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| | issue date = 09/07/1982 | | | issue date = 09/07/1982 |
| | title = Documents Discussions on Request for License Amend Granting Relief from Requirements of Tech Spec 3.5.2 Re Safety Injection Pump Out of Svc.Safety Evaluation Prepared in Conjunction W/Westinghouse Supports Relief Request | | | title = Documents Discussions on Request for License Amend Granting Relief from Requirements of Tech Spec 3.5.2 Re Safety Injection Pump Out of Svc.Safety Evaluation Prepared in Conjunction W/Westinghouse Supports Relief Request |
| | author name = JURGENSEN R W | | | author name = Jurgensen R |
| | author affiliation = AMERICAN ELECTRIC POWER SERVICE CORP. | | | author affiliation = AMERICAN ELECTRIC POWER SERVICE CORP. |
| | addressee name = DENTON H R, VARGA S A | | | addressee name = Denton H, Varga S |
| | addressee affiliation = NRC, NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR) | | | addressee affiliation = NRC, NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR) |
| | docket = 05000315 | | | docket = 05000315 |
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| {{#Wiki_filter:~~ANEPICANELECTR1CPGWERSe>viceCorpo~etio)s2Broadu:ay. | | {{#Wiki_filter:~ ~ |
| Xeu:Pork,.l.}.Ir)00.t(2/2p.S>0.9000September 7,1982AEP:NRC:0739 DonaldC.CookNuclearPlantUnitNo.1DocketNo.50-315LicenseNo.DPR-58REQUESTFORRELIEFFROMTECHNICAL SPECIFICATION 3.5.2Mr.HaroldR.Denton,DirectorOfficeofNuclearReactorRegulation U.S.NuclearRegulatory Commission Washington, D.C.20555Attention: | | ANEPICAN ELECTR1C PGWER Se> vice Corpo~ etio)s 2 Broadu:ay. Xeu: Pork,.l. }. Ir)00.t (2 /2p .S > 0. 9000 September 7, 1982 AEP:NRC:0739 Donald C. Cook Nuclear Plant Unit No. 1 Docket No. 50-315 License No. DPR-58 REQUEST FOR RELIEF FROM TECHNICAL SPECIFICATION 3.5.2 Mr. Harold R. Denton, Director Of fice of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Mr. Steven A. 'Varga |
| Mr.StevenA.'Varga | |
|
| |
|
| ==DearMr.Denton:== | | ==Dear Mr. Denton:== |
| Thisletterdocuments thediscussions heldwithmembersofyourStaffconcerning ourrequestforalicenseamendment grantingrelieffromtherequirements of.Technical Specification 3.5.2.EachUnitoftheCookNuclearPlanthastwoSafetyInjection PumpsplustwoCentrifugal ChargingPumpsandtwoResidualHeatRemovalPumps.Inpreparation fortheblackouttestperformed onAugust14,1982,UnitNo.1'sNorthSafetyInjection Pumpsufferedseveredamage.We",haveproceeded torepairthepumpexpeditiously buthavenotbeenabletocompletetherepairsasofthetimeofthisletter.UnitNo.1Technical Specification (T/S)No.3.5.2requires, amongotherthings,thattwoSafetyInjection (SI)PumpsbeoperableinModes1,2and3.UnitNo.1iscurrently
| | |
| 'completing itsrefueling outageandthereturntoModes3,2and1isprevented bytheinoperability oftheSIpumpandtheT/Smentioned above.Inordertoproceedwiththelowpowerphysicstesting,werequestawaiveroftherequirement tohavetwoSIpumpsoperableinModes2(reactorthermalpowerlessthanorequalto5%)and3foraperiodofoneweekbeginning atthetimetheUnitentersMode3.TheSouthSIpumpwillbedemonstrated operablepriortoenteringMode3.Allremaining technical specification requirements willbemet.Attachment 1tothislettercontainsthesafetyevaluation preparedbyusinconjunction withWestinghouse. | | This letter documents the discussions held with members of your Staff concerning our request for a license amendment granting relief from the requirements of .Technical Specification 3.5.2 . |
| Theconclusion isthatevenifweweretolosetheoperable1SSafetyInjection Pumpsufficient marginKP wy,@5"-&~9'M
| | Each Unit of the Cook Nuclear Plant has two Safety Injection Pumps plus two Centrifugal Charging Pumps and two Residual Heat Removal Pumps. |
