of independent parameters. Theinput parameters to the isolation logics are (a) reactor vessel water level,(b) area ambient temperatures, (c) main steam line (MSL) flowmeasurement, (d) Standby Liquid Control (SLC) System initiation,(e) main steam line pressure, (f) high pressure coolant injection (HPCI)and reactor core isolation cooling (RCIC) steam line flow, (g) drywellpressure, (h) HPCI and RCIC steam line pressure, (i) reactor watercleanup (RWCU) flow, and (j) reactor steam dome pressure. Redundantsensor input signals from each parameter are provided for initiation ofisolation. The only exception is SLC System initiation.Primary containment isolation instrumentation has inputs to the trip logicof the isolation functions listed below.1. Main Steam Line IsolationReactor Vessel Water Level -Low Low and Main Steam Line Pressure -Low Functions receive inputs from four channels. One channelassociated with each Function inputs to one of four trip strings. Two tripstrings make up a trip system and both trip systems must trip to cause anisolation of all main steam isolation valves (MSIVs), MSL drain valves,and reactor sample isolation valves. Any channel will trip the associatedtrip string. Only one trip string must trip to trip the associated trip system.The trip strings are arranged in a one-out-of-two taken twice logic toinitiate isolation of all main steam isolation valves (MSIVs), MSL drainvalves, and recirculation sample isolation valves.MonticelloB 3.3.6.1 -1Revision No. 34 Primary Containment Isolation InstrumentationB 3.3.6.1BASESBACKGROUND (continued)The Main Steam Line Flow -High Function uses 16 flow channels, fourfor each steam line. One channel from each steam line inputs to one ofthe four trip strings. Two trip strings make up each trip system and bothtrip systems must trip to cause an isolation of the MSIVs, MSL drainvalves, and reactor sample isolation valves. Each trip string has fourinputs (one per MSL), any one of which will trip the trip string. The tripstrings are arranged in a one-out-of-two taken twice logic. This iseffectively a one-out-of-eight taken twice logic arrangement to initiateisolation.The Main Steam Line Tunnel Temperature -High Function receives inputfrom 16 channels (four from each of the four tunnel areas). The logic isarranged similar to the Main Steam Line Flow -High Function. Onechannel from each steam tunnel area inputs to one of four trip strings.Two trip strings make up a trip system, and both trip systems must trip tocause isolation.MSL Isolation Functions isolate the Group 1 valves.2. Primary Containment IsolationThe Reactor Vessel Water Level -Low and Drywell Pressure -HighFunctions receive inputs from four channels. One channel associatedwith each Function inputs to one of four trip strings. Two trip strings makeup a trip system and both trip systems must trip to cause an isolation ofthe Group 2 primary containment isolation valves (i.e., drywell and sump).Any channel will trip the associated trip string. Only one trip string musttrip to trip the associated trip system. The trip strings are arranged in aone-out-of-two taken twice logic to initiate isolation.Primary Containment Isolation Drywell Pressure -High and ReactorVessel Water Level -Low Functions isolate the Group 2 drywell andsump isolation valves.3, 4. High Pressure Coolant Iniection System Isolation and Reactor CoreIsolation Cooling System IsolationThe HPCI and RCIC Steam Line Flow -High Functions receive input fromtwo channels for each system. Each channel output for each system isconnected to a time delay relay that provides an output signal to two tripsystems. The output signal is arranged so that any channel that trips willprovide a trip signal to the trip system (one-out-of-two logic in each tripsystem). Each trip system associated with HPCI or RCIC will provide aclosure signal to the associated system isolation valves. The HPCIMonticello B 3.3.6.1-2 Revision No. 34MonticelloB 3.3.6.1-2Revision No. 34 Primary Containment Isolation InstrumentationB 3.3.6.1BASESBACKGROUND (continued)Steam Supply Line Pressure -Low Function receives input from fourchannels. The outputs are arranged in a one-out-of-two-twice logic in onetrip system. The trip system isolates all HPCI isolation valves. The RCICSteam Supply Line Pressure -Low Function receives input from fourchannels. The outputs are arranged in a one-out-of-two twice logic. Theoutput of the logic is directed to two trip systems. Each trip system isable, by itself, to isolate all RCIC isolation valves. The HPCI and RCICSteam Line Area Temperature -High Functions receive input from16 channels for each system. The outputs of the 16 channels aregrouped in four sets of four detectors. Each set is arranged in one-out-two-twice logic. The outputs of each set provide trip signals to each oftwo separate isolation trip systems. Each trip system is able, by itself, toisolate all HPCI and RCIC isolation valves, as applicable.HPCI Functions isolate the Group 4 valves and RCIC Functions isolatethe Group 5 valves.5. Reactor Water Cleanup System IsolationThe RWCU Room Temperature -High, Reactor Vessel Water Level -LowLow, Drywell Pressure -High, and RWCU Flow -High Functions receiveinputs from four channels. One channel associated with each Functioninputs to one of four trip strings. Two trip strings make up a trip systemand both trip systems must trip to cause an isolation of the RWCU valves.Any channel will trip the associated trip string. Only one trip string musttrip to trip the associated trip system. The trip strings are arranged in aone-out-of-two taken twice logic to initiate isolation of all RWCU isolationvalves. The SLC System Initiation Function receives input from the SLCinitiation switch. The switch provides trip signal inputs to both tripsystems in any position other than "OFF." For the purpose of thisSpecification, the SLC initiation switch is considered to provide onechannel input into each trip system. Each of the two trip systems isconnected to one of the two valves on each RWCU penetration.RWCU Functions isolate the Group 3 valves.6. Shutdown Coolinq System IsolationThe Reactor Vessel Water Level -Low Function receives input from fourreactor vessel water level channels. One channel associated with eachFunction inputs to one of four trip strings. Two trip strings make up a tripsystem and both trip systems must trip to cause an isolation of the RHRMonticello B 3.3.6.1-3 Revision No. 34MonticelloB 3.3.6.1-3Revision No. 34 Primary Containment Isolation InstrumentationB 3.3.6.1BASESBACKGROUND (continued)shutdown cooling supply isolation valves. Any channel will trip theassociated trip string. Only one trip string must trip to trip the associatedtrip system. The trip strings are arranged in a one-out-of-two taken twicelogic to initiate isolation of the RHR shutdown cooling supply isolationvalves. The Reactor Steam Dome Pressure -High Function receivesinput from two channels, both of which provide input to two trip systems.Any trip channel will trip both trip systems to initiate isolation of the RHRshutdown cooling supply isolation valves.Shutdown Cooling System Isolation Functions isolate the Group 2 RHRshutdown cooling supply isolation valves.7. Traversing Incore Probe (TIP) System IsolationThe Reactor Vessel Water Level -Low and Drywell Pressure -HighFunctions receive inputs from four channels. One channel associatedwith each Function inputs to one of four trip strings. Two trip strings makeup a trip system and both trip systems must trip to initiate a TIP driveisolation signal. Any channel will trip the associated trip string. Only onetrip string must trip to trip the associated trip system. The trip strings arearranged in a one-out-of-two taken twice logic to initiate a TIP driveisolation signal.When either Function actuates, the TIP drive mechanisms will withdrawthe TIPs, if inserted, and close the inboard TIP System isolation ballvalves when the TIPs are fully withdrawn. The outboard TIP Systemisolation valves are manual shear valves.TIP System Isolation Functions isolate the Group 2 valves (TIP inboardisolation ball valves).APPLICABLE The isolation signals generated by the primary containment isolationSAFETY instrumentation are implicitly assumed in the safety analyses ofANALYSES, LCO, References 1 and 2 to initiate closure of valves to limit offsite doses.and APPLICABILITY Refer to LCO 3.6.1.3, "Primary Containment Isolation Valves (PCIVs),"Applicable Safety Analyses Bases for more detail of the safety analyses.Primary containment isolation instrumentation satisfies Criterion 3 of10 CFR 50.36(c)(2)(ii). Certain instrumentation Functions are retained forother reasons and are described below in the individual Functionsdiscussion.Monticello B 3.3.6.1-4 Revision No. 34MonticelloB 3.3.6.1-4Revision No. 34 Primary Containment Isolation InstrumentationB 3.3.6.1BASESAPPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)The OPERABILITY of the primary containment instrumentation isdependent on the OPERABILITY of the individual instrumentationchannel Functions specified in Table 3.3.6.1-1. Each Function must havea required number of OPERABLE channels, with their setpoints within thespecified Allowable Values, where appropriate. The actual setpoint iscalibrated consistent with applicable setpoint methodology assumptions.Allowable Values are specified for each Primary Containment IsolationFunction specified in the Table. Nominal trip setpoints are specified in thesetpoint calculations. The nominal setpoints are selected to ensure thatthe setpoints do not exceed the Allowable Value between CHANNELCALIBRATIONS. Operation with a trip setpoint less conservative thanthe nominal trip setpoint, but within its Allowable Value, is acceptable. Achannel is inoperable if its actual trip setpoint is not within its requiredAllowable Value. Trip setpoints are those predetermined values of outputat which an action should take place. The setpoints are compared to theactual process parameter (e.g., reactor vessel water level), and when themeasured output value of the process parameter exceeds the setpoint,the associated device (e.g., trip unit) changes state. The analytic limitsare derived from the limiting values of the process parameters obtainedfrom the safety analysis. The Allowable Values and nominal trip setpoints(NTSP) are derived, using the General Electric setpoint methodologyguidance, as specified in the Monticello setpoint methodology. TheAllowable Values are derived from the analytic limits. The differencebetween the analytic limit and the Allowable Value allows for channelinstrument accuracy, calibration accuracy, process measurementaccuracy, and primary element accuracy. The margin between theAllowable Value and the NTSP allows for instrument drift that might occurduring the established surveillance period. Two separate verifications areperformed for the calculated NTSP. The first, a Spurious Trip AvoidanceTest, evaluates the impact of the NTSP on plant availability. The secondverification, an LER Avoidance Test, calculates the probability of avoidinga Licensee Event Report (or exceeding the Allowable Value) due toinstrument drift. These two verifications are statistical evaluations toprovide additional assurance of the acceptability of the NTSP and mayrequire changes to the NTSP. Use of these methods and verificationsprovides the assurance that if the setpoint is found conservative to theAllowable Value during surveillance testing, the instrumentation wouldhave provided the required trip function by the time the process reachedthe analytic limit for the applicable events.Certain Emergency Core Cooling Systems (ECCS) valves (e.g., RHR testline suppression pool cooling isolation) also serve the dual function ofautomatic PCIVs. The signals that isolate these valves are alsoassociated with the automatic initiation of the ECCS. The instrumentationrequirements and ACTIONS associated with these signals are addressedMonticelloB 3.3.6.1-5Revision No. 34 Primary Containment Isolation InstrumentationB 3.3.6.1BASESAPPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)in LCO 3.3.5.1, "Emergency Core Cooling Systems (ECCS)Instrumentation," and are not included in this LCO.In general, the individual Functions are required to be OPERABLE inMODES 1, 2, and 3 consistent with the Applicability for LCO 3.6.1.1,"Primary Containment." Functions that have different Applicabilities arediscussed below in the individual Functions discussion.The specific Applicable Safety Analyses, LCO, and Applicabilitydiscussions are listed below on a Function by Function basis.Main Steam Line Isolationl.a. Reactor Vessel Water Level -Low LowLow reactor pressure vessel (RPV) water level indicates that thecapability to cool the fuel may be threatened. Should RPV water leveldecrease too far, fuel damage could result. Therefore, isolation of theMSIVs and other interfaces with the reactor vessel occurs to preventoffsite dose limits from being exceeded. The Reactor Vessel Water Level-Low Low Function is one of the many Functions assumed to beOPERABLE and capable of providing isolation signals. The ReactorVessel Water Level -Low Low Function associated with isolation isassumed in the analysis of the recirculation line break (Ref. 1). Theisolation of the MSLs on Low Low supports actions to ensure that offsitedose limits are not exceeded for a DBA.Reactor vessel water level signals are initiated from four differentialpressure transmitters that sense the difference between the pressure dueto a constant column of water (reference leg) and the pressure due to theactual water level (variable leg) in the vessel. Four channels of ReactorVessel Water Level -Low Low Function are available and are required tobe OPERABLE to ensure that no single instrument failure can precludethe isolation function.The Reactor Vessel Water Level -Low Low Allowable Value is chosen tobe the same as the ECCS Reactor Vessel Water Level -Low LowAllowable Value (LCO 3.3.5.1) to ensure that the MSLs isolate on apotential loss of coolant accident (LOCA) to prevent offsite doses fromexceeding 10 CFR 50.67 limits.This Function isolates the Group 1 valves.Monticello B 3~3~6.1-6 Revision No. 34MonticelloS3.3.6.1-6Revision No. 34 Primary Containment Isolation InstrumentationB 3.3.6.1BASESAPPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)1.b. Main Steam Line Pressure -LowLow MSL pressure indicates that there may be a problem with the turbinepressure regulation, which could result in a low reactor vessel water levelcondition and the RPV cooling down more than 100°F/hr if the pressureloss is allowed to continue. The Main Steam Line Pressure -LowFunction is directly assumed in the analysis of the pressure regulatorfailure (Ref. 3). For this event, the closure of the MSIVs ensures that theRPV temperature change limit (100°F/hr) is not reached. In addition, thisFunction supports actions to ensure that Safety Limit 2.1.1.1 is notexceeded. (This Function closes the MSIVs prior to pressure decreasingbelow 686 psig, which results in a scram due to MSIV closure, thusreducing reactor power to < 25% RTP.)The MSL low pressure signals are initiated from. four pressure switchesthat are connected to the MSL header close to the turbine stop valves.The pressure switches are arranged such that, even though physicallyseparated from each other, each pressure switch is able to detect lowMSL pressure. Four channels of Main Steam Line Pressure -LowFunction are available and are required to be OPERABLE to ensure thatno single instrument failure can preclude the isolation function.The Allowable Value was selected to be high enough to preventexcessive RPV depressurization.The Main Steam Line Pressure -Low Function is only required to beOPERABLE in MODE 1 since this is when the assumed transient canoccur (Ref. 3).This Function isolates the Group 1 valves.1.c. Main Steam Line Flow -HiqhMain Steam Line Flow -High is provided to detect a break of the MSLand to initiate closure of the MSIVs. If the steam were allowed tocontinue flowing out of the break, the reactor would depressurize and thecore could uncover. If the RPV water level decreases too far, fueldamage could occur. Therefore, the isolation is initiated on high flow toprevent or minimize core damage. The Main Steam Line Flow -HighFunction is one of the Functions assumed in the analysis of the mainsteam line break (MSLB) (Ref. 2). The isolation action, along with thescram function of the Reactor Protection System (RPS), ensures that thefuel peak cladding temperature remains below the limits of 10 CFR 50.46and offsite doses do not exceed the 10 CFR 50.67 limits.MonticelloB 3.3.6.1-7Revision No. 34 Primary Containment Isolation InstrumentationB 3.3.6.1BASESAPPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)The MSL flow signals are initiated from 16 differential pressure indicatingswitches that