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| | issue date = 10/11/1995 | | | issue date = 10/11/1995 |
| | title = Provides Pressurized Thermal Shock Assessment for Reactor Pressure Vessel Limiting Circumferential Weld for Review & Approval | | | title = Provides Pressurized Thermal Shock Assessment for Reactor Pressure Vessel Limiting Circumferential Weld for Review & Approval |
| | author name = MERCREDY R C | | | author name = Mercredy R |
| | author affiliation = ROCHESTER GAS & ELECTRIC CORP. | | | author affiliation = ROCHESTER GAS & ELECTRIC CORP. |
| | addressee name = JOHNSON A R | | | addressee name = Johnson A |
| | addressee affiliation = NRC (Affiliation Not Assigned), NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) | | | addressee affiliation = NRC (Affiliation Not Assigned), NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| | docket = 05000244 | | | docket = 05000244 |
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| =Text= | | =Text= |
| {{#Wiki_filter:I7PRIORITY3y(ACCELERATED RZDSPROCESSING) | | {{#Wiki_filter:I PRIORITY 3y 7 |
| REGULATORY INFORMATION DISTRIBUTION SYSTEM(RIDS)ACCESSION NBR:9510180239 DOC.DATE: | | (ACCELERATED RZDS PROCESSING) |
| 95/10/11NOTARIZED: | | REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS) |
| NOFACIL:50-244 RobertEmmetGinnaNuclearPlant,Unit1,Rochester GAUTH.NAMEAUTHORAFFILIATION MERCREDY,R.C.
| | ACCESSION NBR:9510180239 DOC.DATE: 95/10/11 NOTARIZED: NO DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 P AUTH. NAME AUTHOR AFFILIATION MERCREDY,R.C. Rochester Gas & Electric Corp. |
| Rochester Gas&ElectricCorp.RECIP.NAME RECIPIENT AFFILIATION JOHNSON,A.R. | | RECIP.NAME RECIPIENT AFFILIATION JOHNSON,A.R. |
|
| |
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| ==SUBJECT:== | | ==SUBJECT:== |
| Providespressurized thermalshockassessment forreactorpressurevessellimitingcircumferential weldforreview&approval.
| | Provides pressurized thermal shock assessment for reactor pressure vessel limiting circumferential weld for review & |
| DISTRIBUTION CODE:A049DCOPIESRECEIVED:LTR ENCLSIZE:TITLE:ORSubmittal: | | approval. 0 DISTRIBUTION CODE: A049D COPIES RECEIVED:LTR ENCL SIZE: |
| ThermalShocktoReactorVesselDOCKET05000244P0RNOTES:License Expdateinaccordance with10CFR2,2.109(9/19/72).
| | TITLE: OR Submittal: Thermal Shock to Reactor Vessel R NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244 RECIPIENT COPIES RECIPIENT COPIES 1D CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-1 PD INTERNAL: AEOD/DOA NRR/DE OGC/HDS3 1'' |
| 05000244RECIPIENT 1DCODE/NAME PD1-1PDINTERNAL:
| | 1 1 0 |
| AEOD/DOANRR/DEOGC/HDS3RES/DE/MEB/MES RES/DSR/RPSB RGN1ADMSTREXTERNAL: | | JOHNSO F |
| NOAC~+t'g5'3/~iCOPIESLTTRENCL111''01011111111RECIPIENT IDCODE/NAME JOHNSOF1NRR/DSSA/SRXB RESDERES/DSIR/EIB RES/DST/PRAB NRCPDRCOPIESLTTRENCL1'1111111111NNOTE,TOALL"RZDS"RECIPIENTS:
| | NRR/DSSA/SRXB 1 |
| PLEASEHELPUSTOREDUCEWASTE!CONTACTTHEDOCUMENTCONTROLDESK,ROOMOWFN5D8(415-2083)
| | 1' 1 |
| TOELIMINATE YOURNAMEFROMDISTRIBUTION LISTSFORDOCUMENTS YOUDON'TNEEDITOTAL,NUMBEROFCOPIESREQUIRED:
| | 1 1 |