| ~-'a wrMr.H.R.Dent.AEP:NRC:0739wouldstillexisttothelimitsspecified in10CFR50.46.Thus,thisreliefisnotdetrimental tothehealthandsafetyofthepublic.IWewouldappreciate theexpeditious handlingofthisrequestbyyourStaff.AEPSCinterprets 10CFR170.22asrequiring thataClassIIIAmendment Feebepaidforthechange.Acheckintheamountof$4,000willbetransmitted toyouinafutureletter.hThisTechnical Specification reliefrequesthasbeenreviewedbytheCookPlantPNSRC.ItwillbereviewedbyAEPSC'sNSDRCatthenext~scheduled meeting.Duetothisletterbeingwrittenonshortnotice,ithasnotbeenpreparedfollowing ourstanda"dCorporate Procedures forsuchletters.Weshall,however,reviewtheletteraccording toourCorporate Procedures andwil'nform youifanymodification isrequired. | | In preparation for the blackout test performed on August 14, 1982, Unit No. 1's North Safety Injection Pump suffered severe damage. We |
| Verytrulyyours,/emcAttachment cc:JohnE.Dolan-ColumbusR.S.HunterM.P.AlexichW.G.Smith,Jr.-BridgmanR.C.CallenG.CharnoffJoeWilliams, Jr.NRCResidentInspector atCookPlant-Bridgman'~~
| | ",have proceeded to repair the pump expeditiously but have not been able to complete the repairs as of the time of this letter. Unit No. 1 Technical Specification (T/S) No. 3.5.2 requires, among other things, that two Safety Injection (SI) Pumps be operable in Modes 1, 2 and 3. |
| Attachment toAEP:NRC:0739 LOCAEvaluation forD.C.CookUnit1withOneSafetInectionPumOutofServiceThepurposeofthisevaluation istoassesstheeffectofonesafetyinjection pumpoutofservicefortheCookUnit1NuclearPlantonLossofCoolantAccident(LOCA)consequences. | | Unit No. 1 is currently 'completing its refueling outage and the return to Modes 3, 2 and 1 is prevented by the inoperability of the SI pump and the T/S mentioned above. In order to proceed with the low power physics testing, we request a waiver of the requirement to have two SI pumps operable in Modes 2 (reactor thermal power less than or equal to 5%) and 3 for a period of one week beginning at the time the Unit enters Mode 3. |
| Presently, theplantisfueledbyExxonNuclearCompany.However,theevaluation providedbelowisjudgedtobeapplicable tothenon'Westinghouse fuel,sincetherearenoknownmajordesigndifferences thatwouldhaveasignificant impactontheLOCAbehaviorimportant forthisevaluation. | | The South SI pump will be demonstrated operable prior to entering Mode |
| LareBreakLOCASafetyinjection pumpflowprovidesaninsignificant proportion ofthetotalECCSflowduringalargebreakaccident, whereRCSpressurerapidlydropstonearatmospheric.
| | : 3. All remaining technical specification requirements will be met. |
| Accumulator andlowheadsafetyinjection (RHR)flowareimportant forthisaccident. | | Attachment 1 to this letter contains the safety evaluation prepared by us in conjunction with Westinghouse. |
| Therefore, thelossofasafetyinjection pumphasanegligible effectonlargeLOCAcalculated peakcladtemperature. | | we were to lose the operable 1S Safety The conclusion is that even if K |
| SmallBreakLOCATheplant'sprotection againstsmallLOCAscomesfromatwotrainsystemincluding atotaloftwosafetyinjection pumpsandtwohighheadchargingpumps.SmallLOCAFSARlicensing analysesassumetheworstsinglefailuretobelossofatrain,leavingoneintermediate headSIpumpandonechargingpump.ThesmallLOCAanalysisyieldscladtemperatures wellbelow10CFR50.46limits.Thisanalysisassumption boundsthepresentplantconfiguration withonesafetyinjection pumpoutofserviceandnosinglefailure.Iftheworstsinglefailureassumption isconsidered inadditiontothe;lossofthesafetyinjection pump,andfurther,thetrainlostisassumed'tohavetheoperational safetyinjection pump,ECCSflowisdelivered fromonlythehighheadchargingpump.Thefollowing paragraphs evaluatethisscenario.