are connected to the four MSLs (differential pressureindicating switches sense differential pressure across a flow restrictor).The differential pressure indicating switches are arranged such that, eventhough physically separated from each other, all four connected to oneMSL would be able to detect the high flow. Four channels of Main SteamLine Flow -High Function for each MSL (two channels per trip system)are available and are required to be OPERABLE so that no singleinstrument failure will preclude detecting a break in any individual MSL.The Allowable Value is chosen to ensure that offsite dose limits are notexceeded due to the break.This Function isolates the Group 1 valves.1 .d. Main Steam Line Tunnel Temperature -HighMain steam line tunnel temperature is provided to detect a leak in theRCPB in the steam tunnel and provides diversity to the high flowinstrumentation. Temperature is sensed in four different areas of thesteam tunnel above each main steam line. The isolation occurs when avery small leak has occurred in any of the four areas. If the small leak isallowed to continue without isolation, offsite dose limits may be reached.However, credit for these instruments is not taken in any transient oraccident analysis in the USAR, since bounding analyses are performedfor large breaks, such as MSLBs.Main steam line tunnel temperature signals are initiated from bimetallictemperature switches located in the four areas being monitored. Eventhough physically separated from each other, any temperature switch inany of the four areas is able to detect a leak. Therefore, sixteen channelsof Main Steam Line Tunnel Temperature -High Function are available butonly eight channels (two channels in each of the four trip strings) arerequired to be OPERABLE to ensure that no single instrument failure canpreclude the isolation function.The Main Steam Line Tunnel Temperature -High Allowable Value ischosen to detect a leak equivalent to between 5 gpm and 10 gpm.This Function isolates the Group 1 valves.Monticello B 3.3.6.1-8 Revision No. 34MonticelloB 3.3.6.1-8Revision No. 34 Primary Containment Isolation InstrumentationB 3.3.6.1BASESAPPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)Primary Containment Isolation2.a. Reactor Vessel Water Level -LowLow RPV water level indicates that the capability to cool the fuel may bethreatened. The valves whose penetrations communicate with theprimary containment are isolated to limit the release of fission products.The isolation of the primary containment on low RPV water level supportsactions to ensure that offsite dose limits of 10 CFR 50.67 are notexceeded. The Reactor Vessel Water Level -Low Function associatedwith isolation is implicitly assumed in the USAR analysis as these leakagepaths are assumed to be isolated post LOCA.Reactor Vessel Water Level -Low signals are initiated from leveltransmitters that sense the difference between the pressure due to aconstant column of water (reference leg) and the pressure due to theactual water level (variable leg) in the vessel. Four channels of ReactorVessel Water Level -Low Function are available and are required to beOPERABLE to ensure that no single instrument failure can preclude theisolation function.The Reactor Vessel Water Low Level -Low Allowable Value was chosento be the same as the RPS Reactor Vessel Water Level -Low AllowableValue (LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation"),since isolation of these valves is not critical to orderly plant shutdown.This Function isolates the Group 2 drywell and sump isolation valves.2.b. Drywell Pressure -HiqhHigh drywell pressure can indicate a break in the RCPB inside theprimary containment. The isolation of some of the primary containmentisolation valves on high drywell pressure supports actions to ensure thatoffsite dose limits of 10 CFR 50.67 are not exceeded. The DrywellPressure -High Function, associated with isolation of the primarycontainment, is implicitly assumed in the USAR accident analysis asthese leakage paths are assumed to be isolated post LOCA.High drywell pressure signals are initiated from pressure switches thatsense the pressure in the drywell. Four channels of Drywell Pressure -High are available and are required to be OPERABLE to ensure that nosingle instrument failure can preclude the isolation function.The Allowable Value was selected to be the same as the ECCS DrywellPressure -High Allowable Value (LCO 3.3.5.1), since this may beindicative of a LOCA inside primary containment.MonticelloB 3.3.6.1-9Revision No. 34 Primary Containment Isolation InstrumentationB 3.3.6.1BASESAPPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)This Function isolates the Group 2 drywell and sump isolation valves.High Pressure Coolant Injection and Reactor Core Isolation CoolingSystems Isolation3.a, 4.a. HPCI and RCIC Steam Line Flow -HiqhSteam Line Flow -High Functions are provided to detect a break of theRCIC or HPCI steam lines and initiate closure of the steam line isolationvalves of the appropriate system. If the steam is allowed to continueflowing out of the break, the reactor will depressurize and the core canuncover. Therefore, the isolations are initiated on high flow to prevent orminimize core damage. The isolation action, along with the scramfunction of the RPS, ensures that the fuel peak cladding temperatureremains below the limits of 10 CFR 50.46. Specific credit for theseFunctions is not assumed in any USAR accident analyses since thebounding analysis is performed for large breaks such as recirculation andMSL breaks. However, these instruments prevent the RCIC or HPCIsteam line breaks from becoming bounding. The HPCI and RCIC SteamLine Flow- High channels are each provided with a time delay relay toprevent false isolations on HPCI or RCIC Steam Line Flow -High, asapplicable, during system startup transients and therefore improvessystem reliability.The HPCI and RCIC Steam Line Flow -High signals are initiated fromdifferential pressure switches (two for HPCI and two for RCIC) that areconnected to the system steam lines. Two channels of both HPCI andRCIC Steam Line Flow -High Functions are available and are required tobe OPERABLE to ensure that no single instrument failure can precludethe isolation function. In addition, each flow channel is connected to atime delay relay to delay the tripping of the associated HPCI or RCICisolation trip system for a short time.The Allowable Values are chosen to be low enough to ensure that the tripoccurs to prevent fuel damage and maintains the MSLB event as thebounding event. The Allowable Values associated with the time delay arechosen to be long enough to prevent false isolations due to system startsbut not so long as to impact offsite dose calculations.These Functions isolate the Groups 4 and 5 valves, as appropriate.3.b, 4.b. HPCI and RCIC Steam Supply Line Pressure -LowLow HPCI or RCIC steam supply line pressure indicates that the pressureof the steam in the HPCI or RCIC turbine, as applicable, may be too lowto continue operation of the associated systems turbine. These isolationsMonticelloB 3.3.61-10Revision No. 34 Primary Containment Isolation InstrumentationB 3.3.6.1BASESAPPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)are for equipment protection and are not assumed in any transient oraccident analysis in the USAR. However, they also provide a diversesignal to indicate a possible system break. These instruments areincluded in Technical Specifications (TS) because of the potential for riskdue to possible failure of the instruments preventing HPCI and RCICinitiations. Therefore, they meet Criterion 4 of 10 CFR 50.36(c)(2)(ii).The HPCI and RCIC Steam Supply Line Pressure -Low signals areinitiated from pressure switches (four for HPCI and four for RCIC) that areconnected to the system steam line. Four channels of both HPCI andRCIC Steam Supply Line Pressure -Low Functions are available and arerequired to be OPERABLE to ensure that no single instrument failure canpreclude the isolation function.The Allowable Values are selected to be high enough to prevent damageto the systems turbine.These Functions isolate the Groups 4 and 5 valves, as appropriate.3.c, 4.c. HPCI and RCIC Steam Line Area Temperature -HighHPCI and RCIC steam line area temperatures are provided to detect aleak from the associated system steam piping. The isolation occurs whena very small leak has occurred and is diverse to the high flowinstrumentation. If the small leak is allowed to continue without isolation,offsite dose limits may be reached. These Functions are not assumed inany USAR transient or accident analysis, since bounding analyses areperformed for large breaks such as recirculation or MSL breaks.HPCI and RCIC Steam Line Area Temperature -High signals are initiatedfrom bimetallic temperature switches that are appropriately located toprotect the system that is being monitored. Eight instruments monitoreach area. Sixteen channels for each HPCI and RCIC Steam Line AreaTemperature -High Function are available and are required to beOPERABLE to ensure that no single instrument failure can preclude theisolation function.The Allowable Values are set low enough to detect a break in theassociated system piping to ensure the core will not be uncovered andthe radiological consequences are bounded by the main steam line breakanalysis.These Functions isolate the Groups 4 and 5 valves, as appropriate.Monticello B 3.3.6.1-11 Revision No. 34MonticelloB 3.3.6. 1-11Revision No. 34 Primary Containment Isolation InstrumentationB 3.3.6.1BASESAPPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)Reactor Water Cleanup System Isolation5.a. RWCU Flow -HighThe high flow signal is provided to detect a break in the RWCU System.This will detect leaks in the RWCU System when room temperature wouldnot provide detection (i.e., a cold leg break). Should the reactor coolantcontinue to flow out of the break, offsite dose limits may be exceeded.Therefore, isolation of the RWCU System is initiated when high flow issensed to prevent exceeding offsite doses. A time delay is provided toprevent spurious trips during most RWCU operational transients. ThisFunction is not assumed in any USAR transient or accident analysis,since bounding analyses are performed for large breaks such as MSLBs.The high flow signals are initiated from transmitters that monitor RWCUSystem flow. In addition, each flow channel is connected to a time delayrelay to delay the tripping of the flow channel for a short time. Fourchannels of RWCU Flow -High Function are available and are required tobe OPERABLE to ensure that no single instrument failure can precludethe isolation function.The RWCU Flow -High Allowable Value ensures that a break of theRWCU piping is detected. The Allowable Value associated with the timedelay is chosen to be long enough to prevent false isolations due tosystem starts but not so long as to impact offsite dose calculations.This Function isolates the Group 3 valves.5.b. RWCU Room Temperature -HighRWCU room temperatures are provided to detect a leak from the RWCUSystem. The isolation occurs even when very small leaks have occurredand is diverse to the high differential flow instrumentation for the hotportions of the RWCU System. If the small leak continues withoutisolation, offsite dose limits may be reached. Credit for these instrumentsis not taken in any transient or accident analysis in the USAR, sincebounding analyses are performed for large breaks such as recirculation orMSL breaks.RWCU room temperature signals are initiated from temperature elementsthat are located in the room that is being monitored. Four resistancetemperature detectors provide input to the RWCU Room Temperature -High Function. Four channels are required to be OPERABLE to ensurethat no single instrument failure can preclude the isolation function.Monticello B 3.3.6.1-12 Revision No. 34MonticelloB 3.3.6.1-12Revision No. 34 Primary Containment Isolation InstrumentationB 3.3.6.1BASESAPPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)The RWCU Room Temperature -High Allowable Value is set low enoughto detect a leak equivalent to 210 gpm.This Function isolates the Group 3 valves.5.c. Drywell Pressure -HighHigh drywell pressure can indicate a break in the RCPB inside theprimary containment. The isolation of some of the primary containmentisolation valves on high drywell pressure supports actions to ensure thatoffsite dose limits of 10 CFR 50.67 are not exceeded. The DrywellPressure -High Function, associated with isolation of the primarycontainment, is implicitly assumed in the USAR accident analysis asthese leakage paths are assumed to be isolated post LOCA.High drywell pressure signals are initiated from pressure switches thatsense the pressure in the drywell. Four channels of Drywell Pressure -High Function are available and are required to be OPERABLE to ensurethat no single instrument failure can preclude the isolation function.The Allowable Value was selected to be the same as the ECCS DrywellPressure -High Allowable Value (LCO 3.3.5.1), since this may beindicative of a LOCA inside primary containment.This Function isolates the Group 3 valves.5.d. SLC System InitiationThe isolation of the RWCU System is required when the SLC System hasbeen initiated to prevent dilution and removal of the boron solution by theRWCU System (Ref. 4). SLC System initiation signals are initiated fromthe SLC initiation switch.Two channels of the SLC System Initiation Function are available and arerequired to be OPERABLE only in MODES 1 and 2, since these are theonly MODES where the reactor can be critical, and these MODES areconsistent with the Applicability for the SLC System (LCO 3.1.7, "StandbyLiquid Control (SLC) System").There is no Allowable Value associated with this Function since thechannels are mechanically actuated based solely on the position of theSLC System initiation switch.This Function isolates the Group 3 valves.Monticello B 3.3.6.1-13 Revision No. 34MonticelloB 3.3.6.1-13Revision No. 34 Primary Containment Isolation InstrumentationB 3.3.6.1BASESAPPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)5.e. Reactor Vessel Water Level -Low LowLow RPV water level indicates that the capability to cool the fuel may bethreatened. Should RPV water level decrease too far, fuel damage couldresult. Therefore, isolation of some interfaces with the reactor vesseloccurs to isolate the potential sources of a break. The isolation of theRWCU System on low low RPV water level supports actions to ensurethat the fuel peak cladding temperature remains below the limits of10 CFR 50.46. The Reactor Vessel Water Level -Low Low Functionassociated with RWCU isolation is not directly assumed in the USARsafety analyses because the RWCU System line break is bounded bybreaks of larger systems (recirculation and MSL breaks are morelimiting).Reactor Vessel Water Level -Low Low signals are initiated from fourdifferential pressure transmitters that sense the difference between thepressure due to a constant column of water (reference leg) and thepressure due to the actual water level (variable leg) in the vessel. Fourchannels of Reactor Vessel Water Level -Low Low Function are availableand are required to be OPERABLE to ensure that no single instrumentfailure can preclude the isolation function.The Reactor Vessel Water Level -Low Low Allowable Value was chosento be the same as the ECCS Reactor Vessel Water Level -Low LowAllowable Value (LCO 3.3.5.1), since the capability to cool the fuel maybe threatened.This Function isolates the Group 3 valves.Shutdown Cooling System Isolation6.a. Reactor Steam Dome Pressure -HighThe Reactor Steam Dome Pressure -High Function is provided to isolatethe shutdown cooling portion of the Residual Heat Removal (RHR)System. This interlock is provided only for equipment protection toprevent an intersystem LOCA scenario, and credit for the interlock is notassumed in the accident or transient analysis in the USAR.The Reactor Steam Dome Pressure -High signals are initiated from twotransmitters that are connected to different taps on the RPV. Twochannels of Reactor Steam Dome Pressure -High Function are availableand are required to be OPERABLE to ensure that no single instrumentfailure can preclude the isolation function. The Function is only requiredto be OPERABLE in MODES 1, 2, and 3, since these are the onlyMODES in which the reactor can be pressurized; thus, equipmentMonticelloB 3.3.6.1-14Revision No- 34 Primary Containment Isolation InstrumentationB 3.3.6.1BASESAPPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)protection is needed. The Allowable Value was chosen to be low enoughto protect the system equipment from overpressurization.This Function isolates the Group 2 RHR shutdown cooling supplyisolation valves.6.b. Reactor Vessel Water Level -LowLow RPV water level indicates that the