| LTTR15ENCL13 04 ANDROCHE/TER GASANDELECTRICCORPORATION
| | 1 1 0 RES DE 1 1 RES/DE/MEB/MES 1 1 RES/DSIR/EIB 1 1 RES/DSR/RPSB 1 1 RES/DST/PRAB 1 1 RGN1 ADMSTR 1 1 |
| ~89FASTAYENLIE, ROCHESTER, N.Y.Idio-000I AREACODE716Sd6-2700ROBERTC.MECREDYVicePresident NvcleorOperotions, October11,1995U.S.NuclearRegulatory Commission DocumentControlDeskAttn:AllenR.JohnsonProjectDirectorate I-1Washington, D.C.20555 | | ~+ |
| | EXTERNAL: NOAC 1 1 NRC PDR t'g 5'3/~i N |
| | NOTE,TO ALL "RZDS" RECIPIENTS: |
| | PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM OWFN 5D8 (415-2083) TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEEDI TOTAL, NUMBER OF COPIES REQUIRED: LTTR 15 ENCL 13 |
| | |
| | 0 4 AND ROCHE/TER GAS AND ELECTRIC CORPORATION ~ 89 FASTAYENLIE, ROCHESTER, N.Y. Idio-000I AREA CODE 716 Sd6-2700 ROBERT C. MECREDY Vice President N vcleor Operotions, October 11, 1995 U.S. Nuclear Regulatory Commission Document Control Desk Attn: Allen R. Johnson Project Directorate I-1 Washington, D.C. 20555 |
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| ==Subject:== | | ==Subject:== |
| Pressurized ThermalShockAssessment forGinnaReactorVesselR.E.GinnaNuclearPowerPlantDocketNo.50-244Ref.(a):LetterfromAllenR.Johnson(NRC),toRobertC.Mecredy(RG&E),"SummaryofMeetingwithRochester GasandElectricCorporation onMay16,1995-Pressurizer ThermalShockAssessment ofReactorVessels," | | Pressurized Thermal Shock Assessment for Ginna Reactor Vessel R.E. Ginna Nuclear Power Plant Docket No. 50-244 Ref. (a): Letter from Allen R. Johnson (NRC), to Robert C. Mecredy (RG&E), "Summary of Meeting with Rochester Gas and Electric Corporation on May 16, 1995 Pressurizer Thermal Shock Assessment of Reactor Vessels," dated June 5, 1995. |
| datedJune5,1995.
| | |
| | ==Dear Mr. Johnson:== |
| | |
| | The purpose of this correspondence is to provide Rochester Gas and Electric Corporation's (RG&E) pressurized thermal shock (PTS) assessment for the R.E. Ginna reactor pressure vessel (RPV) limiting circumferential weld for your review and approval. The assessment, as presented in the attached, is provided in response to the referenced letter and is based on the guidance provided in 10CFR50.61 and Regulatory Guide 1.99 revision 2. As provided for by this guidance, the RG&E assessment includes the use of the results of the RPV surveillance capsule program. As the attached PTS determination shows, the Ginna RPV can be expected to operate past the end of license (EOL) of 2009 without exceeding the PTS screening criterion of 10CFR50.61 for the limiting case weld. |
| | Very truly yours, Robert C. Mecredy REJtl,393 i.70jg J 95i0180239 'st5lOii PDR ADOCK 05000244 P PDR pyre+ p 8'K/ prtl. |
| | |
| | xc Mr'. Allen R. Johnson (Mail Stop 14B2) |
| | Project Directorate I-1 Washington, D.C. 20555 U. S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna Senior Resident Inspector |
| | |