| | Injection Pump sufficient margin P |
| Reduction ofECCSflowintherangeof600to1200psiahasanadverse--"effect oncalculated cladtemperature forarangeofsmallLOCAbreaksizes.Thelossofasafetyinjectionpumphastheeffectofreducingdelivered ECCSflowinthatimportant pressurerange.TotalECCSflowwillbedegradedbyapproximately 56%averagedoverthispressureinterval. | | |
| Established sensitivity studieshaveindicated thatsuchadegradation resultsinasmuchasa550oFsmallLOCAPCTincrease. | | wy, |
| ~~0t~.ThesmallbreakanalysisforCook1doesnotusethelatestNRCapprovedWsmallLOCAEvaluation Model.ThecurrentsmallbreakLOCAEMwouldcalculate aPCTofapproximately 1200oF,reducedfrom1493F,predicted bytheanalysisintheFSAR.ThisnewPCTisestablished fromanalysisofasubstantially equivalent plant(3250MWt,4Loop,sameSISdesign)analyzedinHCAP-8970-P-A, | | @5"-&~9'M~ 'a |
| ,"Westinghouse Emergency CoreCo'olingSystemSmall8reakOctober,1975Model",andappliestoCook.Additionally, creditforconservative assumptions inthesmallLOCAFSARanalysiscanmitigatethePCTpenalty.Also,thecurrentsmallLOCAFSARanalysisisperformed at100%powerwhileduringtheperiodoftimeforwhichweareseekingT/Sreliefthereactorwillnotexceed5%power.Thelowerpower.levelof5%versus100%wouldmorethanoffsettheconsequences ofreducedECCSflow.Inconclusion, operation ofCook1withasafetyinjection pumpoutofserviceandamaximumpowerlevelof5%forabrief'periodoftimeisstillboundedbythesmallLOCAFSARanalysis. | | |
| Inaddition,thefactthatthepresentanalysishassignificant marginto10CFR50.46PCTlimitsindicates thatstartupoftheplantfromthisrefueling outageisnotasafetyconcern.}}
| | wr Mr. H. R. Dent . AEP: NRC: 0739 would still exist to the limits specified in 10 CFR 50.46. Thus, this relief is not detrimental to the health and safety of the public. |
| | I We would appreciate the expeditious handling of this request by your Staff. |
| | AEPSC interprets 10 CFR 170.22 as requiring that a Class III Amendment Fee be paid for the change. A check in the amount of $ 4,000 will be transmitted to you in a future letter.h This Technical Specification relief request has been reviewed by the Cook Plant PNSRC. It will be reviewed by AEPSC's NSDRC at the next |
| | ~ |
| | scheduled meeting. |
| | Due to this letter being written on short notice, it has not been prepared following our standa"d Corporate Procedures for such letters. |
| | We shall, however, review the letter according to our Corporate Procedures and wil'nform you if any modification is required. |
| | Very truly yours, |
| | /emc Attachment cc: John E. Dolan - Columbus R. S. Hunter M. P. Alexich W. G. Smith, Jr. Bridgman R. C. Callen ' ~ ~ |
| | G. Charnoff Joe Williams, Jr. |
| | NRC Resident Inspector at Cook Plant - Bridgman |
| | |
| | Attachment to AEP:NRC:0739 LOCA Evaluation for D. C. Cook Unit 1 with One Safet In ection Pum Out of Service The purpose of this evaluation is to assess the effect of one safety injection pump out of service for the Cook Unit 1 Nuclear Plant on Loss of Coolant Accident (LOCA) consequences. Presently, the plant is fueled by Exxon Nuclear Company. However, the evaluation provided below is judged to be applicable to the non 'Westinghouse fuel, since there are no known major design differences that would have a significant impact on the LOCA behavior important for this evaluation. |
| | Lar e Break LOCA Safety injection pump flow provides an insignificant proportion of the total ECCS flow during a large break accident, where RCS pressure rapidly drops to near atmospheric. Accumulator and low head safety injection (RHR) flow are important for this accident. Therefore, the loss of a safety injection pump has a negligible effect on large LOCA calculated peak clad temperature. |
| | Small Break LOCA The plant's protection against small LOCAs comes from a two train system including a total of two safety injection pumps and two high head charging pumps. Small LOCA FSAR licensing analyses assume the worst single failure to be loss of a train, leaving one intermediate head SI pump and one charging pump. The small LOCA analysis yields clad temperatures well below 10 CFR 50.46 limits. This analysis assumption bounds the present plant configuration with one safety injection pump out of service and no single failure. |
| | If the worst single fai lure assumption is considered in addition to the |
| | ;loss of the safety injection pump, and further, the train lost is assumed |
| | 'to have the operational safety injection pump, ECCS flow is delivered from only the high head charging pump. The following paragraphs evaluate this scenario. |
| | Reduction of ECCS flow in the range of 600 to 1200 psia has an adverse |
| | --"effect on calculated clad temperature for a range of small LOCA break sizes. The loss of a safety i njecti on pump has the effect of reducing delivered ECCS flow in that important pressure range. Total ECCS flow will be degraded by approximately 56% averaged over this pressure interval. |
| | Established sensitivity studies have indicated that such a degradation results in as much as a 550oF small LOCA PCT increase. |
| | |
| | ~ |
| | 0 |
| | ~ |
| | t ~ . |
| | The small break analysis for Cook 1 does not use the latest NRC approved W small LOCA Evaluation Model. The current small break LOCA EM would calculate a PCT of approximately 1200oF, reduced from 1493 F, predicted by the analysis in the FSAR. This new PCT is established from analysis of a substantially equivalent plant (3250 MWt,4 Loop, same SI S design) analyzed in HCAP-8970-P-A, ,"Westinghouse Emergency Core Co'oling System Small 8reak October, 1975 Model", and applies to Cook. |
| | Additionally, credit for conservative assumptions in the small LOCA FSAR analysis can mitigate the PCT penalty. Also, the current small LOCA FSAR analysis is performed at 100% power while during the period of time for which we are seeking T/S relief the reactor will not exceed 5% power. |
| | The lower power. level of 5% versus 100% would more than offset the consequences of reduced ECCS flow. |
| | In conclusion, operation of Cook 1 with a safety injection pump out of service and a maximum power level of 5% for a brief 'period of time is still bounded by the small LOCA FSAR analysis. In addi tion, the fact that the present analysis has significant margin to 10 CFR 50.46 PCT limits indicates that startup of the plant from this refueling outage is not a safety concern.}} |
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Forwards LER 99-004-01 Re Failure to Perform TS Surveillance Analyses of Reactor Coolant Chemistry with Fuel Removed. Commitments Made by Util Are Listed
ML17326A0281999-06-28028 June 1999
Provides Response to 981116 & 960228 RAIs Re GL 92-01. Revised Pressurized Thermal Shock Evaluation Based on New Weld Chemistry Info & Copy of W Rept WCAP-15074, Evaluation of 1P3571 Weld Metal from Surveillance Programs... Encl
ML17326A0241999-06-23023 June 1999
Responds to GL 98-01,Suppl 1, Y2K Readiness of Computer Sys at Nuclear Power Plants. Y2K Readiness Disclosure for Plant & List of Commitments Encl
ML17326A0121999-06-18018 June 1999
Forwards LER 99-014-00 Re Requirement of TS 4.0.5 Not Met for Boron Injection Tank Bolting.Commitments Identified in Submittal Listed
ML17326A0111999-06-11011 June 1999
Provides Response to NRC RAI Re GL 97-01, Degradation of Crdm/Cedm Nozzle & Other Vessel Closure Head Penetrations.