capability to cool the fuel may bethreatened. Should RPV water level decrease too far, fuel damage couldresult. Therefore, isolation of some reactor vessel interfaces occurs tobegin isolating the potential sources of a break. The Reactor VesselWater Level -Low Function associated with RHR Shutdown CoolingSystem isolation is not directly assumed in safety analyses because abreak of the RHR Shutdown Cooling System is bounded by breaks of therecirculation and MSL. The RHR Shutdown Cooling System isolation onlow RPV water level supports actions to ensure that the RPV water leveldoes not drop below the top of the active fuel during a vessel draindownevent caused by a leak (e.g., pipe break or inadvertent valve opening) inthe RHR Shutdown Cooling System.Reactor Vessel Water Level -Low signals are initiated from fourdifferential pressure transmitters that sense the difference between thepressure due to a constant column of water (reference leg) and thepressure due to the actual water level (variable leg) in the vessel. Fourchannels (two channels per trip system) of the Reactor Vessel WaterLevel -Low Function are available and are required to be OPERABLE toensure that no single instrument failure can preclude the isolationfunction. As noted (footnote (a) to Table 3.3.6.1-1), only one channel pertrip system (with an isolation signal available to one shutdown coolingpump supply isolation valve) of the Reactor Vessel Water Level -LowFunction is required to be OPERABLE in MODES 4 and 5, provided RHRShutdown Cooling System integrity is maintained. System integrity ismaintained provided the piping is intact and no maintenance is beingperformed that has the potential for draining the reactor vessel throughthe system.The Reactor Vessel Water Level -Low Allowable Value was chosen to bethe same as the RPS Reactor Vessel Water Level -Low Allowable Value(LCO 3.3.1.1), since the capability to cool the fuel may be threatened.The Reactor Vessel Water Level -Low Function is only required to beOPERABLE in MODES 3, 4, and 5 to prevent this potential flow path fromlowering the reactor vessel level to the top of the fuel. In MODES 1and 2, another isolation (i.e., Reactor Steam Dome Pressure -High) andMonticelloB 3.3.6.1-15Revision No. 34 Primary Containment Isolation InstrumentationB 3.3.6.1BASESAPPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)administrative controls ensure that this flow path remains isolated toprevent unexpected loss of inventory via this flow path.This Function isolates the Group 2 RHR shutdown cooling supplyisolation valves.Traversing Incore Probe System Isolation7.a. Reactor Vessel Water Level -LowLow RPV water level indicates that the capability to cool the fuel may bethreatened. The valves whose penetrations communicate with theprimary containment are isolated to limit the release of fission products.The isolation of the primary containment on low RPV water level supportsactions to ensure that offsite dose limits of 10 CFR 50.67 are notexceeded. The Reactor Vessel Water Level -Low Function associatedwith isolation is implicitly assumed in the USAR analysis as these leakagepaths are assumed to be isolated post LOCA.Reactor Vessel Water Level -Low signals are initiated from differentialpressure transmitters that sense the difference between the pressure dueto a constant column of water (reference leg) and the pressure due to theactual water level (variable leg) in the vessel. Two channels of ReactorVessel Water Level -Low Function are available and are required to beOPERABLE to ensure that no single instrument failure can initiate aninadvertent isolation actuation. The isolation function is ensured by themanual shear valve in each penetration.The Reactor Vessel Water Level -Low Allowable Value was chosen to bethe same as the RPS Reactor Vessel Water Level -Low Allowable Value(LCO 3.3.1.1), since isolation of these valves is not critical to orderly plantshutdown.This Function isolates the Group 2 TIP inboard isolation ball valves.7.b. Drywell Pressure -HighHigh drywell pressure can indicate a break in the RCPB inside theprimary containment. The isolation of some of the primary containmentisolation valves on high drywell pressure supports actions to ensure thatoffsite dose limits of 10 CFR 50.67 are not exceeded. The DrywellPressure -High Function, associated with isolation of the primarycontainment, is implicitly assumed in the USAR accident analysis asthese leakage paths are assumed to be isolated post LOCA.MonticelloB 3.3.6.1-16Revision No. 34 Primary Containment Isolation InstrumentationB 3.3.6.1BASESAPPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)High drywell pressure signals are initiated from pressure transmitters thatsense the pressure in the drywell. Two channels of Drywell Pressure -High Function are available and are required to be OPERABLE to ensurethat no single instrument failure can initiate an inadvertent actuation. Theisolation function is ensured by the manual shear valve in eachpenetration.The Allowable Value was selected to be the same as the ECCS DrywellPressure -High Allowable Value (LCO 3.3.5.1), since this may beindicative of a LOCA inside primary containment.This Function isolates the Group 2 TIP inboard isolation ball valves.ACTIONS The ACTIONS are modified by two Notes. Note 1 allows penetration flowpath(s) to be unisolated intermittently under administrative controls.These controls consist of stationing a dedicated operator at the controlsof the valve, who is in continuous communication with the control room.In this way, the penetration can be rapidly isolated when a need forprimary containment isolation is indicated. Note 2 has been provided tomodify the ACTIONS related to primary containment isolationinstrumentation channels. Section 1.3, Completion Times, specifies thatonce a Condition has been entered, subsequent divisions, subsystems,components, or variables expressed in the Condition, discovered to beinoperable or not within limits, will not result in separate entry into theCondition. Section 1.3 also specifies that Required Actions of theCondition continue to apply for each additional failure, with CompletionTimes based on initial entry into the Condition. However, the RequiredActions for inoperable primary containment isolation instrumentationchannels provide appropriate compensatory measures for separateinoperable channels. As such, a Note has been provided that allowsseparate Condition entry for each inoperable primary containmentisolation instrumentation channel.A.1Because of the diversity of sensors available to provide isolation signalsand the redundancy of the isolation design, an allowable out of servicetime of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , depending on the Function (12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months forthose Functions that have channel components common to RPSinstrumentation and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for those Functions that do not havechannel components common to RPS instrumentation), has been shownto be acceptable (Refs. 5 and 6) to permit restoration of any inoperablechannel to OPERABLE status. This out of service time is only acceptableprovided the associated Function is still maintaining isolation capabilityMonticelloB 3.3.6.1-17Revision No. 34 Primary Containment Isolation InstrumentationB 3.3.6.1BASESACTIONS (continued)(refer to Required Action B.1 Bases). If the inoperable channel cannot berestored to OPERABLE status within the allowable out of service time, thechannel must be placed in the tripped condition per Required Action A.1.Placing the inoperable channel in trip would conservatively compensatefor the inoperability, restore capability to accommodate a single failure,and allow operation to continue with no further restrictions. Alternately, ifit is not desired to place the channel in trip (e.g., as in the case whereplacing the inoperable channel in trip would result in an isolation),Condition C must be entered and its Required Action taken.B. 1Required Action B.1 is intended to ensure that appropriate actions aretaken if multiple, inoperable, untripped channels within the same Functionresult in redundant primary containment isolation capability being lost forthe associated penetration flow path(s). The MSL, Primary Containment,most of the RWCU System, Shutdown Cooling System Reactor VesselWater Level -Low, and TIP Isolation Functions are considered to bemaintaining primary containment isolation capability when sufficientchannels are OPERABLE or in trip, such that both trip systems willgenerate a trip signal from the given Function on a valid signal. The otherisolation Functions are considered to be maintaining primary containmentisolation capability when sufficient channels are OPERABLE or in trip,*such that one trip system will generate a trip signal from the givenFunction on a valid signal. This ensures that one of the two PCIVs in theassociated penetration flow path can receive an isolation signal from thegiven Function. For Functions 1.a, 1.b, 2.a, 2.b, 5.a, 5.b, 5.c, 5.e, 6.b,7.a, and 7.b, this would require both trip systems to have one channelOPERABLE or in trip. For Function 1.c, this would require both tripsystems to have one channel, associated with each MSL, OPERABLE orin trip. Function 1.d channels monitor several locations within a givenarea (e.g., different locations within the main steam tunnel area).However, since any channel can detect a leak in any area, this wouldrequire both trip systems to have one channel OPERABLE or in trip. ForFunctions 3.a, 4.a, and 5.d, this would require one trip system to haveone channel OPERABLE or in trip. For Function 3.b, this would requireone channel in each trip string to be OPERABLE or in trip for the tripsystem. For Function 4.b, this would require one channel in each tripstring to be OPERABLE or in trip for one trip system. For Functions 3.cand 4.c, eight channels monitor each area. These channels are arrangedin two sets of four detectors, with each set of detectors arranged in a one-out-of-two-twice logic. Therefore, this would require a set in each area tohave sufficient channels OPERABLE or in the tripped condition for onetrip system.MonticelloB 3.3.6.1-18Revision No. 34 Primary Containment Isolation InstrumentationB 3.3.6.1BASESACTIONS (continued)The Completion Time is intended to allow the operator time to evaluateand repair any discovered inoperabilities. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time isacceptable because it minimizes risk while allowing time for restoration ortripping of channels.C._ 1Required Action C.1 directs entry into the appropriate Conditionreferenced in Table 3.3.6.1-1. The applicable Condition specified inTable 3.3.6.1-1 is Function and MODE or other specified conditiondependent and may change as the Required Action of a previousCondition is completed. Each time an inoperable channel has not metany Required Action of Condition A or B and the associated CompletionTime has expired, Condition C will be entered for that channel andprovides for transfer to the appropriate subsequent Condition.D.1, D.2.1, and D.2.2If the channel is not restored to OPERABLE status or placed in trip withinthe allowed Completion Time, the plant must be placed in a MODE orother specified condition in which the LCO does not apply. This is doneby placing the plant in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in MODE 4within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (Required Actions D.2.1 and D.2.2). Alternately, theassociated MSLs may be isolated (Required Action D.1), and, if allowed(i.e., plant safety analysis allows operation with an MSL isolated),operation with that MSL isolated may continue. Isolating the affectedMSL accomplishes the safety function of the inoperable channel. TheCompletion Times are reasonable, based on operating experience, toreach the required plant conditions from full power conditions in anorderly manner and without challenging plant systems.E. 1If the channel is not restored to OPERABLE status or placed in trip withinthe allowed Completion Time, the plant must be placed in a MODE orother specified condition in which the LCO does not apply. This is doneby placing the plant in at least MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based onoperating experience, to reach MODE 2 from full power conditions in anorderly manner and without challenging plant systems.Monticello B 3.3.6.1-19 Revision No. 34MonticelloB 3.3.6.1-19Revision No. 34 Primary Containment Isolation InstrumentationB 3.3.6.1BASESACTIONS (continued)F. 1If the channel is not restored to OPERABLE status or placed in trip withinthe allowed Completion Time, plant operations may continue if theaffected penetration flow path(s) is isolated. Isolating the affectedpenetration flow path(s) accomplishes the safety function of theinoperable channels.For the RWCU Room Temperature -High Function, the affectedpenetration flow path(s) may be considered isolated by isolating only thatportion of the system in the associated room monitored by the inoperablechannel. That is, if the RWCU pump room A area channel is inoperable,the pump room A area can be isolated while allowing continued RWCUoperation utilizing the B RWCU pump.The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is acceptable because it minimizes riskwhile allowing sufficient time for plant operations personnel to isolate theaffected penetration flow path(s).G.1If the channel is not restored to OPERABLE status or placed in trip withinthe allowed Completion Time, plant operations may continue if theaffected penetration flow path(s) is isolated. Isolating the affectedpenetration flow path(s) accomplishes the safety function of theinoperable channels. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is acceptable due tothe fact that these Functions provide a TIP System isolation, and the TIPSystem penetration is a small bore (approximately 1 inch), its isolation ina design basis event (with loss of offsite power) would be via themanually operated shear valves, and the ability to manually isolate byeither the normal isolation valve or the shear valve is unaffected by theinoperable instrumentation.H.1 and H.2If the channel is not restored to OPERABLE status or placed in trip withinthe allowed Completion Time, the associated SLC subsystem(s) isdeclared inoperable or the RWCU System is isolated. Since this Functionis required to ensure that the SLC System performs its intended function,sufficient remedial measures are provided by declaring the associatedSLC subsystems inoperable or isolating the RWCU System.The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is acceptable because it minimizes riskwhile allowing sufficient time for personnel to isolate the RWCU System.MonticelloB 3.3.6.1-20Revision No. 34 Primary Containment Isolation InstrumentationB 3.3.6.1BASESACTIONS (continued)1.1 and 1.2If the channel is not restored to OPERABLE status or placed in trip withinthe allowed Completion Time, the associated penetration flow path shouldbe closed. However, if the shutdown cooling function is needed toprovide core cooling, these Required Actions allow the penetration flowpath to remain unisolated provided action is immediately initiated torestore the channel to OPERABLE status or to isolate the RHR ShutdownCooling System (i.e., provide alternate decay heat removal capabilities sothe penetration flow path can be isolated). Actions must continue until thechannel is restored to OPERABLE status or the RHR Shutdown CoolingSystem is isolated.SURVEILLANCE As noted at the beginning of the SRs, the SRs for each PrimaryREQUIREMENTS Containment Isolation instrumentation Function are found in the SRscolumn of Table 3.3.6.1-1.The Surveillances are modified by a Note to indicate that when a channel(a channel that is directed to two trip systems is considered to be onechannel) is placed in an inoperable status solely for performance ofrequired Surveillances, entry into associated Conditions and RequiredActions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associatedFunction maintains primary containment isolation capability. Uponcompletion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, thechannel must be returned to OPERABLE status or the applicableCondition entered and Required Actions taken. This Note is based on thereliability analysis (Refs. 5 and 6) assumption of the average timerequired to perform channel surveillance. That analysis demonstratedthat the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance does not significantly reduce theprobability that the PCIVs will isolate the penetration flow path(s) whennecessary.SR 3.3.6.1.1Performance of the CHANNEL CHECK once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ensures thata gross failure of instrumentation has not occurred. A CHANNEL CHECKis normally