| | ATTACHMENT R.E. Ginna Reactor Vessel EOL PTS Assessment Discussion The purpose of this assessment is to show that the expected reference temperature PTS (RT~,) at EOL for the Ginna RPV limiting weld remains below the screening criteria when using surveillance program data. The limiting material for the Ginna RPV is the intermediate to lower shell weld having weld wire heat number 61782. The material used for the surveillance capsule program is weld wire SA-1036 having material heat number 61782 which has been determined to be acceptable for use as a surveillance material+. The determination of the reference temperature for RT~ for the RPV-PTS assessment uses the 10CFR50.61(b)(3) plant specific surveillance program method which includes the surveillance data adjustments per Regulatory Guide 1.99 revision 2 position 2, where: |
| | (1) RTpgg = I + M + aRTggg (2) I = initial RT~~~ |
| | (4) ~RT~Y = reference temperature shift"> |
| | |
| | 8 II. Material Values Used A. RPV material heat 61782 1~ I= the mean of the test values where: 1 o F(() |
| | i=1 n |
| | i1 n |
| | n = |
| | = |
| | 2 |
| | -38'F) |
| | ~ I= -19.5'F and o', = 18.5'F 2 ~ fluence 9 EOL = 3.68E19(" |
| | 3 ~ Chemistry(') |
| | Cu = ~ 25 Ni = .54 4 ~ CF('): |
| | 167.6oF 5 ~ M = 2ltr~ + e,~ where e, = 18.5'F (II.A.1) and 14 o F(4) |
| | ~ M = 46.4'F B. Surveillance Capsules Weld SA 1036 1~ ~RTND V 5.56E18 140oF R 1.15E19 165oF T 1.97E19 150oF S 3.87E19 205oF 2 ~ Mean Value Chemistry+(')+: |
| | Cu = .214 Ni = .505 3 ~ CF>>: |
| | 150 'oF |
| | |
| | IXX. Ca'lculations A. Chemistry factor ratio<4>: |
| | CF of RPV (II.A.4 above) : CF of capsule (II.B.3 above) |
| | ~ Ratio = 167.6 = 1.11 150.9 B. Best fit CF<" |
| | : 1. Summing adjusted capsule ~RTNDT Adjusted V 140 X 1.11 155.4 R 165 X 1.11 183.2 T 150 X 1.11 166.5 S 205 X 1.11 227.6 2 ~ Determine sum of fluence factor x adjusted aRT~~ |
| | where fluence factor (ff) = f< " |
| | Adjusted |
| | ~Ca sule FF ~RT m Product V 0.836 155. 4 129.9. |
| | R 1.039 183.2 190.3 T 1.185 166.5 197.3 S 1.349 227.6 307. |
| | sum = 824.5 3 ~ Summing squares of fluence factors (FF) for capsules using fluence factor values from II.B.1: |
| | ~ca sule V 0. 836 0. 699 R 1. 039 1. 080 T 1.185 1. 405 S 1.349 1.820 sum = 5.004 4 Sum of adjusted aRTN T |
| | : sum of squares of fluence 824.5 (XII.B.2) : 5.004 (III.B.3) = 164.8 F |
|
| |
|
| ==DearMr.Johnson:==
| | C.'RT~~ for surveillance fit 9 EOL EOL gRT~~ (CF) frogs-oioi~o where f= 3.68 |
| Thepurposeofthiscorrespondence istoprovideRochester GasandElectricCorporation's (RG&E)pressurized thermalshock(PTS)assessment fortheR.E.Ginnareactorpressurevessel(RPV)limitingcircumferential weldforyourreviewandapproval.
| | ( I.B.4) x 3.68 8 F) (1. 338) = 220. '164. |
| Theassessment, aspresented intheattached, isprovidedinresponsetothereferenced letterandisbasedontheguidanceprovidedin10CFR50.61 andRegulatory Guide1.99revision2.Asprovidedforbythisguidance, theRG&Eassessment includestheuseoftheresultsoftheRPVsurveillance capsuleprogram.AstheattachedPTSdetermination shows,theGinnaRPVcanbeexpectedtooperatepasttheendoflicense(EOL)of2009withoutexceeding thePTSscreening criterion of10CFR50.61 forthelimitingcaseweld.Verytrulyyours,RobertC.MecredyREJtl,393 i.70jgJ95i0180239
| | 5 D. RTrrs ~ EOL RT~ = I+ M + aRTNg~ |
| 'st5lOiiPDRADOCK05000244PPDRpyre+p8'K/prtl.