ML17325B6401999-06-0101 June 1999
Forwards LER 99-013-00 Re Safety Injection & Centrifugal Charging Throttle Valve Cavitation During LOCA Could Lead to ECCS Pump Failure.Listed Commitments Identified in Submittal
ML17325B6281999-06-0101 June 1999
Forwards LER 99-S03-00,re Nonconforming Vital Area Barriers.Commitments Made by Util Are Listed
ML17325B6331999-05-28028 May 1999
Forwards LER 99-S02-00,re Vulnerability in Safeguard Sys That Could Allow Unauthorized or Undetected Access to Protected Area.Commitments Made by Util Are Listed
ML17265A8201999-05-24024 May 1999
Forwards LER 98-037-01,representing Extensive Rev to Original LER & Replacing Rept in Entirety.Listed Commitments Identified in Submittal
ML17325B6111999-05-21021 May 1999
Forwards Annual Radioactive Effluent Release Rept for 980101-1231 for DC Cook Nuclear Plant,Units 1 & 2. Transmittal of Submittal Was Delayed Due to Administrative Error in Regulatory Affairs Dept
ML17325B6031999-05-21021 May 1999
Provides Response to NRC GL 98-04, Potential for Degradation of ECCS & Containment Spray Sys After LOCA Because of Const & Protective Coating Deficiencies & Foreign Matl in Containment.
ML17325B5971999-05-20020 May 1999
Forwards LER 99-012-00,re Auxiliary Building ESF Ventilation Sys Not Being Capable of Maintaining ESF Room Temps post-accident.Commitment,listed
ML17335A5281999-05-12012 May 1999
Forwards DC Cook Nuclear Plant Fitness for Duty Program Performance Dtd for six-month Period of 980701-1231,IAW 10CFR26.71(d).Info Was Delayed Due to Administrative Error in Regulatory Affairs Dept
ML17335A5271999-05-11011 May 1999
Forwards Details Re Sources & Levels of Insurance Maintained for DC Cook,Units 1 & 2,as of 990401,per 10CFR50.54(w)(3). Info Was Delayed Beyond Required Date Due to Internal Oversight
ML17325B5841999-05-10010 May 1999
Forwards LER 99-002-00 Re TS 4.0.5 Requirements Not Being Met Due to Improperly Performed Test.Commitments Identified in Ler,Listed
ML17325B5871999-05-0707 May 1999
Forwards Current Revs of Expanded Sys Readiness Review (Essr) Implementing Procedures,For Info Purposes to Support Current NRC Insps.Current Esrr Schedule Provided for Info Purposes,Reflecting Revised Target Dates
ML17325B5821999-05-0404 May 1999
Provides Addl Background,Description & Clarification of Previous & Revised Commitments Re UFSAR Revalidation Effort. Commitment Change Involved Alignment of UFSAR Revalidation Program Methodology to Strategy Contained in Current Plan
ML17325B5791999-05-0404 May 1999
Forwards LER 99-011-00,concerning Air Sys for EDG Not Supporting Long Term Operability.Commitments Made by Util Listed
ML17325B5741999-05-0303 May 1999
Forwards LER 99-010-00 Re RCS Leak Detection Sys Sensitivity Not in Accoradnce with Design Requirements.Listed Commitments Identified in Submittal
ML17325B5631999-04-22022 April 1999
Forwards Results of Independent Chemical Evaluations Performed from Sept 1997 Through Feb 1999,re Resolution of Issues Related to License Amend 227
ML17325B5561999-04-16016 April 1999
Forwards LER 99-006-00, Fuel Crane Loads Lifted Over SFP Could Impact Energies Greater than TS Limits, IAW 10CFR50.73.Submittal Was Delayed to Allow for Resolution of Questions.Commitment Made by Licensee,Listed
ML17325B5451999-04-12012 April 1999
Forwards LER 99-009-00 Re as-found RHR Safety Relief Valve Lift Setpoint Greater than TS Limit.Commitments Identified in Submittal Listed
ML17325B5301999-04-0707 April 1999
Forwards LER 99-S01-01, Vulnerability in Locking Mechanism of Four Vital Area Gates, Per 10CFR50.73.Commitments Made by Util,Listed
ML17325B5241999-04-0505 April 1999
Forwards Revs 0 & 1 to Cook Nuclear Plant Restart Plan, Dtd 980307 & 0407.Rev 5 Is Current Cook Nuclear Plant Restart & Supercedes Previous Revs in All Respects
ML17325B5121999-04-0101 April 1999
Forwards LER 99-007-00, Calculations Show That Divider Barrier Between Upper & Lower Containment Vols May Be Overstressed. Commitments Made by Util Are Listed
ML17325B5141999-03-30030 March 1999
Forwards Rept on Status of Decommissioning Funding.Attached Rept Includes Amount of Decommissioning Funds Estimated to Be Required Pursuant to 10CFR50.75(b) & (C)
ML17325B5191999-03-29029 March 1999