a comparison of the parameter indicated on one channel to asimilar parameter on other channels. It is based on the assumption thatinstrument channels monitoring the same parameter should readapproximately the same value. Significant deviations between theinstrument channels could be an indication of excessive instrument drift inone of the channels or of something even more serious. A CHANNELCHECK will detect gross channel failure; thus, it is key to verifying theinstrumentation continues to operate properly between each CHANNELCALIBRATION.MonticelloB 3.3.6.1-21Revision No. 34 Primary Containment Isolation InstrumentationB 3.3.6.1BASESSURVEILLANCE REQUIREMENTS (continued)Agreement criteria are determined by the plant staff based on acombination of the channel instrument uncertainties, including indicationand readability. If a channel is outside the criteria, it may be an indicationthat the instrument has drifted outside its limit.The Frequency is based on operating experience that demonstrateschannel failure is rare. The CHANNEL CHECK supplements less formal,but more frequent, checks of channels during normal operational use ofthe displays associated with the channels required by the LCO.SR 3.3.6.1.2A CHANNEL FUNCTIONAL TEST is performed on each required channelto ensure that the channel will perform the intended function. Asuccessful test of the required contact(s) of a channel relay may beperformed by the verification of the change of state of a single contact ofthe relay. This clarifies what is an acceptable CHANNEL FUNCTIONALTEST of a relay. This is acceptable because all of the other requiredcontacts of the relay are verified by other Technical Specifications andnon-Technical Specifications tests at least once per refueling interval withapplicable extensions.Any setpoint adjustment shall be consistent with the assumptions of thecurrent plant specific setpoint methodology.The 92 day Frequency of SR 3.3.6.1.2 is based on the reliability analysesdescribed in References 5 and 6.SR 3.3.6.1.3Calibration of trip units provides a check of the actual trip setpoints(including any specified time delay). The channel must be declaredinoperable if the trip setting is discovered to be less conservative than theAllowable Value specified in Table 3.3.6.1-1. If the trip setting isdiscovered to be less conservative than accounted for in the appropriatesetpoint methodology, but is not beyond the Allowable Value, the channelperformance is still within the requirements of the plant safety analysis.Under these conditions, the setpoint must be readjusted to be equal to ormore conservative than that accounted for in the appropriate setpointmethodology.The Frequency of 92 days is based on the reliability analyses ofReferences 5 and 6.MonticelloB 3.3.6.1-22Revision No. 34 Primary Containment Isolation InstrumentationB 3.3.6.1BASESSURVEILLANCE REQUIREMENTS (continued)SR 3.3.6.1.4 and SR 3.3.6.1.5A CHANNEL CALIBRATION is a complete check of the instrument loopand the sensor. This test verifies the channel responds to the measuredparameter within the necessary range and accuracy. CHANNELCALIBRATION leaves the channel adjusted to account for instrumentdrifts between successive calibrations consistent with the plant specificsetpoint methodology.The Frequency of SR 3.3.6.1.4 is based on the assumption of a 92 daycalibration interval in the determination of the magnitude of equipmentdrift in the setpoint analysis. The Frequency of SR 3.3.6.1.5 is based onthe assumption of a 24 month calibration interval in the determination ofthe magnitude of equipment drift in the setpoint analysis.SR 3.3.6.1.6The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITYof the required isolation logic for a specific channel. The system functionaltesting performed on PCIVs in LCO 3.6.1.3 overlaps this Surveillance toprovide complete testing of the assumed safety function. The 24 monthFrequency is based on the need to perform this Surveillance under theconditions that apply during a plant outage and the potential for an unplannedtransient if the Surveillance were performed with the reactor at power.Operating experience has shown these components usually pass theSurveillance when performed at the 24 month Frequency.REFERENCES 1. USAR, Section 14.7.2.2. USAR, Section 14.7.3.3. USAR, Section 7.6.3.2.4.4. USAR, Section 6.6.1.1.5. NEDC-31677P-A, "Technical Specification Improvement Analysis forBWR Isolation Actuation Instrumentation," July 1990.6. NEDC-30851P-A Supplement 2, "Technical SpecificationsImprovement Analysis for BWR Isolation Instrumentation Common toRPS and ECCS Instrumentation," March 1989.Monticello B 3.3.6.1-23 Revision No. 34MonticelloB 3.3.6.1-23Revision No. 34 Primary Containment Isolation InstrumentationB 3.3.6.1BASESREFERENCES (continued)7. Amendment No. 185, "Issuance of Amendment to Reduce theReactor Steam Dome Pressure Specified in the Reactor Core SafetyLimits," dated November 25, 2014. (ADAMS Accession No.ML14281A318)Monticello B 3.3.6.1-24 -Last Revision No. 34MonticelloB 3.3.6.1-24 -LastRevision No. 34 ECCS -Operating3.5.1B 3.5 EMERGENCY CORE COOLING SYSTEM (ECCS) AND REACTOR CORE ISOLATIONCOOLING (RCIC) SYSTEMB 3.5.1 ECCS -OperatingBASESBACKGROUNDThe ECCS is designed, in conjunction with the primary and secondarycontainment, to limit the release of radioactive materials to theenvironment following a loss of coolant accident (LOCA). The ECCSuses two independent methods (flooding and spraying) to cool the coreduring a LOCA. The ECCS network consists of the High PressureCoolant Injection (HPCI) System, the Core Spray (CS) System, the lowpressure coolant injection (LPCI) mode of the Residual Heat Removal(RHR) System, and the Automatic Depressurization System (ADS). Thesuppression pool provides the required source of water for the ECCS.Although no credit is taken in the safety analyses for the condensatestorage tanks (CSTs), they are capable of providing a source of water forthe HPCI, LPCI, and CS Systems.On receipt of an initiation signal, ECCS pumps automatically start and thesystem aligns and the pumps inject water, taken either from the CSTs orsuppression pool, into the Reactor Coolant System (RCS) as RCSpressure is overcome by the discharge pressure of the ECCS pumps.Although the system is initiated, ADS action is delayed, allowing theoperator to interrupt the timed sequence if the system is not needed. TheHPCI pump discharge pressure almost immediately exceeds that of theRCS, and the pump injects coolant into the vessel to cool the core. If thebreak is small, the HPCI System will maintain coolant inventory as well asvessel level while the RCS is still pressurized. If HPCI fails, it is backedup by ADS in combination with LPCI and CS. In this event, the ADStimed sequence would be allowed to time out and open the selectedsafety/relief valves (S/RVs) depressurizing the RCS, thus allowing theLPCI and CS to overcome RCS pressure and inject coolant into thevessel. If the break is large, RCS pressure initially drops rapidly and theLPCI and CS cool the core.Water from the break returns to the suppression pool where it is usedagain and again. Water in the suppression pool is circulated through aheat exchanger cooled by the RHR Service Water System. Dependingon the location and size of the break, portions of the ECCS may beineffective; however, the overall design is effective in cooling the coreregardless of the size or location of the piping break.The combined operation of all ECCS subsystems are designed to ensurethat no single active component failure will prevent automatic initiationand successful operation of the minimum required ECCS equipment.MonticelloB 3.5. 1-1Revision No. 34 ECCS -Operating3.5.1BASESBACKGROUND (continued)The CS System (Ref. 1) is composed of two independent subsystems.Each subsystem consists of a motor driven pump, a spray sparger abovethe core, and piping and valves to transfer water from the suppressionpool to the sparger. The CS System is designed to provide cooling to thereactor core when reactor pressure is low. Upon receipt of an initiationsignal, the CS pumps in both subsystems are automatically started inapproximately 15 seconds after AC power is available. When the RPVpressure drops sufficiently, CS System flow to the RPV begins. A full flowtest line is provided to route water from and to the suppression pool toallow testing of the CS System without spraying water in the RPV.LPCI is an independent operating mode of the RHR System. There aretwo LPCI subsystems (Ref. 2), each consisting of two motor drivenpumps in the same RHR loop and piping and valves to transfer waterfrom the suppression pool to the RPV via the selected recirculation loop.Each LPCI subsystem consists of a common suction line from thesuppression pool, parallel flowpaths through the two RHR pumps, and acommon injection line to the RPV. An inoperable "LPCI pump" refers tothe condition where inoperable components associated with the flowpaththrough one of the two parallel RHR pumps renders that LPCI pumpflowpath inoperable, but the common portions of the associated LPCIsubsystem are OPERABLE.The LPCI System is equipped with a loop select logic that determineswhich, if any, of the recirculation loops has been broken and selects thenon-broken loop for injection. If neither loop is determined to be broken,a preselected loop is used for injection. The LPCI System cross-tie valvemust be open to support OPERABILITY of both LPCI subsystems.Similarly, the LPCI swing bus, consisting of two motor control centerswhich are directly connected together, is required to be energized fromthe Division 1 power supply (normal source), with automatic transfercapability to the Division 2 power supply (backup source) to support bothLPCI subsystems. The LPCI subsystems are designed to provide corecooling at low RPV pressure. Upon receipt of an initiation signal, all fourLPCI pumps are automatically started (pumps A and B approximately5 seconds after AC power is available and pumps C and D approximately10 seconds after AC power is available). RHR System valves in the LPCIflow path are automatically positioned to ensure the proper flow path forwater from the suppression pool to inject into the selected recirculationloop. When the RPV pressure drops sufficiently, the LPCI flow to theRPV, via the selected recirculation loop, begins. The water then entersthe reactor through the jet pumps. Full flow test lines are provided foreach LPCI subsystem to route water from and to the suppression pool, toallow testing of the LPCI pumps without injecting water into the RPV.These test lines also provide suppression pool cooling capability, asdescribed in LCO 3.6.2.3, "RHR Suppression Pool Cooling." An intertieMonticelloB 3.5.1-2Revision No. 34 ECCS -Operating3.5.1BASESBACKGROUND (continued)line is provided to connect the RHR shutdown cooling suction line with thetwo RHR shutdown cooling loop return lines to the associatedrecirculation loop. This line includes two RHR intertie return line isolationvalves that are normally closed and a RHR intertie suction line isolationvalve that is normally open. The purpose of this line is to reduce thepotential for water hammer in the recirculation and RHR systems. Theisolation valves are opened during a cooldown to establish recirculationflow through the RHR suction line and return lines, thereby ensuring auniform cooldown of this piping. The RHR intertie loop return lineisolation valves receive a closure signal on LPCI initiation. In the event ofan inoperable RHR intertie loop return line isolation valve, there is apotential for some of the LPCI flow to be diverted to the broken loopduring a LOCA. This may cause early transition boiling during a LOCAbut this condition was evaluated in the safety analysis and foundacceptable. The RHR intertie line is to be isolated within 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months ifdiscovered open in MODE 1 to eliminate the need to compensate for thesmall change in jet pump drive flow and a reduction in core flow during aloss of coolant accident.The HPCI System (Ref. 3) consists of a steam driven turbine pump unit,piping, and valves to provide steam to the turbine, as well as piping andvalves to transfer water from the suction source to the core via thefeedwater system line, where the coolant is distributed within the RPVthrough the feedwater sparger. Suction piping for the system is providedfrom the CSTs and the suppression pool. Pump suction for HPCI isnormally aligned to the CSTs to minimize injection of suppression poolwater into the RPV. However, if the water level in any CST is low, or ifthe suppression pool level is high, an automatic transfer to thesuppression pool water source ensures a water supply for continuousoperation of the HPCI System. The steam supply to the HPCI turbine ispiped from a main steam line upstream of the associated inboard mainsteam isolation valve.The HPCI System is designed to provide core cooling for a wide range ofreactor pressures (150 psig to 1120 psig). Upon receipt of an initiationsignal, the HPCI turbine stop valve and turbine steam supply valve openand the turbine accelerates to a specified speed. As the HPCI flowincreases, the turbine governor valve is automatically adjusted tomaintain design flow. Exhaust steam from the HPCI turbine is dischargedto the suppression pool. A full flow test line is provided to route waterfrom and to the CSTs to allow testing of the HPCI System during normaloperation without injecting water into the RPV.The ECCS pumps are provided with minimum flow bypass lines, whichdischarge to the suppression pool. The valves in these linesautomatically open or remain open to prevent pump damage due toMonticelloB 3.5.1-3Revision No. 34 ECCS -Operating3.5.1BASESBACKGROUND (continued)overheating when other discharge line valves are closed. To ensurerapid delivery of water to the RPV and to minimize water hammer effects,all ECCS pump discharge lines are filled with water. The LPCI and CSSystem discharge lines are kept full of water using a "keep fill" system(Condensate Service System). The HPCI System is normally aligned tothe CSTs. The height of water in the CSTs maintains the piping full ofwater up to the first closed isolation valve in the discharge piping. TheHPCI System discharge piping near the normally closed injection valve tothe Feedwater System absorbs heat from the feedwater via conductionand valve leakage. This has the potential to form a localized steam voidin the HPCI discharge piping and cause a momentum transient uponHPCI initiation. Although the momentum transient has been evaluatedand shown not to adversely affect HPCI System operation, theCondensate System is utilized as a "keep-fill" system to maintain theHPCI discharge piping between the normally closed injection valve andthe pump discharge check valve charged with water to prevent possiblevoid formation and minimize momentum transient effects. This "keep-fill"system is relied upon during normal operation, but is not required for theoperability of the HPCI System under normal plant conditions. Additionalassessment of operability may be required under off-normal conditions,such as HPCI suction aligned to the suppression pool. The relativeheight of the feedwater line connection for HPCI is such that the water inthe feedwater lines keeps the remaining portion of the HPCI dischargeline full of water.The ADS (Ref. 4) consists of three of the eight S/RVs. It is designed toprovide depressurization of the RCS during a small break LOCA if HPCIfails or is unable to maintain required water level in the RPV. ADSoperation reduces the RPV pressure to within the operating pressurerange of the low pressure ECCS subsystems (CS and LPCI), so thatthese subsystems can provide coolant inventory makeup. The ADSvalves are normally supplied by the Instrument Nitrogen System. Thispneumatic supply will automatically transfer to the Instrument Air Systemon high or low Instrument Nitrogen System pressure. However, both ofthese pneumatic supplies are non-safety related and are not assumed tooperate following an accident. The safety grade pneumatic supply to twoof the ADS valves is the Alternate Nitrogen System and to the third ADSvalve is the S/RV Accumulator bank. The Alternate Nitrogen Systemcontains two independent trains (i.e., subsystems) of safety relatedreplaceable gas cylinders that supply two of the three ADS valves (S/RVsA and C). One Alternate Nitrogen System train supplies one ADS valveand other non-ADS related pneumatic loads and the other AlternateNitrogen System train supplies a different ADS valve and other non-ADSrelated pneumatic loads. The S/RV Accumulator Bank supplies the thirdADS valve (S/RV D), and consists of a dedicated safety related backupaccumulator bank and an associated inlet check valve.MonticelloB 3.5.1-4Revision No. 34 ECCS -Operating3.5.1BASESAPPLICABLE