| | -19 ~ 5 F (II.A.1) + 46.4 F (II.A.5) + 220.5 (III.C) 247.4oF IV. Results The RT~ value at EOL for the R.E. Ginna vessel is 247.4 F. |
| xcMr'.AllenR.Johnson(MailStop14B2)ProjectDirectorate I-1Washington, D.C.20555U.S.NuclearRegulatory Commission RegionI475Allendale RoadKingofPrussia,PA19406GinnaSeniorResidentInspector ATTACHMENT R.E.GinnaReactorVesselEOLPTSAssessment Discussion Thepurposeofthisassessment istoshowthattheexpectedreference temperature PTS(RT~,)atEOLfortheGinnaRPVlimitingweldremainsbelowthescreening criteriawhenusingsurveillance programdata.ThelimitingmaterialfortheGinnaRPVistheintermediate tolowershellweldhavingweldwireheatnumber61782.Thematerialusedforthesurveillance capsuleprogramisweldwireSA-1036havingmaterialheatnumber61782whichhasbeendetermined tobeacceptable foruseasasurveillance material+.
| | This is 52.64F below the PTS screening criterion described in 10CFR50.61. |
| Thedetermination ofthereference temperature forRT~fortheRPV-PTSassessment usesthe10CFR50.61(b)(3) plantspecificsurveillance programmethodwhichincludesthesurveillance dataadjustments perRegulatory Guide1.99revision2position2,where:(1)RTpgg=I+M+aRTggg(2)I=initialRT~~~(4)~RT~Y=reference temperature shift">
| |
| 8II.MaterialValuesUsedA.RPVmaterialheat617821~I=themeanofthetestvaluesi=1nwhere:1oF(()1i=-38'F)nn=2~I=-19.5'Fando',=18.5'F2~3~4~5~fluence9EOL=3.68E19("
| |
| Chemistry(')
| |
| Cu=~25Ni=.54CF('):167.6oFM=2ltr~+e,~wheree,=18.5'F(II.A.1)and14oF(4)~M=46.4'FB.Surveillance CapsulesWeldSA10361~~RTNDVRTS5.56E181.15E191.97E193.87E19140oF165oF150oF205oF2~3~MeanValueChemistry+(')+:
| |
| Cu=.214Ni=.505CF>>:150'oF IXX.Ca'lculations A.Chemistry factorratio<4>:
| |
| CFofRPV(II.A.4above)-:CFofcapsule(II.B.3above)~Ratio=167.6=1.11150.9B.BestfitCF<"1.Summingadjustedcapsule~RTNDTAdjustedVRTS140165150205X1.11X1.11X1.11X1.11155.4183.2166.5227.62~Determine sumoffluencefactorxadjustedaRT~~wherefluencefactor(ff)=f<"~CasuleFFAdjusted~RTmProductVRTS0.8361.0391.1851.349155.4183.2166.5227.6sum=129.9.190.3197.3307.824.53~Summingsquaresofcapsulesfluencefactors(FF)forusingfluencefactorvaluesfromII.B.1:~casuleVRTS0.8361.0391.1851.349sum=0.6991.0801.4051.8205.0044SumofadjustedaRTNT-:sumofsquaresoffluence824.5(XII.B.2)
| |
| -:5.004(III.B.3)
| |
| =164.8F C.'RT~~forsurveillance fit9EOLEOLgRT~~(CF)frogs-oioi~o wheref=3.68(I.B.4)x3.68'164.8F)(1.338)=220.5D.RTrrs~EOLRT~=I+M+aRTNg~-19~5F(II.A.1)+46.4F(II.A.5)+220.5(III.C)247.4oFIV.ResultsTheRT~valueatEOLfortheR.E.Ginnavesselis247.4F.Thisis52.64FbelowthePTSscreening criterion described in10CFR50.61.
| |
|
| |
|
| ==REFERENCES:== | | ==REFERENCES:== |
|
| |
|
| (1)BAW1803Revision1,"Correlations forPredicting theEffectsofNeutronRadiation onLinde80Submerged | | (1) BAW 1803 Revision 1, "Correlations for Predicting the Effects of Neutron Radiation on Linde 80 Submerged Arc Welds." |
| -ArcWelds."(2)WCAP-13902, "Analysis ofCapsuleSfromtheRochester GasandElectricCorporation R.e.GinnaReactorVesselRadiation Surveillance Program,"
| | (2) WCAP-13902, "Analysis of Capsule S from the Rochester Gas and Electric Corporation R.e. Ginna Reactor Vessel Radiation Surveillance Program," dated December 1993. |
| datedDecember1993.(3)CodeofFederalRegulations, 10CFR50.61, "Fracture Toughness Requirements forProtection AgainstPressurized ThermalShockEvents."(4)Regulatory Guide1.99,revision2,"Radiation Embrittlement ofReactorVesselMaterials."