Forwards LER 99-001-00,re Degraded Component Cw Flow to Containment Main Steam Line Penetrations.Commitment, Listed
ML20204F6401999-03-19019 March 1999
Responds to NRC 981013 NOV & Proposed Imposition of Civil Penalty.Violations Cited in Subject NOV Were Initially Identified in Referenced Five Insp Repts.Corrective Actions: Ice Condensers Have Been Completely Thawed of Any Blockage
ML17325B4751999-03-18018 March 1999
Forwards LER 99-004-00,re Failure to Perform TS Surveillance Analyses of Reactor Coolant Chemistry with Fuel Removed. Commitment Made by Util,Listed
ML17325B4721999-03-18018 March 1999
Forwards LER 99-005-00,re Reactor Trip Breaker Manual Actuations During Rod Drop Testing Not Previously Reported. Listed Commitments Identified in Submittal
ML17325B4641999-03-17017 March 1999
Withdraws Response to Issue 1 of NRC Cal,Dtd 970919. Comprehensive Design Review Effort in Progress to Validate Resolution of Issue for Future Operation
1999-09-30
[Table view]
Category:UTILITY TO NRC
MONTHYEARML17328A4481990-09-21021 September 1990
Requests Withdrawal of Mode 6 Proposed Tech Spec Change Since Issue Currently Being Addressed by New STSs
ML17328A4331990-08-27027 August 1990
Informs of Preliminary Assessment of 900713 Electrical Contact Accident at Facility.Investigation Concludes That No Safety Rules Violated
ML17328A4001990-08-24024 August 1990
Responds to 900725 Ltr Re Commitments Made in Response to Generic Ltr 88-14 on Instrument Air Supply Problems Affecting safety-related Equipment.Simulator Training to Be Provided as Stated
ML17334B3781990-08-24024 August 1990
Forwards Info to Certify Plant Simulator Facility.Results of Evaluation of Dual Unit Simulation Facilities Demonstrate That Plant Simulator Performance Compares Favorably W/Units
ML17328A4231990-08-21021 August 1990
Forwards Under Separate Cover,Semiannual Radioactive Effluent Release Rept for Jan-June 1990
ML17328A3901990-08-17017 August 1990
Responds to NRC 900720 Ltr Re Violations Noted in Onsite Audit of Spds.Corrective Actions:Apparent Disparity in Selection of Reactor Trip & Sys Capacity or Anticipatory Values Rectified
ML17328A3891990-08-15015 August 1990
Forwards Performance Data Sheets for Plant fitness-for-duty Program for Period Jan-June 1990,per 10CFR26.Encl Includes Statistics on Various Categories of Testing,Substances Tested for & cut-off Levels Used
ML17328A3571990-08-0202 August 1990
Responds to Open Items in Safety Evaluation of Util Response to Unresolved Issues on post-fire Safe Shutdown Methodology. Open Items 1 & 21 Will Be Closed by Implementing Plant Procedures Providing Equivalent Degree of Protection
ML17328A3431990-07-23023 July 1990
Requests Withdrawal of 890830 Proposed Tech Spec Changes Re Sections 3.0 & 4.0.Util Will Resubmit Generic Ltr 87-09 Recommended Improvement Items on Case by Case Basis
ML17328A3421990-07-23023 July 1990
Provides Results of Offsite Dose Calculation for Reactor Coolant Pump Locked Rotor Event for Facility Cycle 8.Util Identified Previously Issued SERs Addressing Short Term Containment Analysis & LOCA Containment Integrity
ML20055G7551990-07-18018 July 1990
Responds to NRC Re Violations Noted in Insp Repts 50-315/90-10 & 50-316/90-10.Corrective Actions:Required Review Performed & Updated Procedure 12 Mhp 5021.019.001 Revised & Issued
ML17328A3191990-07-12012 July 1990
Responds to NRC 900601 Ltr Re Violations Noted in Insp Repts 50-315/89-31 & 50-316/89-31.Corrective Actions:Electrical Testing Techniques Improved & Surveillance Procedures for Feedwater Pumps Will Be Rewritten
ML17328A3181990-07-0909 July 1990
Responds to Generic Ltr 90-03,Suppl 1, Relaxation of Staff Position in Generic Ltr 83-28,Item 2.2,Part 2, 'Vendor Interface for Safety-Related Components.' Licensee Will Contact safety-related Vendors on Annual Basis
ML17328A3111990-07-0303 July 1990
Provides Certification of Funding Plan for Decommissioning of Plant,Per 10CFR50.33 & 50.75
ML17328A2961990-06-25025 June 1990
Forwards Response to Generic Ltr 90-04 Re Closeout of Generic Safety Issues.Util Supports Concern & Desire to Close long-standing Generic Safety Issues
ML17328A2921990-06-22022 June 1990
Forwards WCAP-12483, Analysis of Capsule U from American Electric Power Co DC Cook Unit 1 Reactor Vessel Radiation Surveillance Program.