The ECCS performance is evaluated for the entire spectrum of breakSAFETY sizes for a postulated LOCA. The accidents for which ECCS operation isANALYSES required are presented in References 5 and 6. The required analysesand assumptions are defined in Reference 7. The results of theseanalyses are also described in References 5 and 6.This LCO helps to ensure that the following acceptance criteria for theECCS (Ref. 8), established by 10 CFR 50.46 (Ref. 9), will be metfollowing a LOCA, assuming the worst case single active componentfailure in the ECCS:a. Maximum fuel element cladding temperature is < 2200°F;b. Maximum cladding oxidation is < 0.170times the total claddingthickness before oxidation;c. Maximum hydrogen generation from a zirconium water reaction is< 0.01 times the hypothetical amount that would be generated if all ofthe metal in the cladding surrounding the fuel, excluding the claddingsurrounding the plenum volume, were to react;d. The core ig maintained in a coolable geometry; ande. Adequate long term cooling capability is maintained.The limiting single failures are discussed in Reference 10. For a largedischarge pipe break LOCA, failure of the LPCI valve on the unbrokenrecirculation loop is considered the most limiting break/failurecombination. For a small break LOCA, HPCI failure is the most severefailure. Extended Power Uprate removed the allowance for one ADSvalve out-of-service (Ref. 17). The remaining OPERABLE ECCSsubsystems provide the capability to adequately cool the core andprevent excessive fuel damage.The ECCS satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).LCO Each ECCS injection/spray subsystem and three ADS valves are requiredto be OPERABLE. The ECCS injection/spray subsystems are defined asthe two CS subsystems, the two LPCI subsystems, and one HPCISystem. The low pressure ECCS injection/spray subsystems are definedas the two CS subsystems and the two LPCI subsystems.With less than the required number of ECCS subsystems OPERABLE,the potential exists that during a limiting design basis LOCA concurrentwith the worst case single failure, the limits specified in Reference 9 couldbe exceeded. All ECCS subsystems must therefore be OPERABLE tosatisfy the single failure criterion required by Reference 9.MonticelloB 3.5.1-5Revision No. 34 ECCS -Operating3.5.1BASESLCO (continued)As noted, LPCI subsystems may be considered OPERABLE duringalignment and operation for decay heat removal when below the actualRHR shutdown cooling supply isolation interlock in MODE 3, if capable ofbeing manually realigned (remote or local) to the LPCI mode and nototherwise inoperable. Alignment and operation for decay heat removalincludes when the required RHR pump is not operating or when thesystem is realigned from or to the RHR shutdown cooling mode. Thisallowance is necessary since the RHR System may be required tooperate in the shutdown cooling mode to remove decay heat and sensibleheat from the reactor. At these low pressures and decay heat levels, areduced complement of ECCS subsystems should provide the requiredcore cooling, thereby allowing operation of RHR shutdown cooling whennecessary.APPLICABILITYAll ECCS subsystems are required to be OPERABLE during MODES 1,2, and 3, when there is considerable energy in the reactor core and corecooling would be required to prevent fuel damage in the event of a breakin the primary system piping. In MODES 2 and 3, when reactor steamdome pressure is < 150 psig, ADS and HPCI are not required to beOPERABLE because the low pressure ECCS subsystems can providesufficient flow below this pressure. ECCS requirements for MODES 4and 5 are specified in LCO 3.5.2, "ECCS -Shutdown."ACTIONSA Note prohibits the application of LCO 3.0.4.b to an inoperable HPCIsubsystem. There is an increased risk associated with entering a MODEor other specified condition in the Applicability with an inoperable HPCIsubsystem and the provisions of LCO 3.0.4.b, which allow entry into aMODE or other specified condition in the Applicability with the LCO notmet after performance of a risk assessment addressing inoperablesystems and components, should not be applied in this circumstance.A. 1If one LPCI pump is inoperable, the inoperable pump must be restored toOPERABLE status within 30 days. In this condition, the remainingOPERABLE pumps provide adequate core cooling during a LOCA.However, overall LPCI reliability is reduced, because a single failure inone of the remaining OPERABLE LPCI subsystems, concurrent with aLOCA, may result in the LPCI subsystems not being able to perform theirintended safety function. The 30 day Completion Time is based on areliability study cited in Reference 11 that evaluated the impact on ECCSavailability, assuming various components and subsystems were takenout of service. The results were used to calculate the average availabilityof ECCS equipment needed to mitigate the consequences of a LOCA asa function of allowable repair times (i.e., Completion Times).Monticello B 3.5.1-6 Revision No. 34MonticelloB 3.5.1-6Revision No. 34 ECCS -Operating3.5.1BASESACTIONS (continued)B. 1If a LPCI subsystem is inoperable for reasons other than Condition A, or aCS subsystem is inoperable, the inoperable low pressure injection/spraysubsystem must be restored to OPERABLE status within 7 days. In thiscondition, the remaining OPERABLE subsystems provide adequate corecooling during a LOCA. However, overall ECCS reliability is reduced,because a single failure in one of the remaining OPERABLE subsystems,concurrent with a LOCA, may result in the ECCS not being able toperform its intended safety function. The 7 day Completion Time is basedon a reliability study (Ref. 11) that evaluated the impact on ECCSavailability, assuming various components and subsystems were takenout of service. The results were used to calculate the average availabilityof ECCS equipment needed to mitigate the consequences of a LOCA asa function of allowed outage times (i.e., Completion Times).C._1If one LPCI pump in each subsystem is inoperable, one inoperable LPCIpump must be restored to OPERABLE status within 7 days. In thiscondition, the remaining OPERABLE ECCS subsystems provideadequate core cooling during a LOCA. However, overall ECCS reliabilityis reduced because a single failure in one of the remaining OPERABLEECCS subsystems, concurrent with a LOCA, may result in the ECCS notbeing able to perform its intended safety function. The 7 day CompletionTime is based on a reliability study (Ref. 11) that evaluated the impact onECCS availability, assuming various components and subsystems weretaken out of service. The results were used to calculate the averageavailability of ECCS equipment needed to mitigate the consequences of aLOCA as a function of allowed outage times (i.e., Completion Times).D. 1If two LPCI subsystems are inoperable for reasons other than Condition Cor G, one inoperable subsystem must be restored to OPERABLE statuswithin 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . In this condition, the remaining OPERABLECS subsystems provide adequate core cooling during a LOCA. However,overall ECCS reliability is reduced, because a single failure in one of theremaining CS subsystems, concurrent with a LOCA, may result in ECCSnot being able to perform its intended safety function. The 72 hourCompletion Time is based on a reliability study cited in Reference 11 thatevaluated the impact on ECCS availability, assuming various componentsand subsystems were taken out of service; and on previous BWRlicensing precedents, and was approved for Monticello by AmendmentMonticelloB 3.5.1-7Revision No. 34 ECCS -Operating3.5.1BASESACTIONS (continued)162 (Reference 14). The results were used to calculate the averageavailability of ECCS equipment needed to mitigate the consequences of aLOCA as a function of allowable repair times (i.e., Completion Times).E.1, E.2 and E.3If any one low pressure CS subsystem is inoperable in addition to eitherone LPCI subsystem OR one or two LPCI pump(s), adequate corecooling is ensured by the OPERABILITY of HPCI and the remaining lowpressure ECCS subsystems. This condition results in a complement ofremaining OPERABLE low pressure ECCS (i.e., one CS and either two orthree LPCI pumps) whose makeup capacity is bounded by the minimummakeup capacity evaluated in the accident analysis, which assumes thelimiting single component failure (Reference 10). However, overall ECCSreliability is reduced, because a single active component failure in theremaining low pressure ECCS, concurrent with a design basis LOCA,could result in the minimum required ECCS equipment not beingavailable. Since both a CS subsystem is inoperable and a reduction inthe makeup capability of the LPCI System has occurred, a morerestrictive Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is required to restore either aCS subsystem or, either a LPCI subsystem OR the LPCI pump(s) toOPERABLE status. The Completion Time was developed usingengineering judgment based on a reliability study cited in Reference 11,previous BWR licensing precedents, and approved for Monticello byAmendment 162 (Reference 14). This Completion Time has been foundto be acceptable through operating experience.Monticello B 3.5.1-8 Revision No. 34MonticelloB 3.5.1-8Revision No. 34 ECCS -Operating3.5.1BASESACTIONS (continued)F.1 and F.2If any Required Action and associated Completion Time of Condition A,B, C, D, or E is not met, the plant must be brought to a MODE in whichthe LCO does not apply. To achieve this status, the plant must bebrought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within36 hours. The allowed Completion Times are reasonable, based onoperating experience, to reach the required plant conditions from fullpower conditions in an orderly manner and without challenging plantsystems.G.1If two LPCI subsystems are inoperable due to open RHR intertie returnline isolation valve(s), the RHR intertie line must be isolated within18 hours. The line can be isolated by closing both RHR intertie return lineisolation valves or by closing one RHR intertie return line isolation valveand the RHR intertie suction line isolation valve. The 18 hour2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months CompletionTime is reasonable, considered the low probability of a DBA occurringduring this period.H._1If the Required Action and associated Completion Time of Condition G isnot met, the plant must be brought to a MODE in which the RHR intertiereturn line isolation valves are not required to be closed. To achieve thisstatus, the plant must be brought to at least MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Theallowed Completion Time is reasonable, based on operating experience,to reach the required plant conditions from full power conditions in anorderly manner and without challenging plant systems.1.1 and 1.2If the HPCI System is inoperable and the RCIC System is verified to beOPERABLE, the HPCI System must be restored to OPERABLE statuswithin 14 days. In this condition, adequate core cooling is ensured by theOPERABILITY of the redundant and diverse low pressure ECCSinjection/spray subsystems in conjunction with ADS. Also, the RCICSystem will automatically provide makeup water at most reactor operatingpressures. Verification of RCIC OPERABILITY is therefore requiredimmediately when HPCI is inoperable. This may be performed as anadministrative check by examining logs or other information to determineMonticelloB 3.5.1-9Revision No. 34 ECCS -Operating3.5.1BASESACTIONS (continued)if RCIC is out of service for maintenance or other reasons. It does notmean to perform the Surveillances needed to demonstrate theOPERABILITY of the RCIC System. If the OPERABILITY of the RCICSystem cannot be immediately verified, however, Condition M must beentered. In the event of component failures concurrent with a designbasis LOCA, there is a potential, depending on the specific failures, thatthe minimum required ECCS equipment will not be available. A 14 dayCompletion Time is based on a reliability study cited in Reference 11 andhas been found to be acceptable through operating experience.J.1 and J.2If any one low pressure ECCS injection/spray subsystem, or one LPCIpump in both LPCI subsystems, is inoperable in addition to an inoperableHPCI System, the inoperable low pressure ECCS injection/spraysubsystem(s) or the HPCI System must be restored to OPERABLE statuswithin 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . In this condition, adequate core cooling is ensured bythe OPERABILITY of the ADS and the remaining low pressure ECCSsubsystems. However, the overall ECCS reliability is significantlyreduced because a single failure in one of the remaining OPERABLEsubsystems concurrent with a design basis LOCA may result in the ECCSnot being able to perform its intended safety function. Since both a highpressure system (HPCI) and a low pressure subsystem(s) are inoperable,a more restrictive Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is required to restoreeither the HPCI System or the low pressure ECCS injection/spraysubsystem(s) to OPERABLE status. This Completion Time is based on areliability study cited in Reference 11 and has been found to beacceptable through operating experience.K. 1The LCO requires three ADS valves to be OPERABLE in order to providethe ADS function. Reference 12 contains the results of an analysis thatevaluated the effect of one ADS valve being out of service. Per thisanalysis, operation of only two ADS valves will provide the requireddepressurization. However, overall reliability of the ADS is reduced,because a single failure in the OPERABLE ADS valves could result in areduction in depressurization capability. Therefore, operation is onlyallowed for a limited time. The 14 day Completion Time is based on areliability study cited in Reference 11 and has been found to beacceptable through operating experience.Monticello B 3.5.1-10 Revision No. 34MonticelloB 3.5.1-10Revision No. 34 ECCS -Operating3.5.1BASESACTIONS (continued)L.1 and L.2If any Required Action and associated Completion Time of Condition I, J,or K is not met, or if one ADS valve is inoperable and Condition A, B, C,D, or G are entered, or if two or more ADS valves are inoperable, or if theHPCI System is inoperable and Condition D, E, or G are entered, then theplant must be brought to a condition in which the LCO does not apply. Toachieve this status, the plant must be brought to at least MODE 3 within12 hours and reactor steam dome pressure reduced to < 150 psig within36 hours. The allowed Completion Times are reasonable, based onoperating experience, to reach the required plant conditions from fullpower conditions in an orderly manner and without challenging plantsystems.M._ 1If two or more low pressure ECCS injection/spray systems are inoperablefor reasons other than Conditions C, D, E, or G, the plant is in a degradedcondition not specifically justified for continued operation, and may be in acondition outside of the accident analyses. Therefore, LCO 3.0.3 must beentered immediately.For some cases, per the single failure assumptions of the accidentanalysis the plant may not be in an unanalyzed condition (Ref. 10) but theallowable duration for operation in the condition has not been justified,therefore LCO 3.0.3 must be entered immediately.Monticello B 3.5.1-11 Revision No. 34MonticelloB 3.5.1-11Revision No. 34 ECCS -Operating3.5.1BASESSURVEILLANCE SR 3.5.1.1REQUIREMENTSThe flow path piping has the potential to develop voids and pockets ofentrained air. Maintaining the pump discharge lines of the CS Systemand LPCI subsystems full of water ensures that the ECCS will performproperly, injecting its full capacity into the RCS upon demand. This willalso prevent a water hammer following an ECCS initiation signal. Oneacceptable method of ensuring that the lines are full is to vent at the highpoints. While the potential for developing voids in the HPCI Systemexists, the effects of a void have been analyzed and shown to beacceptable. The 31 day Frequency is based on the gradual nature of voidbuildup in the ECCS piping, the procedural controls governing systemoperation, and operating experience.SR 3.5.1.2Verifying the correct alignment for manual, power operated, andautomatic valves in the ECCS flow paths provides assurance that theproper flow paths will exist for ECCS operation. This SR does not applyto valves that are locked, sealed, or otherwise secured in position sincethese were verified to be in the correct position prior to locking, sealing, orsecuring. A valve that receives an initiation signal is allowed to be in anonaccident position provided the valve will automatically reposition in theproper stroke time. This SR does not require any testing or valvemanipulation; rather, it involves verification that those valves capable ofpotentially being