| | (3) Code of Federal Regulations, 10CFR50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events. " |
| (5)BAW1920P,"Analysis ofCapsuleDB1-LG1," | | (4) Regulatory Guide 1.99, revision 2, "Radiation Embrittlement of Reactor Vessel Materials." |
| October1986.(6)LettertoRogerW.KoberfromMortonB.Fairtile, "SafetyEvaluation bytheOfficeofNuclearReactorRegulation RelatedtoAmendment No.15toFacilityOperating LicenseNo.DPR-18Rochester GasandElectricCorporation R.E.GinnaNuclearPowerPlantDocketNo.50-244,"datedJune12(1986.(7)BAW-1500, "Chemistry of177-FAB&WOwner'sGroupReactorVesselBeltlineWelds,"September 1978.(8)10CFR50.61, "Fracture toughness requirements forprotection againstpressurized thermalshockevents."(9)BAW-2121P, "Chemical Composition ofB&WFabricated ReactorVesselBeltlineWelds,"April1991.}}
| | (5) BAW 1920P, "Analysis of Capsule DB1-LG1," October 1986. |
| | (6) Letter to Roger W. Kober from Morton B. Fairtile, "Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 15 to Facility Operating License No. DPR-18 Rochester Gas and Electric Corporation R.E. |
| | Ginna Nuclear Power Plant Docket No. 50-244," dated June 12( 1986. |
| | (7) BAW-1500, "Chemistry of 177-FA B& W Owner's Group Reactor Vessel Beltline Welds," September 1978. |
| | (8) 10CFR50.61, "Fracture toughness requirements for protection against pressurized thermal shock events." |
| | (9) BAW-2121P, "Chemical Composition of B&W Fabricated Reactor Vessel Beltline Welds," April 1991.}} |
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Supplements 980806 Relief Request with Attached Table.Util Third 10-Yr ISI of Reactor Vessel Being Performed During 1999 Refueling Outage,Beginning on 990301
ML17265A5361999-02-0202 February 1999
Forwards Response to NRC 981203 RAI Re Resolution of Unresolved Safety Issue USI A-46.Util Does Not Agree with NRCs Interpretation.Detailed Bases,Encl
ML17311A0691999-01-25025 January 1999
Forwards Revs to Ginna Station Emergency Plan Implementing Procedures (Epips).Previous Rev Had Incorrect Effective Date
ML17311A0671999-01-14014 January 1999
Forwards Revised Emergency Operating Procedures for Re Ginna NPP
ML17265A5131999-01-12012 January 1999
Forwards Revised Cover Page for Ginna Station Technical Requirements Manual (Trm), Rev 7,correcting Error in Effective Date
1999-09-30
[Table view]
Category:INCOMING CORRESPONDENCE
MONTHYEARML17250B3041999-10-20020 October 1999
Forwards Changes to Tech Specs Bases for Re Ginna Nuclear Power Plant.Change Bars Indicate Those Revs Which Have Been Incorporated
ML17265A7651999-10-0808 October 1999
Forwards Fifteen Relief Requests That Will Be Utilized for Ginna NPP Fourth Interval ISI Program That Will Start on Jan 1,2000.Attachment 1 Includes Summaries & Detailed Description of Each Relief Request
ML17265A7631999-10-0505 October 1999
Requests Approval for Use of Relief Request Number 43 to Address Volumetric Examination Limitations (Less than 90%) Associated with a & B RHR Heat Exchanger Outlet Nozzle to Shell Welds.Approval Is Requested by Dec 31,2000
ML17265A7641999-10-0505 October 1999
Requests Approval for Use of Relief Request Number 42 to Address Volumetric Examinations Limitations (Less than 90%) Associated with Eight Class 1 Identified Welds or Areas of Reactor Pressure Vessel
ML20212J3561999-09-30030 September 1999
Forwards Four Copies of Re Ginna NPP Training & Qualification Plan for Security Officers, Rev 7,dtd 990930. Synopsis of Changes,Encl.Encl Withheld Per 10CFR73.21