ML20044A3521990-06-22022 June 1990
Submits Info Re Sensitivity Study Performed on Number of Fuel Axial Intervals,Per Topical Rept, American Electric Power Reactor Core Thermal-Hydraulic Analysis Using Cobra III-C/MIT-2 Computer Code.
ML17328A2821990-06-15015 June 1990
Submits Ltr Re Proposed Control Room Habitability Tech Spec Changes & Supporting Analyses,Per 900521 Discussion.Revised Calculations of Control Room Thyroid Doses Will Be Submitted within 60 Days of Receipt of Proposed Generic Ltr
ML17328A2741990-06-12012 June 1990
Submits Followup,Per 900205 & 0308 Ltrs & Provides Update Re Inoperable Fire Barrier.Fire Seal Repaired & Restored to Operability on 900419
ML17328A2551990-06-0505 June 1990
Forwards Addl Info Re Util 900126 Revised Response to NRC Bulletin 88-002,per NRC 900509 Request
ML17328A7361990-06-0101 June 1990
Forwards Addl Info Re 890825 & 1212 Applications for Amends to Licenses DPR-58 & DPR-74,per Request.Amends Make Changes to Administrative Controls
ML17328A7331990-05-29029 May 1990
Forwards Nonproprietary WCAP-12577 & Proprietary WCAP-12576, Westinghouse Revised Thermal Design Procedure Instrument... Methodology for American Electric Power DC Cook Unit 2 Nuclear Power Station, Per 900419 Commitment
ML17328A7321990-05-24024 May 1990
Responds to NRC Bulletin 90-001, Loss of Fill Oil in Transmitters Mfg by Rosemount. No Rosemount Model 1153 Series B or D or Model 1154 Transmitters Installed at Facility.Transmitters Purchased as commercial-grade Units
ML17328A6991990-05-0909 May 1990
Provides Suppl to 900103 Response,Certifying That Fitness for Duty Program Implemented at Plant.Change Also Clarifies When More Stringent cut-off Levels Adopted at Plant Program Apply
ML17328A7001990-05-0909 May 1990
Responds to NRC 900406 Ltr Re Inadequacies of Spds,Per Audit on 900221-22.Corrective Actions:Software Mod Will Prevent User from Accessing Displays Other than Iconic Displays from SPDS Dedicated Terminal
ML17328A7071990-05-0707 May 1990
Forwards Monthly Operating Rept for Apr 1990 & Corrected Rept Mar 1990 for Donald C Cook Nuclear Plant Unit 1
ML17328A7031990-05-0707 May 1990
Forwards Monthly Operating Rept for Apr 1990 & Corrected Rept for Mar 1990 for DC Cook Nuclear Plant Unit 2
ML17328A6951990-04-30030 April 1990
Certifies That Training Programs for Initial Licensing & Requalification Training of Operators & Senior Operators at Plant Accredited by Inpo,Per Generic Ltr 87-07
ML17328A6861990-04-30030 April 1990
Forwards Annual Environ Operating Rept,Jan-Dec 1989, & DC Cook Nuclear Plant Units 1 & 2 Operational Radiological Environ Monitoring Program 1989 Rept Jan-Dec 1989.