mispositioned are in the correct position. This SR doesnot apply to valves that cannot be inadvertently misaligned, such ascheck valves. For the HPCI System, this SR also includes the steam flowpath for the turbine and the flow controller position.The 31 day Frequency of this SR was derived from the Inservice TestingProgram requirements for performing valve testing at least once every92 days. The Frequency of 31 days is further justified because the valvesare operated under procedural control and because improper valveposition would only affect a single subsystem. This Frequency has beenshown to be acceptable through operating experience.SR 3.5.1.3Verification every 31 days that each ADS pneumatic pressure is withinthe analysis limits (S/RV Accumulator Bank header pressure >_ 88.3 psigand Alternate Nitrogen System supply (ALT N2 TRAIN A (or B) SUPPLY)pressure _ 410 psig (Ref. 13)) ensures adequate pressure for reliableADS operation. The supply associated with each ADS valve providespneumatic pressure for valve-actuation. The design pneumatic supplypressure requirements for the S/RV accumulator bank and AlternateMonticelloB 3.5.1-12Revision No. 34 ECCS -Operating3.5.1BASESSURVEILLANCE REQUIREMENTS (continued)Nitrogen System trains (replaceable gas cylinders) are such that,following a failure of the pneumatic supply to them, at least five valveactuations can occur over a ten hour period (Ref. 10). The ECCS safetyanalysis assumes only one actuation to achieve the depressurizationrequired for operation of the low pressure ECCS. The 31 day Frequencytakes into consideration administrative controls over operation of thesystem and alarms for low pressure.Each Alternate Nitrogen System is designed for the three upstreamnitrogen bottles to maintain OPERABILITY while the fourth, downstream,bottle is being replaced with a fully charged bottle. During bottlechangeout the capacity of the system is temporarily reduced. This isacceptable based on the remaining capacity (only one actuation isnecessary to depressurize), the low rate of usage, the fact thatprocedures have been initiated for replenishment, and the low probabilityof an event during this brief period.SR 3.5.1.4Verification every 31 days that the RHR System intertie return lineisolation valves are closed ensures that each LPCI subsystem will providethe required flow rate to the reactor pressure vessel. The 31 dayFrequency has been found acceptable, considering that these valves areunder strict administrative controls that will ensure the valves continue toremain closed.The SR is modified by a Note stating that the SR is only required to bemet in MODE 1. During MODE 1 operations with the RHR Systemintertie line isolation valves open, some of the LPCI flow may be divertedto the broken recirculation loop during a LOCA, potentially resulting inearly transition boiling. In other MODES, the intertie line may be openedbecause the impact on the LOCA analyses is negligible.SR 3.5.1.5Verification of correct breaker alignment to the LPCI swing busdemonstrates that the normal AC electrical power source is powering theswing bus and the backup AC electrical power source is available toensure proper operation of the LPCI injection valves and the recirculationpump discharge valves. If either the normal source is not powering theLPCI swing bus or the backup source is not available to the LPCI swingbus, one of the LPCI subsystems must be considered inoperable. The31 day Frequency has been found acceptable based on engineeringjudgment and operating experience.MonticelloB 3.5.1-13Revision No. 34 ECCS -Operating3.5.1BASESSURVEILLANCE REQUIREMENTS (continued)SR 3.5.1.6Cycling the recirculation pump discharge valves through one completecycle of full travel demonstrates that the valves are mechanicallyOPERABLE and will close when required. Upon initiation of an automaticLPCI subsystem injection signal, these valves are required to be closed toensure full LPCI subsystem flow injection in the reactor via therecirculation jet pumps. De-energizing the valve in the closed position willalso ensure the proper flow path for the LPCI subsystem. Acceptablemethods of de-energizing the valve include de-energizing breaker controlpower, racking out the breaker or removing the breaker.The Frequency of this SR is in accordance with the Inservice TestingProgram. If any recirculation pump discharge valve is inoperable and inthe open position, both LPCI subsystems must be declared inoperable.SR 3.5.1.7, SR 3.5.1.8, and SR 3.5.1.9The performance requirements of the low pressure ECCS pumps aredetermined through application of the 10 CFR 50, Appendix K criteria(Ref. 7). This periodic Surveillance is performed (in accordance with theASME Operation and Maintenance (OM) Code requirements for theECCS pumps) to verify that the ECCS pumps will develop the flow ratesrequired by the respective analyses. The low pressure ECCS pump flowrates ensure that adequate core cooling is provided to satisfy theacceptance criteria of Reference 9. The pump flow rates are verifiedagainst a system head equivalent to the reactor to containment pressureexpected during a LOCA. In addition, for LPCI the system head for thetested pump must include a head correction that corresponds to two LPCIpumps delivering 7,740 gpm. The total system pump outlet pressure isadequate to overcome the elevation head pressure between the pumpsuction and the vessel discharge, the piping friction losses, and RPVpressure present during a LOCA. These values are establishedanalytically.The flow tests for the HPCI System are performed at two differentpressure ranges such that system capability to provide rated flow againsta system head corresponding to reactor pressure is tested at both thehigher and lower operating ranges of the system. The required systemhead should overcome the RPV pressure and associated discharge linelosses. Adequate reactor steam pressure must be available to performthe tests. Additionally, adequate steam flow must be passing through themain turbine or turbine bypass valves to continue to control reactorpressure when the HPCI System diverts steam flow. Therefore, sufficienttime is allowed after adequate pressure and flow are achieved to performMonticelloB 3.5.1-14Revision No. 34 ECCS -Operating3.5.1BASESSURVEILLANCE REQUIREMENTS (continued)these tests. Reactor steam pressure must be > 950 psig to performSR 3.5.1.8 and > 150 psig to perform SR 3.5.1.9. Adequate steam flow isrepresented by at least one turbine bypass valve 80% open. Reactorstartup is allowed prior to performing the low pressure Surveillance testbecause the reactor pressure is low and the time allowed to satisfactorilyperform the Surveillance test is short. The reactor pressure is allowed tobe increased to normal operating pressure since it is assumed that thelow pressure test has been satisfactorily completed and there is noindication or reason to believe that HPCI is inoperable.Therefore, SR 3.5.1.8 and SR 3.5.1.9 are modified by Notes that state theSurveillances are not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after thereactor steam pressure and flow are adequate to perform the test. The12 hours allowed for performing the flow test after the required pressureand flow are reached is sufficient to achieve stable conditions for testingand provides reasonable time to complete the SRs. The Frequency forSR 3.5.1.7 and SR 3.5.1.8 is in accordance with the Inservice TestingProgram requirements. The 24 month Frequency for SR 3.5.1.9 is basedon the need to perform the Surveillance under the conditions that applyduring a startup from a plant outage. Operating experience has shownthat these components usually pass the SR when performed at the24 month Frequency, which is based on the refueling cycle. Therefore,the Frequency was concluded to be acceptable from a reliabilitystandpoint.SR 3.5.1.10The ECCS subsystems are required to actuate automatically to performtheir design functions. This Surveillance verifies that, with a requiredsystem initiation signal (actual or simulated), the automatic initiation logicof HPCI, CS, and LPCI will cause the systems or subsystems to operateas designed, including actuation of the system throughout its emergencyoperating sequence, automatic pump startup and actuation of allautomatic valves to their required positions. This SR also ensures thatthe HPCI System will automatically restart on a Reactor Vessel WaterLevel -Low Low signal received subsequent to a Reactor Vessel WaterLevel -High trip and that the suction is automatically transferred from theCSTs to the suppression pool on a Suppression Pool Water Level -Highor Condensate Storage Tank Level -Low signal. The LOGIC SYSTEMFUNCTIONAL TEST performed in LCO 3.3.5.1 overlaps this Surveillanceto provide complete testing of the assumed safety function.Monticello B 3.5.1-15 Revision No. 34MonticelloB 3.5.1-15Revision No. 34 ECCS -Operating3.5.1BASESSURVEILLANCE REQUIREMENTS (continued)The 24 month Frequency is based on the need to perform theSurveillance under the conditions that apply during a plant outage and thepotential for an unplanned transient if the Surveillance were performedwith the reactor at power.Operating experience has shown that these components usually pass theSR when performed at the 24 month Frequency, which is based on therefueling cycle. Therefore, the Frequency was concluded to beacceptable from a reliability standpoint.This SR is modified by a Note that excludes vessel injection/spray duringthe Surveillance. Since all active components are testable and full flowcan be demonstrated by recirculation through the test line, coolantinjection into the RPV is not required during the Surveillance.SR 3.5.1.11The ADS designated S/RVs are required to actuate automatically uponreceipt of specific initiation signals. A system functional test is performedto demonstrate that the mechanical portions of the ADS function (i.e.,solenoids) operate as designed when initiated either by an actual orsimulated initiation signal, causing proper actuation of all the requiredcomponents. SR 3.5.1.12 and the LOGIC SYSTEM FUNCTIONAL TESTperformed in LCO 3.3.5.1 overlap this Surveillance to provide completetesting of the assumed safety function.The 24 month Frequency is based on the need to perform theSurveillance under the conditions that apply during a plant outage and thepotential for an unplanned transient if the Surveillance were performedwith the reactor at power. Operating experience has shown that thesecomponents usually pass the SR when performed at the 24 monthFrequency, which is based on the refueling cycle. Therefore, theFrequency was concluded to be acceptable from a reliability standpoint.This SR is modified by a Note that excludes valve actuation since thevalves are individually tested in accordance with SR 3.5.1.12.SR 3.5.1.12This Surveillance verifies that each ADS valve is capable of beingopened, which can be determined by either of two means, i.e., Method 1or Method 2. Applying Method 1, approved in Reference 15, valveOPERABILITY and setpoints for overpressure protection are verified inMonticelloB 3.5.1-16Revision No. 34 ECCS -Operating3.5.1BASESSURVEILLANCE REQUIREMENTS (continued)accordance with the ASME OM Code. Applying Method 2, a manualactuation of the ADS valve is performed to verify the valve is functioningproperly.Method 1Valve OPERABILITY and setpoints for overpressure protection areverified in accordance with the requirements of the ASME OM Code(Ref. 16). Proper ADS valve function is verified through performance ofinspections and overlapping tests on component assemblies,demonstrating the valve is capable of being opened. Testing isperformed to demonstrate that each:* ADS S/RV main stage opens and passes steam when theassociated pilot stage actuates; and* ADS S/RV second stage actuates to open the associated mainstage when the pneumatic actuator is pressurized;* ADS S/RV solenoid valve ports pneumatic pressure to theassociated S/RV actuator when energized;" ADS S/RV actuator stem moves when dry lift tested in-situ.(With exception of main and pilot stages this test demonstratesmechanical operation without steam.)The solenoid valves and S/RV actuators are functionally tested once percycle as part of the Inservice Testing Program. The S/RV assembly isbench tested as part of the certification process, at intervals determined inaccordance with the Inservice Testing Program. Maintenance proceduresensure that the S/RV is correctly installed in the plant, and that the S/RVand associated piping remain clear of foreign material that might obstructvalve operation or full steam flow.This methodology provides adequate assurance that the ADS valves willoperate when actuated, while minimizing the challenges to the valves andthe likelihood of leakage or spurious operation.Method 2A manual actuation of each ADS valve is performed to verify that thevalve and solenoid are functioning properly and that no blockage exists inthe S/RV discharge lines. This is demonstrated by the response of theturbine bypass valves, by a change in the measured flow, or by any othermethod suitable to verify steam flow. Adequate steam flow must beMonticelloB 3.5.1-17Revision No. 34 ECCS -Operating3.5.1BASESSURVEILLANCE REQUIREMENTS (continued)passing through the turbine bypass valves to continue to control reactorpressure when the ADS valves divert steam flow upon opening.Sufficient time is therefore allowed after the required flow is achieved toperform this SR. Adequate steam flow is represented by at least oneturbine bypass valve 80% open. This SR is modified by a Note thatstates the Surveillance is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months afterreactor steam flow is adequate to perform the test. Reactor startup isallowed prior to performing this SR because valve OPERABILITY and thesetpoints for overpressure protection are verified, per ASMErequirements, prior to valve installation. The 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowed for manualactuation after the required flow is reached is sufficient to achieve stableconditions and provides adequate time to complete the Surveillance.SR 3.5.1.11 and the LOGIC SYSTEM FUNCTIONAL TEST performed inLCO 3.3.5.1, "ECCS Instrumentation," overlap this Surveillance to providecomplete testing of the assumed safety function.The Frequency of "In accordance with the Inservice Testing Program" isbased on ASME OM Code requirements. Industry operating experiencehas shown that these components usually pass the SR when performedat the Code required Frequency. Therefore, the Frequency wasconcluded to be acceptable from a reliability standpoint.SR 3.5.1.13The LPCI System injection valves, recirculation pump discharge valves,recirculation pump suction valves, and the RHR discharge intertie lineisolation valves are powered from the LPCI swing bus, which must beenergized after a single failure, including loss of power from the normalsource to the swing bus. Therefore, the automatic transfer capability fromthe normal power source to the backup power source must be verified toensure the automatic capability to detect loss of normal power and initiatean automatic transfer to the swing bus backup power source. Verificationof this capability every 24 months ensures that AC electrical power isavailable for proper operation of the associated LPCI injection valves,recirculation pump discharge valves, recirculation pump suction valves,and the RHR discharge intertie line isolation valves. The swing busautomatic transfer scheme must be OPERABLE for both LPCIsubsystems to be OPERABLE. The Frequency of 24 months is based onthe need to perform the Surveillance under the conditions that applyduring a startup from a plant outage. Operating experience has shownthat the components usually pass the SR when performed at the 24month Frequency, which is based on the refueling cycle. Therefore, theFrequency was concluded to be acceptable from a reliability standpoint.MonticelloB 3.5.1-18Revision No. 34 ECCS -Operating3.5.1BASESREFERENCES 1. USAR, Section 6.2.2.2. USAR, Section 6.2.3.3. USAR, Section 6.2.4.4. USAR, Section 6.2.5.5. USAR, Section 14.7.2.6. USAR, Section 14.7.3.7. 10 CFR 50, Appendix K.8. USAR, Section 6.2.1.1.9. 10 CFR 50.46.10. USAR, Section 14.7.2.3.2.11. Memorandum from R.L. Baer (NRC) to V. Stello, Jr. (NRC),"Recommended Interim Revisions to LCOs for ECCS Components,"December 1, 1975.12. USAR, Section 14.7.2.3.1.5.13. Amendment No. 155, "Issuance of Amendment Re: Request toRevise Technical Specification Surveillance Requirement 3.5.1.3 toCorrect the Alternate Nitrogen System Pressure," dated February 21,2008. (ADAMS Accession Nos. ML080380638 and ML080590541)14. Amendment No. 162, "Issuance of Amendment RegardingCompletion Time to Restore a Low-Pressure Emergency CoreCooling Subsystem to Operable Status," dated July 10, 2009.