ML20212J3801999-09-30030 September 1999
Forwards Four Copies of Rev to Re Ginna NPP Security Plan.Rev Changes Contingency Weapons Available to Response Force to Those Most Effective in Current Defensive Strategy.Encl Withheld Per 10CFR73.21
IR 05000244/19990051999-09-24024 September 1999
Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs
ML17265A7571999-09-24024 September 1999
Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs
ML17265A7461999-08-31031 August 1999
Submits Response to NRC Administrative Ltr 95-03,rev 2, Availability of Reactor Vessel Integrity Database,Version, Dtd 990726
ML17265A7401999-08-26026 August 1999
Requests Approval for Use of Relief Request Number 35 Re Use of ASME Section XI Code,1995 Edition,1996 Addenda.Code Will Be Used to Develop Plant Fourth 10-year Interval ISI Program on Class 1,2 & 3 Components
ML17265A7411999-08-26026 August 1999
Forwards LER 99-004-01 Re Plant Being Outside Design Basis Due to Containment Recirculation Fan Moisture Separator Vanes Being Incorrectly Installed.Part 21 Notification of 990512 Is Being Rescinded
ML17250B3021999-08-23023 August 1999
Informs That Util & NRC Had Conference Call on 990816 to Review Approach in Responding to Questions,As Result of Questions Re Main Steam Check Valve Performance Included in Insp Rept 50-244/99-05,dtd 990806
ML17265A7451999-08-23023 August 1999
Forwards fitness-for-duty Performance Data Rept for Six Months Ending 990630,per 10CFR26.71(d)
ML17265A7271999-07-30030 July 1999
Forwards 10CFR21 Interim Rept Per Reporting of Defects & Noncompliance,Section 21 (a) (2).Interim Rept Prepared Because Evaluation Cannot Be Completed within 60 Days from Discovery of Deviation or Failure to Comply
ML17265A7151999-07-23023 July 1999
Forwards LER 99-007-01 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Reactor Trip.Further Investigation of Event Identified Addl Corrective Actions
ML17265A7061999-07-22022 July 1999
Forwards LER 98-003-02,re Actuations of CR Emergency Air Treatment Sys.All Creats Actuations,Including Those Originally Believed to Be Valid Actuations,Were,In Fact Invalid Actuations
ML17265A7191999-07-21021 July 1999
Forwards Ginna Station ISI Rept for Refueling Outage Conducted in 1999
ML17265A7141999-07-21021 July 1999
Withdraws Relief Request 35 for Plant Inservice Insp Program Section XI Requirements,Submitted on 980806.Licensee Plans to Resubmit Relief Request,Which Includes Addl Level of Detail,In Near Future
ML17265A7041999-07-16016 July 1999
Submits Info Re Specific Licensing Actions Which May Be Expected to Generate Complex Reviews,In Response to Administrative Ltr 99-02,dtd 990603
ML17265A6911999-06-30030 June 1999
Responds to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Gl 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Readiness Disclosure Attached
ML17265A6871999-06-22022 June 1999
Forwards Response to RAI Made During 990225 Telcon Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves. Calculation Encl. Encl
ML17265A6841999-06-21021 June 1999
Informs That Util Wishes to Amend Extend of Alternate Exams Provided for Relief Request Re ISI Program ASME Section XI Require Exams for First 10-Yr Interval for Containment
ML17265A6741999-06-15015 June 1999
Submits Annual ECCS Rept IAW 10CFR50.46(a)(3)(ii) Requirements.No Changes Have Been Made to Large Break LOCA PCT & Small Break LOCA PCT
ML17265A6751999-06-11011 June 1999