ML17328A6801990-04-23023 April 1990
Responds to NRC Bulletin 88-004, Potential Safety-Related Pump Loss. Study Planned to Evaluate Alternative Actions to Be Taken for Protecting RHR Pumps
ML17328A6771990-04-23023 April 1990
Forwards DC Cook Nuclear Plant 1990 Annual Emergency Preparedness Exercise. Exercise Scope & Objectives & Detailed Scenario Documentation Encl
ML17334B3641990-04-11011 April 1990
Responds to NRC 900301 Ltr Re Violations Noted in Insp Repts 50-315/89-31 & 50-316/89-31.Corrective Actions:Reliability Centered Maint Program Initiated
ML17328A6601990-04-11011 April 1990
Forwards Updated QA Program Description for Cook Nuclear Plant, Incorporating Editorial,Organizational & Position Title Changes
ML17328A6551990-04-0909 April 1990
Responds to NRC 900309 Ltr Re Violations Noted in Insp Rept 50-316/90-08.Corrective Actions:Procedure for Steam Generator Stop Valve Operability Test Revised to Require That MSIV Be Declared Inoperable After Failing Stroke Test
ML17334B3601990-04-0606 April 1990
Forwards Revised Figures for Loss of Load Event Previously Submitted in Attachment 4,App B of Vantage 5 Reload Transition Safety Rept Supplied by Westinghouse in Jan 1990. Update Is Editorial in Nature
ML17334B3621990-04-0606 April 1990
Forwards 1989 Annual Rept & Projected Cash Flow
ML17328A6341990-03-30030 March 1990
Forwards Application for Renewal of Plant NPDES Permit,For Info,Per Section 3.2 of App B to Plant Tech Specs.W/O Encl
ML17328A6331990-03-30030 March 1990
Responds to NUMARC 900104 Request for Addl Info Re Station Blackout Submittals.Target Reliability for Emergency Diesel Generator of 0.0975 Established for 4 H ac-independent Coping Category Will Be Maintained
ML17328A6321990-03-30030 March 1990
Forwards Response to Generic Ltr 90-01, NRC Regulatory Impact Survey. Util Ack Recommendation by NUMARC
ML17328A6221990-03-27027 March 1990
Responds to 900226 Ltr Transmitting Notice of Violation & Proposed Imposition of Civil Penalty in Amount of $75,000. Corrective Actions:Flow Retention Circuitry Setpoints Set to Compensate for Missized Process Flow Orifice
ML17328A6191990-03-27027 March 1990
Forwards Payment in Amount of $75,000 for Civil Penalty Imposed Through Notice of Violation Issued in Insp Rept 50-316/89-02 on 891016-20,24-26 & 1204.Response to Violation & Corrective Actions Provided in Separate Submittal
ML17328A6161990-03-20020 March 1990
Responds to Generic Ltr 89-19 Re Safety Implications of Control Sys in LWRs (USI A-47).Both Units Have Steam Overfill Protection Sys That Would Automatically Prevent Water Carryover Into Steam Lines If Control Sys Failed
ML17328A6121990-03-13013 March 1990
Forwards Monthly Performance Monitoring Rept,Jan 1990.
ML17325B3901990-03-0606 March 1990
Forwards Annual Rept to NRC Per 10CFR50.54(W)(2) Re Nuclear Property Insurance
ML17325B3921990-03-0606 March 1990
Modifies Application for Amend to License DPR-74 Re Cycle 8, Per 900228 Telcon
ML17325B3781990-02-28028 February 1990
Forwards Response to NRC Info Notice 89-056 Re Questionable Certification of Matl Supplied to Dod by Nuclear Suppliers. Matl Capable of Performing Design Function & Acceptable for Continued Use
ML17328A5971990-02-27027 February 1990
Withdraws 890203 Application for Amend to License DPR-74, Modifying Tech Spec Table 3.2-1 Re DNB Parameters to Express RCS Flow Rate on Volumetric Rather than Mass Basis
ML17334B3511990-02-22022 February 1990
Submits Annual Rept of Changes to or Errors in Acceptable LOCA Evaluation Models or Application of Models for Plants. Wflash Analyses Will Be Superceded by Notrump Analyses for Unit 2,Cycle 8 Reload
ML17328A5881990-02-21021 February 1990
Responds to NRC 890914 Request for Addl Info Re Safe Shutdown Methodology.App R Fire Barriers Being Maintained & Surveilled Under 3/4.7.10 for Units 1 & 2
ML17328A5861990-02-16016 February 1990
Provides Revised Comments in Response to NRR Comments During 891213 Telcon Re Allowable Stresses for Piping & Piping Supports
1990-09-21
[Table view] |
Text
~ ~
ANEPICAN ELECTR1C PGWER Se> vice Corpo~ etio)s 2 Broadu:ay. Xeu: Pork,.l. }. Ir)00.t (2 /2p .S > 0. 9000 September 7, 1982 AEP:NRC:0739 Donald C. Cook Nuclear Plant Unit No. 1 Docket No. 50-315 License No. DPR-58 REQUEST FOR RELIEF FROM TECHNICAL SPECIFICATION 3.5.2 Mr. Harold R. Denton, Director Of fice of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Mr. Steven A. 'Varga
Dear Mr. Denton:
This letter documents the discussions held with members of your Staff concerning our request for a license amendment granting relief from the requirements of .Technical Specification 3.5.2 .
Each Unit of the Cook Nuclear Plant has two Safety Injection Pumps plus two Centrifugal Charging Pumps and two Residual Heat Removal Pumps.
In preparation for the blackout test performed on August 14, 1982, Unit No. 1's North Safety Injection Pump suffered severe damage. We
",have proceeded to repair the pump expeditiously but have not been able to complete the repairs as of the time of this letter. Unit No. 1 Technical Specification (T/S) No. 3.5.2 requires, among other things, that two Safety Injection (SI) Pumps be operable in Modes 1, 2 and 3.