(ADAMS Accession No. ML091480782)15. Amendment No. 168, "Issuance of Amendment Re: Testing of MainSteam Safety/Relief Valves," dated July 27, 2012. (ADAMSAccession No. ML12185A216)16. ASME Operation and Maintenance (OM) Code.17. Amendment No. 176, "Monticello Nuclear Generating Plant -Issuance of Amendment No. 176 to Renewed Facility OperatingLicense Regarding Extended Power Uprate," (ADAMS AccessionNo. ML13316C459)MonticelloB 3.5.1-19Revision No. 34 ECCS -Operating3.5.1BASESREFERENCES (continued)18. Amendment No. 184, "Monticello Nuclear Generating Plant -Issuance of Amendment to Revise Technical Specification 3.5.1,"ECCS [Emergency Core Cooling System] -Operating," datedNovember 3, 2014. (ADAMS Accession No. ML14246A449)[Condition F previously allowed two Core Spray subsystems to beinoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .]MonticelloB 3.5.1-20 -LastRevision No. 34 OPERATING LICENSE AND TECHNICAL SPECIFICATIONSUPDATING INSTRUCTIONSTECHNICAL SPECIFICATIONSREMOVE INSERTPage Amendment Document Type Page AmendmentTable 1 186 Operating Table 1 186License andTS LEFPTable 2 186 Record of TS Table 2 U186'.Changes and ...IOL Amend.2.0-1 165 Specification 2.0-1, .1852.03.5.1-2 162 Specification 3 5 1 2 184 .3.5.1-3 176 351 35113 1843.5.1-4 176 3514 184(Do not insert in TS binder, only an updating aid.)NOTE: This update also removes an existing TS "Flag" for Condition F for "Both Core Spraysubystems inoperable" since Amendment 184 removes this Condition and re-letters theconditions, eliminating the need for the flag.(Adds Am. 184 and 185)MonticelloAm 186 TABLE I (Page 1 of 3)MONTICELLO NUCLEAR GENERATING PLANTOPERATING LICENSE AND TECHNICAL SPECIFICATIONS LIST OF EFFECTIVE PAGESPaieiiiii1.1-11.1-21.1-31.1-41.1.51.1-61.2-11.2-21.3-11.3-21.3-31.3-41.3-51.3-61.3-71.3-81.3-91.3-101.3-111.4-11.4-21.4-31.4-41.4-51.4-6Am. No.(2)(2)(2)146148146176175146146146146146146146146146146146146146146146146146158146146Page3.1.4-23.1.4-33.1.5-13.1.5-23.1.5-33.1.6-13.1.6-23.1.7-13.1.7-23.1.7-33.1.7-43.1.7-53.1.7-63.1.8-13.1.8-23.2.1-13.2.2-13.2.2-23.2.3-13.3.1.1-13.3.1.1-23.3.1.1-33.3.1.1-43.3.1.1-53.3.1.1-63.3.1.1-73.3.1.1-83.3.1.1-93.3.1.1-103.3.1.2-13.3.1.2-23.3.1.2-33.3.1.2-43.3.1.2-53.3.2.1-13.3.2.1-23.3.2.1-33.3.2.1-43.3.2.1-53.3.2.2-13.3.2.2-23.3.3.1-13.3.3.1-23.3.3.1-33.3.3.2-13.3.4.1-13.3.4.1-23.3.4.1-3Am. No.146158146146146146146148146146146146146146146146146146146171176180180180180180180180180146146146146146146146159159173146146146146146146146146146Page3.3.5.1-13.3.5.1-23.3.5.1-33.3.5.1-43.3.5.1-53.3.5.1-63.3.5.1-73.3.5.1-83.3.5.1-93.3.5.1-103.3.5.1-113.3.5.2-13.3.5.2-23.3.5.2-33.3.5.2-43.3.6.1-13.3.6.1-23.3.6.1-33.3.6.1-43.3.6.1-53.3.6.1-63.3.6.1-73.3.6.2-13.3.6.2-23.3.6.2-33.3.6.3-13.3.6.3-23.3.6.3-33.3.7.1-13.3.7.1-23.3.7.1-33.3.7.2-13.3.7.2-23.3.8.1-13.3.8.1-23.3.8.1-33.3.8.2-13.3.8.2-23.3.8.2-33.4.1-13.4.1-23.4.2-13.4.3-13.4.3-23.4.4-13.4.4-23.4.5-1Am. No.146146146146151176146176161146146146146146146146146146146176146164146146146146146146Pane3.4.5-23.4.6-13.4.6-23.4.7-13.4.7-23.4.8-13.4.8-23.4.9-13.4.9-23.4.9-33.4.10-13.5.1-13.5.1-23.5.1-33.5.1-43.5.1-53.5.1-63.5.1-73.5.2-13.5.2-23.5.2-33.5.3-13.5.3-23.6.1.1-13.6.1.1-23.6.1.2-1Am. No.146148148146146146146172172172146146184184184162(7)1671681461461671461461461461461.4-7 1462.0-1 1853.0-13.0-23.0-33.0-43.0-53.1.1-13.1.1-23.1.1-33.1.2-13.1.2-23.1.3-13.1.3-23.1.3-33.1.3-43.1.4-11571461571461461461461461461461461581581581461481481481481481461461471461461461591591461461681461461463.6.1.2-2 1463.6.1.2-33.6.1.2-43.6.1.3-13.6.1.3-23.6.1.3-33.6.1.3-43.6.1.3-53.6.1.3-63.6.1.3-73.6.1.3-83.6.1.4-13.6.1.5-13.6.1.5-23.6.1.6-13.6.1.6-2146146148146148146180148146176146146168146146Am. 186 TABLE 1 (Page 2 of 3)MONTICELLO NUCLEAR GENERATING PLANTOPERATING LICENSE AND TECHNICAL SPECIFICATIONS LIST OF EFFECTIVE PAGESPage Am. No. Page Am. No. Page Am. No. OperatingLicense3.6.1.7-1 146 3.8.2-2 148 3.10.6-1 146 Cover pgi.3.6.1.7-2 146 3.8.2-3 146 3.10.6-2 146 1 1563.6.1.8-1 146 3.8.3-1 178 3.10.7-1 146 2 1563.6.1.8-2 146 3.8.3-2 178 3.10.7-2 146 3 1863.6.2.1-1 146 3..3-3 146 3.10.8-1 146 4 1863.6.2.1-2 146 3.8.4-1 146 3.10.8-2 146 5 1563.6.2.1-3 146 3.8.4-2 153(8) 3.10.8-3 146 6 1763.6.2.2-1 146 3.8.4-3 146 4.0-1 182 7 176(')3.6.2.3-1 146 3.8.5-1 148 4.0-2 182 8 1763.6.2.3-2 146 3.8.5-2 146 5.1-1 146 9 1763.6.3.1-1 146 3.8.6-1 146 5.2-1 146 10 1763.6.4.1-1 146 3.8.6-2 146 5.2-2 163 11 1693.6.4.1-2 146 3.8.6-3 146 5.3-1 146 12 1603.6.4.2-1 146 3.8.6-4 146 5.4-1 1463.6.4.2-2 146 3.8.7-1 146 5.5-1 1463.6.4.2-3 146 3.8.7-2 146 5.5-2 146 OL App. A3.6.4.3-1 146 3.8.8-1 148 5.5-3 146 Cover pg.3.6.4.3-2 181 3.8.8-2 146 5.5-4 1463.7.1-1 146 3.9.1-1 146 5.5-5 1813.7.1-2 146 3.9.1-2 146 5.5-6 181 Op. License3.7.2-1 146 3.9.2-1 146 5.5-7 181 Appendix C3.7.2-2 146 3.9.3-1 146 5.5-8 146 C-i 1753.7.3-1 146 3.9.4-1 146 5.5-9 146 C-2 1023.7.4-1 175 3.9.4-2 146 5.5-10 176 (6) C-3 1753.7.4-2 160 3.9.5-1 146 5.5-11 160 C-4 1103.7.4-3 181 3.9.6-1 146 5.5-12 1823.7.5-1 154 3.9.7-1 146 5.5-13 1823.7.5-2 154 3.9.7-2 146 5.6-1 1463.7.6-1 146 3.9.8-1 146 5.6-2 1803.7.6-2 146 3.9.8-2 146 5.6-3 1803.7.7-1 146 3.10.1-1 174 5.7-1 1463.7.8-1 146 3.10.1-2 146 5.7-2 1463.8.1-1 146 3.10.1-3 146 5.7-3 1463.8.1-2 146 3.10.2-1 146 5.7-4 1463.8.1-3 146 3.10.2-2 1463.8.1-4 146 3.10.3-1 1463.8.1-5 146 3.10.3-2 1463.8.1-6 146 3.10.3-3 1463.8.1-7 146 3.10.4-1 1463.8.1-8 146 3.10.4-2 1463.8.1-9 146 3.10.4-3 1463.8.1-10 146 3.10.5-1 1463.8.2-1 148 3.10.5-2 146Am. 186 TABLE 1 (Page.3 of 3)MONTICELLO NUCLEAR GENERATING PLANTOPERATING LICENSE AND TECHNICAL SPECIFICATIONS LIST OFEFFECTIVE PAGESNOTESNOTE 1: (Removed)NOTE 2: Am. 152 removed Table of Contents (TOC) from NRC issued Tech. Specs. TOO at Revision 0.NOTE 3: (Removed)NOTE 4: (Removed)NOTE 5: (Removed)NOTE 6: NRC Correction Letter to Amendment 176, dated 12/16/2013, corrected spelling of "gage" to,.gauge" in OL License Condition 15(a). Added previously issued Amendment No. 175 (struckthrough) to bottom of page 5.5-10 and removed old revision bar.NOTE 7: SR 3.5.1.3.b (Alternate Nitrogen System supply pressure to ADS) is annotated and is beingtreated as a non-conservative TS. AN2 pressure is increased as specified within Ops. ManualB.08.04.03-05 as an interim measure.NOTE 8: The required 125 VDC charger amperage in SR 3.8.4.2 Option 1 of 50 amps is annotated. Theinterim required amperage is 75 amps. This condition is being treated as a non-conservative TS(see AR 01456839).Am. 186 TABLE 2 (Page 1 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNSP PageRevision(REV) No.LicenseDPR-22Amend No. & DateAEC Tech SpecChanqe IssuanceNo. & DateMajor SubjectOriginal123456781 1/19/71Note 2Note 2Note 22 2/20/73Note 2Note 2Note 2Note 2Note 2Note 2Note 2Note 234 6/17/746 8/20/74Note 3578910 7/8/7512111314Note 12 1/14/723 10/31/724 12/8/72Note 15 3/2/731 4/28/71 &6 4/3/737 5/4/738 7/2/739 8/24/7310 10/2/7311 11/27/73&12 11/15/7313 3/30/7414 5/14/74Note 1Note 1Note 3 10/24/7415 1/15/7516 2/3/7517 2/26/7518 4/10/75Note 120 9/15/7519 9/17/7521 10/6/7522 10/30/75Appendix A Technical Specificationsincorporated in DPR-22 on 9/8/70Removed 5 MWt restrictionMOGS Technical Specification changesissued by AEC but never distributed or putinto effect, superseded by TS Change 1211/15/73RHR service water pump capability changeTemporary surveillance test waiverIncrease in U-235 allowed in fission chambersMiscellaneous Technical Specification changes,Respiratory Protection, & AdministrativeControl ChangesRespiratory Protection ChangesRelief Valve and CRD Scram Time ChangesFuel Densification LimitsSafety Valve Setpoint ChangeOffgas Holdup System, RWM, andMiscellaneous Changes8x8 Fuel Load Authorization8x8 Full Power authorizationChanged byproduct material allowanceChanged byproduct material allowanceInverted Tube (CRD) LimitsREMP ChangesReactor Vessel Surveillance Program ChangesVacuum Breaker Test ChangesCorrects Errors & Provides ClarificationIncreased allowed quantity of U-235Snubber RequirementsRemoved byproduct material allowanceSuppression Pool Temperature LimitsAppendix K and GETAB Limits910111213141516171819Am. 186 TABLE 2 (Page 2 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNSP PageRevision(REV) No.2021222324252627282930313233343536373839NOTE 14041424344454647LicenseDPR-22Amend No. & Date15 1/22/76 NOTE 416 2/3/7617 3/16/76NOTE 3 4/13/7618 4/14/7621 5/20/7619 5/27/7620 6/18/7622 7/13/7623 9/27/7624 10/15/7625 10/27/7626 4/1/7727 5/24/7728 6/10/7729 9/16/7730 9/28/7731 10/14/7732 12/9/7733 1/25/7834 4/14/7835 9/15/7836 10/30/7837 11/6/78NOTE 3 11/24/7838 3/15/7939 5/15/7940 6/5/7941 8/29/79Reporting RequirementsCRD Collet Failure SurveillanceNSP Organization ChangesAdoption of GETABContainment Isolation Valve TestingInterim Appendix B, Section 2.4 Tech. Specs.Low Steamline Pressure Setpoint and MCPR ChangesAPLHGR, LHGR, MCPR LimitsCorrection of Errors and Environmental ReportingStandby Gas Treatment System SurveillanceCRD Test FrequencySnubber Testing ChangesAPRS Test Method.MAPLHGR Clamp at Reduced FlowRadiation Protection Supervisor QualificationREMP ChangesMore Restrictive MCPRInservice Inspection ChangesReporting RequirementsFire Protection RequirementsIncrease in spent fuel storage capacityRPT RequirementsSuppression Pool Surveillance8x8R Authorization, MCPR Limits & SRV SetpointsCorrected Downcomer SubmergenceIncorporation of Physical Security Plan into LicenseRevised LPCI Flow CapabilityRespiratory Protection Program ChangesFire Protection Safety Evaluation ReportMawor SubiectAm. 186 TABLE 2 (Page 3 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNSP PageRevision(REV) No.LicenseDPR-22Amend No. & DateMajor Subject4849424312/28/792/12/8050NOTE 1NOTE 15152NOTE 1NOTE 15354NOTE 144 2/29/80-8/29/80-9/19/80-10/24/80-1/9/81-1/9/81-1/13/811 2/12/812 3/2/81MAPLHGR vs. Exposure TableMCPR & MAPLHGR Changes for Cycle 8 and Extended CoreBurnupILRT RequirementsOrder for Modification of License-EnvironmentalQualificationRevised Order for Modification of License-EnvironmentalQualificationOrder for Modification of License-EnvironmentalQualification RecordsIssuance of Facility Operating License (FTOL)Order for Modification of License Concerning BWRScram Discharge Systems (License conditions removed perAmendment No. 11 dated 10/8/82)Order for Modification Mark I ContainmentRevision of License Conditions Relating to FireProtection ModificationsTMI Lessons Learned & Safety -Related Hydraulic Snubber AdditionsLow voltage protection, organization and miscellaneousIncorporation of Safeguards Contingency Plan and SecurityForce Qualification and Training Plan into LicenseCycle 9 -ODYN Changes, New MAPLHGR's, RPSResponse time changeInservice Inspection ProgramFire Protection Technical Specification ChangesMark I Containment Modifications343/27/813/27/81555657585 5/4/816786/3/816/30/8111/5/8159NOTE 19 12/28/81-1/19/82Inservice Surveillance Requirements for SnubbersRevised Order for ModificationMark I ContainmentScram Discharge VolumeNew Scram Discharge Volumes606110115/20/8210/8/82Am. 186 TABLE 2 (Page 4 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNSP PageRevision(REV) No.LicenseDPR-22Amend No. & DateMajor Subject6263646566676869707172737475767778798081828384858687888990121314151617181920212223242526272829303132333435363738394011/30/8212/6/8212/10/8212/17/824/18/834/17/8311/28/8312/30/831/16/841/23/842/2/844/3/845/1/848/15/849/24/8410/31/8411/2/8411/16/8411/16/8411/27/845/28/8510/7/8510/8/8512/3/8512/23/851/22/862/12/863/13/863/18/86RPS Power MonitorCycle 10Recirc Piping and Coolant Leak DetectionAppendix I Technical Specifications (removed App. B)Organizational ChangesMiscellaneous ChangesSteam Line Temperature Switch SetpointRadiation Protection ProgramSRM Count RateDefinition of OperabilityMiscellaneous Technical Specification ChangesRPS Electrical Protection Assembly Time DelayScram Discharge Volume Vent and Drain ValvesMiscellaneous Technical Specification ChangesCycle 11RHR Intertie Line AdditionHybrid I Control Rod AssemblyARTSLow Low Set LogicDegraded Voltage Protection LogicSurveillance RequirementsScreen Wash/Fire Pump (Partial)Fuel Enrichment LimitsCombustible Gas Control SystemVacuum Breaker CyclingNUREG-0737 Technical SpecificationsEnvironmental Technical SpecificationsAdministrative ChangesClarification of Radiation Monitor RequirementsAm. 186 TABLE 2 (Page 5 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNSP Page License Major SubiectRevision DPR-22(REV) No. Amend No. & Date91 41 3/24/86 250 Volt Battery92 42 3/27/86 Jet Pump Surveillance93 43 4/8/86 Simmer Margin Improvement94 44 5/27/86 Cycle 12 Operation95 45 7/1/86 Miscellaneous Changes96 46 7/1/86 LER Reporting and Miscellaneous Changes97 47 10/22/86 Single Loop Operation98 48 12/1/86 Offgas System Trip99 49 8/26/87 Rod Block Monitor100 50 8/26/87 APRM and IRM Scram Requirements101 51 10/16/87 2R Transformer102 52 11/18/87 Surveillance Intervals -ILRT Schedule103 53 11/19/87 Extension of Operating License104 54 11/25/87 Cycle 13 and Misc Changes105 55 11/25/87 Appendix J Testing106 56 12/11/87 ATWS -Enriched Boron107 57 9/23/88 Increased Boron Enrichment108 58 12/13/88 Physical Security Plan109 59 2/16/89 Miscellaneous Administrative Changes110 60 2/28/89 Miscellaneous Administrative Changes111 61 3/29/89 Fire Protection and Detection System112 62 3/31/89 ADS Logic and S/RV Discharge Pipe Pressure113 63 4/18/89 Miscellaneous Technical Specification Improvements114 64 5/10/89 Containment Vent and Purge Valves115 65 5/30/89 NUREG-0737 -Generic Letter 83-36116 66 5/30/89 Reactor Vessel Level Instrumentation117 67 6/19/89 Extension of MAPLHGR. Exposure for One Fuel Type118 68 7/14/89 SRO Requirements & Organization Chart RemovalAm. 186 TABLE 2 (Page 6 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNSP PageRevision(REV) No.LicenseDPR-22Amend No. & Date119120121122123124125126127128125130131132133134135136137138139140141142143697071727374757677787980818283848586879/12/899/28/8910/19/8911/2/895/1/906/5/9010/12/9012/20/902/15/913/28/914/9/918/12/914/16/927/15/928/18/921/27/936/29/937/12/934/15/94Maior SubjectOperations Committee Quorum RequirementsRelocation of Cycle-Specific Thermal-Hydraulic LimitsDeletion of Primary Containment Isolation Valve TableRG 1.99, Rev 2, ISI & ILRTCombined STA/LSO PositionRemoval of WRGM Automatic ESF ActuationDiesel Fuel Oil StorageMiscellaneous Administrative ChangesRedundant and IST TestingAlarming DosimetrySAFER/GESTRTorus Vacuum Breaker Test Switch/EDG Fuel Oil Tank LevelSurveillance Test Interval Extension -Part IAlternate Snubber Visual Inspection IntervalsRevisions to Reactor Protection System Tech SpecsMELLIA and Increase Core FlowRevision to Diesel Fire Pump Fuel Oil Sampling RequirementsRevisions to Control Rod Drive Testing RequirementsRevised Coolant Leakage Monitoring FrequencyAverage Planar Linear Heat Generation Rate (APLHGR)Specification & Minimum Critical Power Ratio Bases RevisionsRemoval of Chlorine Detection Requirements andChanges to Control Room Ventilation System RequirementsRevisions to Radiological Effluent SpecificationsSecondary Containment System and Standby GasTreatment System Water Level Setpoint ChangeChange in Safety Relief Valves Testing RequirementsRevised Core Spray Pump Flow88 6/30/9489 8/25/94909192939/7/949/9/949/15/947/12/95Am. 186 TABLE 2 (Page 7 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNSP PageRevision(REV) No.LicenseDPR-22Amend No. & DateMajor Subject1441451461471489495969710/2/954/3/964/9/969/17/9698 7/25/9799 10/29/9711/25/97149NOTE 5150NOTE 61001 00a1014/20/984/30/9808/28/98102 09/16/98103 12/23/98Standby Gas Treatment and Secondary Containment SystemsMSIV Combined Leakrate, and Appendix J, Option BPurge and Vent Valve Seal Replacement IntervalImplementation of BRWOG Option I-D core Stability Solutionand re-issue of pages 11, 12, 82 and 231 to reflect pagesissued by NRC amendments.Bases changes on containment overpressure and numberof RHR pumps required to be operable. Reissue pages 207,209, 219, 229k, 229p, 230, 245 to reflect pages issued byNRC amendments.SLMCPR for Cycle 18 and reissue pages vi, 155, 202, 207,219, 229uReissue pages a, b, g, iii, vi, 14, 25a, 155, 198y, 198z, 202,207, 209, 219, 229k, 229p, 229r, 229u, 230, 245SLMCPR for Cycle 19Reissue all pages.Reactor Coolant Equivalent Radioiodine Concentration andControl Room HabitabilityMonticello Power RerateSurveillance Test Interval/Allowed Outage Time ExtensionProgram -Part 2Revision of Statement on Shift Length & other Misc ChangesCST Low Level HPCI/RCIC Suction TransferRevised RPV-PT Curves & remove SBLC RV setpointReactor Pressure Vessel Hydrostatic and Leakage TestingTesting Requirements for Control Room EFT FiltersSafety Limit Minimum Critical Power Ratio for Cycle 20Transfer of Operating Authority from NSP to NMCTransfer of Operating License from NSP to a New UtilityOperating CompanyEmergency Filtration Train Testing Exceptions andTechnical Specification RevisionsAlternate