Responds to NRC RAI Re Licensee GL 96-05 Program.Encl Info Verifies That Util Is Implementing Provisions of JOG Program on MOV Periodic Verification
ML17309A6551999-06-0707 June 1999
Responds to NRC 990310 RAI Re Verification of Seismic Adequacy of Mechanical & Electrical Equipment
ML17265A6671999-06-0101 June 1999
Requests Approval of Ginna QA Program for Radioactive Matl Packages,Form 311,approval Number 0019.Ginna QA Program for Station Operation, Was Most Recently Submitted to NRC by Ltr Dtd 981221 & Supplemented on 990301
ML17265A6641999-05-25025 May 1999
Forwards Addl Info on Use of GIP Method a for Re Ginna Nuclear Power Plant.Copy of Re Ginna Station USI A-46 Outlier Resolution Table as Requested
ML17265A6611999-05-24024 May 1999
Forwards LER 99-007-00 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Rt.Util Is Planning to Submit Suppl to LER by 990730
ML20206U0921999-05-13013 May 1999
Forwards Four Copies of Rev R to Gnpp Security Plan,Per Provisions of 10CFR50.54(p).Rev Clarifies Armed Response Team Assignments & Does Not Degrade Physical Security Effectiveness.Rev Withheld,Per 10CFR73.21
ML17265A6461999-05-12012 May 1999
Forwards 1998 Annual Radioactive Effluent Release Rept & 1998 Annual Radiological Environ Operating Rept, for Re Ginna NPP
ML17265A6411999-05-12012 May 1999
Forwards LER 99-004-00 IAW 10CFR50.73 & 10CFR21.Further Assessment Will Be Provided in Suppl to LER by 990630
ML20206E7221999-04-29029 April 1999
Forwards Four Copies of Rev Q to Re Ginna Nuclear Power Plant Security Plan,Per 10CFR50.54(p).Changes Do Not Degrade Physical Security Effectiveness.Encl Withheld,Per 10CFR73.21
ML20206H6911999-04-22022 April 1999
Forwards Info Requested During Informal Telcon on 980408 Concerning Upcoming Osre at Ginna Station.Info Requested Listed.Without Encls
ML17265A6271999-04-19019 April 1999
Forwards Rev 0 & Rev 1 to Colr,Cycle 28 for Re Ginna NPP, Per TS 5.6.5
ML17309A6501999-04-14014 April 1999
Forwards Revised Ginna Station EOPs & Procedures Index
ML17265A6121999-03-29029 March 1999
Forwards Rept on Status of Decommissioning Funding for Re Ginna Npp,For Which Rg&E Is Sole Owner,Per 10CFR50.75. Data Presented Herein,Current as of 981231
ML17265A6041999-03-24024 March 1999
Forwards LER 99-001-00,IAW 10CFR50.73(a)(2)(ii)(B) & 10CFR21.Addl Analyses Are Being Performed to Support Future Cycle Operation & Supplemental LER Is Scheduled to Be Submitted by 990618
ML17265A5641999-03-0101 March 1999
Forwards Response to NRC 990217 RAI Concerning Changes to QA Program for Re Ginna Station Operation.Rg&E Is Modifying Changes Requested in 981221 Submittal.Modified QA Program,Encl
ML17265A5671999-03-0101 March 1999
Forwards Application for Amend to License DPR-18,to Revise TSs Battery Cell Parameters Limit for Specific Gravity (SR 3.8.6.3 & SR 3.8.6.6).Supporting Tss,Encl
ML17265A5551999-02-25025 February 1999
Informs That Util Is in Process of Revising fitness-for-duty Program,Developed in Accordance with 10CFR26.Util Will Continue to Use Dept of Health & Human Svcs Certified Test Facility for Majority of Tests During Yr
ML17265A5561999-02-22022 February 1999
Forwards FFD Performance Data Rept for Six Months Ending 981231,per 10CFR26.71(d)
ML17265A5451999-02-12012 February 1999
Forwards Simulator Four Year Certification Rept,Per 10CFR55.45(b)(5)(ii)
ML17309A6491999-02-12012 February 1999
Forwards Ginna Station EOPs
ML17265A5431999-02-0909 February 1999