Unit No. 1 is currently 'completing its refueling outage and the return to Modes 3, 2 and 1 is prevented by the inoperability of the SI pump and the T/S mentioned above. In order to proceed with the low power physics testing, we request a waiver of the requirement to have two SI pumps operable in Modes 2 (reactor thermal power less than or equal to 5%) and 3 for a period of one week beginning at the time the Unit enters Mode 3.
The South SI pump will be demonstrated operable prior to entering Mode
- 3. All remaining technical specification requirements will be met.
Attachment 1 to this letter contains the safety evaluation prepared by us in conjunction with Westinghouse.
we were to lose the operable 1S Safety The conclusion is that even if K
Injection Pump sufficient margin P
wy,
@5"-&~9'M~ 'a
wr Mr. H. R. Dent . AEP: NRC: 0739 would still exist to the limits specified in 10 CFR 50.46. Thus, this relief is not detrimental to the health and safety of the public.
I We would appreciate the expeditious handling of this request by your Staff.
AEPSC interprets 10 CFR 170.22 as requiring that a Class III Amendment Fee be paid for the change. A check in the amount of $ 4,000 will be transmitted to you in a future letter.h This Technical Specification relief request has been reviewed by the Cook Plant PNSRC. It will be reviewed by AEPSC's NSDRC at the next
~
scheduled meeting.
Due to this letter being written on short notice, it has not been prepared following our standa"d Corporate Procedures for such letters.
We shall, however, review the letter according to our Corporate Procedures and wil'nform you if any modification is required.
Very truly yours,
/emc Attachment cc: John E. Dolan - Columbus R. S. Hunter M. P. Alexich W. G. Smith, Jr. Bridgman R. C. Callen ' ~ ~
G. Charnoff Joe Williams, Jr.
NRC Resident Inspector at Cook Plant - Bridgman
Attachment to AEP:NRC:0739 LOCA Evaluation for D. C. Cook Unit 1 with One Safet In ection Pum Out of Service The purpose of this evaluation is to assess the effect of one safety injection pump out of service for the Cook Unit 1 Nuclear Plant on Loss of Coolant Accident (LOCA) consequences. Presently, the plant is fueled by Exxon Nuclear Company. However, the evaluation provided below is judged to be applicable to the non 'Westinghouse fuel, since there are no known major design differences that would have a significant impact on the LOCA behavior important for this evaluation.
Lar e Break LOCA Safety injection pump flow provides an insignificant proportion of the total ECCS flow during a large break accident, where RCS pressure rapidly drops to near atmospheric. Accumulator and low head safety injection (RHR) flow are important for this accident. Therefore, the loss of a safety injection pump has a negligible effect on large LOCA calculated peak clad temperature.
Small Break LOCA The plant's protection against small LOCAs comes from a two train system including a total of two safety injection pumps and two high head charging pumps. Small LOCA FSAR licensing analyses assume the worst single failure to be loss of a train, leaving one intermediate head SI pump and one charging pump. The small LOCA analysis yields clad temperatures well below 10 CFR 50.46 limits. This analysis assumption bounds the present plant configuration with one safety injection pump out of service and no single failure.
If the worst single fai lure assumption is considered in addition to the
- loss of the safety injection pump, and further, the train lost is assumed
'to have the operational safety injection pump, ECCS flow is delivered from only the high head charging pump. The following paragraphs evaluate this scenario.
Reduction of ECCS flow in the range of 600 to 1200 psia has an adverse
--"effect on calculated clad temperature for a range of small LOCA break sizes. The loss of a safety i njecti on pump has the effect of reducing delivered ECCS flow in that important pressure range. Total ECCS flow will be degraded by approximately 56% averaged over this pressure interval.
Established sensitivity studies have indicated that such a degradation results in as much as a 550oF small LOCA PCT increase.
~
0
~
t ~ .
The small break analysis for Cook 1 does not use the latest NRC approved W small LOCA Evaluation Model. The current small break LOCA EM would calculate a PCT of approximately 1200oF, reduced from 1493 F, predicted by the analysis in the FSAR. This new PCT is established from analysis of a substantially equivalent plant (3250 MWt,4 Loop, same SI S design) analyzed in HCAP-8970-P-A, ,"Westinghouse Emergency Core Co'oling System Small 8reak October, 1975 Model", and applies to Cook.
Additionally, credit for conservative assumptions in the small LOCA FSAR analysis can mitigate the PCT penalty. Also, the current small LOCA FSAR analysis is performed at 100% power while during the period of time for which we are seeking T/S relief the reactor will not exceed 5% power.
The lower power. level of 5% versus 100% would more than offset the consequences of reduced ECCS flow.
In conclusion, operation of Cook 1 with a safety injection pump out of service and a maximum power level of 5% for a brief 'period of time is still bounded by the small LOCA FSAR analysis. In addi tion, the fact that the present analysis has significant margin to 10 CFR 50.46 PCT limits indicates that startup of the plant from this refueling outage is not a safety concern.