Shutdown System Operability RequirementsSafety/Relief Valve Bellows Leak Detection SystemTest Frequency10410510610710810911011112/24/9803/19/9910/12/9911/24/9912/8/9902/16/0008/07/0008/18/00112 08/18/00113 10/02/00114 11/30/00Am. 186 TABLE 2 (Page 8 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNSP PageRevision(REV) No.LicenseDPR-22Amend No. & DateMaior Subiect115115a12/21/0002/13/01116 03/01/01117 03/07/01118 03/09/01118a 05/10/01119119a12004/05/0106/28/0107/24/01121 07/25/01122 08/01/01122a 10/22/01123 10/26/01123a 10/25/01124 10/30/01124a 12/05/01Administrative Controls and Other Miscellaneous ChangesBases Change to Reflect Modification 98Q145 InstalledControl Room Toxic Gas Air SupplyRelocation of Inservice Inspection Requirements to aLicensee ProgramReactor Water Cleanup (RWCU) System AutomaticIsolation and Miscellaneous Instrumentation System ChangesRevision of Standby Liquid Control System SurveillanceRequirementsBases Change -50'F Loop Temperature, Bus Transfer &Rerate CorrectionFire Protection Technical Specification ChangesBases Change -Added information on cooldown rateRelocation of Radiological Effluent Technical Specificationsto a Licensee-Controlled ProgramClarify air ejector offgas activity sample point and operabilityrequirementsRelocation of Inservice Testing Requirements to a Licensee-Controlled ProgramBases Change -Remove scram setpoints sentence andcorrect typoControl Rod Drive and Core Monitoring Technical SpecificationChangesBases Change -Change to reflect new operation ofdrywell to suppression chamber vacuum breaker valveposition indicating lightsRelocation of Technical Specification Administrative ControlsRelated to Quality Assurance PlanBases Change -Change to reflect revised TechnicalSpecification definition of a containment spray/coolingsubsystemSafety Limit Minimum Critical Power Ratio for Cycle 21Elimination of Local Suppression Pool Temperature LimitsBases Change -Change reflects relocation of samplepoint for the offgas radiation monitor125126126a12/06/0101/18/0202/15/02Am. 186 TABLE 2 (Page 9 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNSP PageRevision(REV) No.LicenseDPR-22Amend No. & Date127 05/31/02128 06/11/02128a 07/11/02129 08/27/02Major Subject129a129b09/12/0209/12/02130 09/23/02130a 09/26/02131 10/02/02Missed Surveillance Requirement TechnicalSpecification Changes (TSTF-358)Changes to the Technical Specifications RevisedReference Point for Reactor Vessel Level Setpoints,Simplification of Safety Limits, and Improvement to the BasesBases Change -Correct Drywell to SuppressionChamber Vacuum Breaker Indicating LightDescriptionRevise Technical Specifications and SurveillanceRequirements Relating to Standby Diesel GeneratorsBases Change -Change to Snubber Operability DescriptionBases Change -Remove Language That ImpliesTrip Settings Can Be Modified By Deviation ValuesContainment Systems Technical Specification ChangesBases Change -HPCI -Change Wording / HPCI & RCIC -Enhance with Wording Consistent with NUREG-1433-Rev 1Update the Multiplier Values for Single Loop OperationAverage Planar Linear Heat Generation Rate (APLHGR)Conversion to Option B for Containment Leak Rate TestingRevision to Pressure-Temperature CurvesBases Change -Adequate Reactor Steam Flow forHPCI/RCIC TestingOne-Time Extension of Containment Integrated Leak-Rate Test IntervalBoiling Water Reactor Vessel and Internals ProjectReactor Pressure Vessel Integrated SurveillanceProgramBases Change -Clarify description of head coolingGroup 2 valvesElimination of Requirements for Post Accident SamplingSystem (TSTF-413)Bases Change -Editorial change to define theabbreviation "EFCV."Drywell Leakage and Sump Monitoring Detection System132133133a02/04/0302/24/0303/28/03134 03/31/03135 04/22/03135a 04/24/03136 06/17/03136a 09/25/03137 08/21/03Am. 186 TABLE 2 (Page 10 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNSP PageRevision(REV) No.LicenseDPR-22Amend No. & DateMajor Subject137a137b10/09/0310/14/03138 05/21/04138a 06/10/04139 06/02/04140140a11/02/0401/13/05141 01/28/05141a 02/24/05141b 03/10/05141c 03/10/05142 02/01/05143 09/30/0510/20/05Bases Change -RCS Leakage Requirements for TS 3.6.4.DBases Change -Clarification of system controlboundary for ASDSElimination of Requirements for Hydrogen Recombiners andHydrogen and Oxygen Monitors (TSTF-447)Bases Change -Clarification of Tech Spec Table 4.1.1Manual ScramRevised Analysis of Long-Term Containment Responseand Net Positive Suction Head (Design Bases and USAR change)Revision to Technical Specification Tables 3.2.1 and 3.2.4Bases Change -Removal of Drywell Vent Coolers from3.6/4.6 BasesRevision to Technical Specifications Table 3.2.3 andSection 3.7/4.7Bases Change -Implement Improved BPWS asDescribed in NEDO-33091-ABases Change -Bases Changes for LicenseAmendments 138 and 140Bases Change -Removal of 3% Delta-K from StandbyLiquid Control Bases 3.4.A/4.4.ADeletion of Requirements for Submittal of OccupationalRadiation Reports, Monthly Operating Reports, and Report ofSafety/Relief Valve Failures and Challenges (TSTF-369)Implementation of 24-Month Fuel CyclesChange to Facility Operating License Pursuant toCommission Order EA-03-086 Regarding RevisedDesign Basis Treat (DBT); and Revisions to PhysicalSecurity Plan, Training and Qualification Plan, andSafeguards Contingency PlanSurveillance Test Intervals for various instruments(Second part of 24-month Fuel Cycle amendment.)TS Bases changes to conform with the implementation ofLicense Amendments 143 and 144 (24-Month Fuel Cycles).Alternate Source Term -Fuel Handling Accident (TSTF-51)NOTE 1:144 01/12/06144a 04/05/06145 04/24/06Am. 186 TABLE 2 (Page 11 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNSP Page LicenseRevision DPR-22(REV) No. Amend No. & Date146 06/05/06Correction Letter 10/12/06147 07/05/06NOTE 1: Renewed OL 11/08/06148 12/07/06Appendix J Exemptions149 01/18/07Correction Letter 02/23/07150 03/09/07151 07/20/07Major SubjectImproved Technical Specifications (TSTF-359, 372, 439,455, 458,460, 464, 479,480, 485)Correction of typo in Amendment 146 (page 3.3.5.1-8)Degraded Voltage Allowable Value Change (Second part ofITS -Follow-on ITS Amendment)Renewed Facility Operating LicenseAlternate Source Term -Full ScopeIn conjunction with issuance of Amendment 148, exemptions to10 CFR 50.54(o) and to 10 CFR 50, Appendix J, Option B,Sections III.A and III.B were issued.One-Time Extension of LPCI Loop Select Logic Time DelayRelay Surveillance IntervalCorrection: Remove Am 148 from first page of OL and addedRenewed License No. DPR-22 to footer on page 2.Increase SFP allowed Heat Load and allow installation of 64 cellPaR Fuel Storage Rack (if required) to maintain Full Core Offloadcapability during ISFSI construction.Extend Surveillance Interval from 92-days to 24-months andincrease Allowable Values for LPCI Loop Select LogicTime Delay Relays. (Also, correct typo on page 3.3.5.1-6.)Conforming License Amendment to incorporate the MitigationStrategies Required by Section B.5.b of Commission OrderEA-02-026 and the Radiological Protection Mitigation StrategiesRequired by Commission Order EA-06-137Remove the Table of Contents (TOC) out of the TechnicalSpecifications and place under licensee control. TS TOC initialrevision is Revision 0.Revise Surveillance Requirement 3.8.4.2 to specify that theDivision 2 battery chargers are verified to supply greater than orequal to 110 amps.Add Action Statement for two inoperable Control RoomVentilation subsystems to Specification 3.7.5 (TSTF-477).Revise Surveillance Requirement 3.5.1.3.b to specify that theAlternate Nitrogen System supply pressure to the ADS valves isverified to be greater than or equal to 410 psig.Transfer of Operating License from NMC to NSP -MinnesotaAdd LCO 3.0.9 on unavailable Barriers (TSTF-427).Revise Control Rod notch testing frequency from once per 7 daysto every 31 days (TSTF-475).Power Range Neutron Monitoring SystemAm. 186NOTE 1:08/23/07152 11/08/07153 01/30/08154 01/23/08155 02/21/08NOTE 1: 15615715809/22/0810/22/0811/19/08159 1/30/09 TABLE 2 (Page 12 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNSP PageRevision(REV) No.LicenseDPR-22Amend No. & DateCorrection Letter 3/9/09160 3/17/09161 4/7/09162 7/10/09163 8/19/09164 9/28/09165 5/4/11166 7/29/11167 1/11/12168 7/27/12169 8/27/12170 9/7/12171 1/25/13172 1/20/13Major SubiectCorrect footer of Specification 3.3.7.2 -AST Amendment 148.Control Room Envelope Habitability (TSTF-448).Revise the Allowable Value and channel calibration frequency forTable 3.3.5.1-1, Function 2.j, Recirculation Riser DifferentialPressure -High (Break Detection).Add new Conditions to Specification 3.5.1 for restoration ofvarious low-pressure ECCS subsystems out-of-service (OOS)combinations (e.g., one low-pressure ECCS division OOS).Deleted paragraph d concerning work hour limitations underSpecification 5.2.2 to align with rule changes under 10 CFR 26,Subpart I (TSTF-511).Corrected Modes in Table 3.3.6.1-1, Function 5.d, RWCU Systemisolation on a SLC System initiation, to match SLC Systemmodes (Specification 3.1.7) after adoption of full-scope AST, i.e.,added Mode 3 to Function 5.dRevise MCPR Safety Limit to 1.15 to reflect reload analyses(which include EPU and MELLLA+ considerations).Add Cyber Security Plan license condition under OL Section 3,Physical Protection.Revise Core Spray flowrate from 2800 gpm to 2835 gpm.Revise surveillance requirements in Specifications 3.4.3, 3.5.1and 3.6.1.5 to remove requirements to lift-test SRVs during plantstartup.Revise licensing basis to reflect removal of the capability toautomatically transfer to the 1AR Transformer as a source ofpower to the essential buses on degraded voltage and insteaddirectly transfer to the EDGs. (Operating License change only.)Revise Table 3.3.5.1-1, Functions 1.e and 2.e, "Reactor SteamDome Pressure Permissive -Bypass Timer (Pump Permissive)",(i.e., 20 minute ADS bypass timer), to remove the lower limit ofthe allowable value.Revise Required Actions table for LCO 3.3.1.1 to provide arestoration period before entering the required actions when theAPRMs are inoperable due to SR 3.3.1.1.2 not met.Revise Specification 3.4.9, add new Specification 5.6.5, and adda new definition to specify the adoption of a PTLR.Am. 186 TABLE 2 (Page 13 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNSP PageRevision(REV) No.LicenseDPR-22Amend No. & Date173 7/15/13174 8/9/13175 8/28/13176 12/9/13Correction Letter 12/16/13(to Amendment 176)177 1/28/14178 1/28/14179 2/28/14180 3/28/14181 5/2/14182 10/24/14183 10/31/14184 11/3/14185 11/25/14186 11/28/14AEC Tech Spec Major SubjectChange IssuanceNo. & DateAdd footnote reflecting that RWM can be bypassed when theimproved BPWS is used for reactor shutdown (TSTF-476).Add allowance to Specification 3.10.1 LCO to allow for scramtime testing during inservice leak testing and hydrostatic testing(TSTF-484).Miscellaneous Operating License and TS editorial andadministrative changes.Extended Power UprateCorrected spelling of "gage" to "gauge" in OL LicenseCondition 15(a). Added previously issued Amendment No. 175(struckthrough) to bottom of page 5.5-10 and removed oldrevision bar.Modifies the Emergency Plan to revise the EAL for the TurbineBuilding Normal Waste Sump Monitor.Revise specification to reflect relocating fuel and lube oil requiredvolumes to the TS Bases and replacing them with duration basedrequirements (TSTF-501).Revises Shutdown Margin (SDM) definition to address advancedfuel designs (TSTF-535).Maximum Extended Load Line Limit Analysis, Plus (MELLLA+)Reduces SBGT and CREF Systems runtime from 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months to15 minutes every 31 days. Removes surveillance requirementsto have heaters operating and TS 5.5.6.e electric heater outputtest requirement from the Ventilation Filter Testing Program(TSTF-522).Revise TS to Support Fuel Storage System ChangesRemove Radwaste Operator as 60-Minute Responder(Emergency Plan change only.)Revise TS 3.5.1 to Remove Condition F allowing two Core Spraysubsystems to be OOS.Reduce the Reactor Steam Dome Pressure Specified in theReactor Core Safety Limits, resolves GE PROF analysis Part 21issue.Revise Cyber Security Implementation Schedule MilestonesAm. 186 TABLE 2 (Page 14 of 14)MONTICELLO NUCLEAR GENERATING PLANTRECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTSNOTES1. License Amendment or Order for Modification of License not affecting Technical Specifications.2. Technical Specification change issued prior to 10 CFR revisions which require issuance of TechnicalSpecification changes as License Amendments.3. Modification to Bases. No Technical Specification change or License Amendment issued.4. Technical Specification change numbers no longer assigned beginning with Amendment 15.5. Pages reissued 11/25/97 to conform with NRC version. Revision number of affected pages not changed.6. All pages reissued using INTERLEAF in different font. Using NRC Amendment Nos. and issue date. ForBases and Table of Contents, spelling errors corrected and editorial corrections made and all AmendmentNos. changed to 1 00a. For remaining Tech Spec pages, no other changes made and current AmendmentNos. used.Am. 186 SLs2.02.0 SAFETY LIMITS (SLs)2.1 SLs2.1.1 Reactor Core SLs2.1.1.1 With the reactor steam dome pressure < 686 psig or core flow< 10% rated core flow:THERMAL POWER shall be _< 25% RTP.2.1.1.2 With the reactor steam dome pressure _> 686 psig and core flow_ 10% rated core flow:MCPR shall be _> 1.15 for two recirculation loop operation or >_ 1.15 forsingle recirculation loop operation.2.1.1.3 Reactor vessel water level shall be greater than the top of activeirradiated fuel.2.1.2 Reactor Coolant System Pressure SLReactor steam dome pressure shall be _ 1332 psig.2.2 SL VIOLATIONSWith any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s:2.2.1 Restore compliance with all SLs; and2.2.2 Insert all insertable control rods.Monticello2.0-1Amendment No. 185 ECCS -Operating3.5.1ACTIONS (continued)CONDITION REQUIRED ACTION COMPLETION TIMEC. One LPCI pump in both C.1 Restore one LPCI pump to 7 daysLPCI subsystems OPERABLE status.inoperable.D. Two LPCI subsystems D.1 Restore one LPCI 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months sinoperable for reasons subsystem to OPERABLEother than Condition C status.or G.E. One Core Spray E.1 Restore Core Spray 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months subsystem inoperable, subsystem to OPERABLEstatus.AND OROne LPCI subsystem E.2 Restore LPCI subsystem to 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months sinoperable. OPERABLE status.OR OROne or two LPCI E.3 Restore LPCI pump(s) to 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months spump(s) inoperable. OPERABLE status.F. Required Action and F.1 Be in MODE 3. 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months sassociated CompletionTime of Condition A, B, ANDC, D, or E not met.F.2 Be in MODE 4. 36 hour4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months sG. Two LPCI subsystems G.1 Isolate the RHR intertie 18 hour2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months sinoperable due to open line.RHR intertie return lineisolation valve(s).Monticello3.5.1-2Amendment No. !46, !62, 184 ECCS -Operating3.5.1ACTIONS (continued)CONDITION REQUIRED ACTION COMPLETION TIMEH. Required Action and H.1 Be in MODE 2. 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months sassociated CompletionTime of Condition G notmet.HPCI System 1.1 Verify by administrative Immediatelyinoperable, means RCIC System isOPERABLE.AND1.2 Restore HPCI System to 14 daysOPERABLE status.J. HPCI System J.1 Restore HPCI System to 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months sinoperable. OPERABLE status.AND ORCondition A, B, or C J.2 Restore low pressure 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months sentered. ECCS injection/spraysubsystem(s) toOPERABLE status.K. One ADS valve K.1 Restore ADS valve to 14 daysinoperable. OPERABLE status.IIIMonticello3.5.1-3Amendment No. 146, 162, 17-6, 184 ECCS -Operating3.5.1ACTIONS (continued)CONDITION REQUIRED ACTION COMPLETION TIMEL. Required Action and L.1 Be in MODE 3. 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months sassociated CompletionTime of Condition I, J, ANDor K not met.L.2 Reduce reactor steam 36 hour4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months sOR dome pressure to< 150 psig.One ADS valveinoperable andCondition A, B, C, D, orG entered.ORTwo or more ADS valvesinoperable.ORHPCI System inoperableand Condition D, E, or Gentered.M. Two or more low M.1 Enter LCO 3.0.3. Immediatelypressure ECCSinjection/spraysubsystems inoperablefor reasons other thanCondition C, D, E, or G.ORHPCI System and one ormore ADS valvesinoperable.IMonticello3.5.1-4Amendment No. 146, 162, 176, 184

ML15110A280

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