Supplements 980806 Relief Request with Attached Table.Util Third 10-Yr ISI of Reactor Vessel Being Performed During 1999 Refueling Outage,Beginning on 990301
ML17265A5361999-02-0202 February 1999
Forwards Response to NRC 981203 RAI Re Resolution of Unresolved Safety Issue USI A-46.Util Does Not Agree with NRCs Interpretation.Detailed Bases,Encl
ML17311A0691999-01-25025 January 1999
Forwards Revs to Ginna Station Emergency Plan Implementing Procedures (Epips).Previous Rev Had Incorrect Effective Date
ML17311A0671999-01-14014 January 1999
Forwards Revised Emergency Operating Procedures for Re Ginna NPP
ML17265A5131999-01-12012 January 1999
Forwards Revised Cover Page for Ginna Station Technical Requirements Manual (Trm), Rev 7,correcting Error in Effective Date
ML17265A5111999-01-11011 January 1999
Requests Relief Per 10CFR50.55a(a)(3)(ii) from Certain Requirements of Section XI of ASME Bp&V Code for ISI Program.Relief Requests 37,38 & 39 Encl
ML17265A5101999-01-11011 January 1999
Requests Relief Per to 10CFR50.55a(a)(3)(ii) from Certain Requirements of Section XI of ASME B&PV Code for ISI Program.Relief Request 40 Encl
1999-09-30
[Table view] |
Text
I PRIORITY 3y 7
(ACCELERATED RZDS PROCESSING)
REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:9510180239 DOC.DATE: 95/10/11 NOTARIZED: NO DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 P AUTH. NAME AUTHOR AFFILIATION MERCREDY,R.C. Rochester Gas & Electric Corp.
RECIP.NAME RECIPIENT AFFILIATION JOHNSON,A.R.
SUBJECT:
Provides pressurized thermal shock assessment for reactor pressure vessel limiting circumferential weld for review &
approval. 0 DISTRIBUTION CODE: A049D COPIES RECEIVED:LTR ENCL SIZE:
TITLE: OR Submittal: Thermal Shock to Reactor Vessel R NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244 RECIPIENT COPIES RECIPIENT COPIES 1D CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-1 PD INTERNAL: AEOD/DOA NRR/DE OGC/HDS3 1
1 1 0
JOHNSO F
NRR/DSSA/SRXB 1
1' 1
1 1
1 1 0 RES DE 1 1 RES/DE/MEB/MES 1 1 RES/DSIR/EIB 1 1 RES/DSR/RPSB 1 1 RES/DST/PRAB 1 1 RGN1 ADMSTR 1 1
~+
EXTERNAL: NOAC 1 1 NRC PDR t'g 5'3/~i N
NOTE,TO ALL "RZDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM OWFN 5D8 (415-2083) TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEEDI TOTAL, NUMBER OF COPIES REQUIRED: LTTR 15 ENCL 13
0 4 AND ROCHE/TER GAS AND ELECTRIC CORPORATION ~ 89 FASTAYENLIE, ROCHESTER, N.Y. Idio-000I AREA CODE 716 Sd6-2700 ROBERT C. MECREDY Vice President N vcleor Operotions, October 11, 1995 U.S. Nuclear Regulatory Commission Document Control Desk Attn: Allen R. Johnson Project Directorate I-1 Washington, D.C. 20555
Subject:
Pressurized Thermal Shock Assessment for Ginna Reactor Vessel R.E. Ginna Nuclear Power Plant Docket No. 50-244 Ref. (a): Letter from Allen R. Johnson (NRC), to Robert C. Mecredy (RG&E), "Summary of Meeting with Rochester Gas and Electric Corporation on May 16, 1995 Pressurizer Thermal Shock Assessment of Reactor Vessels," dated June 5, 1995.
Dear Mr. Johnson:
The purpose of this correspondence is to provide Rochester Gas and Electric Corporation's (RG&E) pressurized thermal shock (PTS) assessment for the R.E. Ginna reactor pressure vessel (RPV) limiting circumferential weld for your review and approval. The assessment, as presented in the attached, is provided in response to the referenced letter and is based on the guidance provided in 10CFR50.61 and Regulatory Guide 1.99 revision 2. As provided for by this guidance, the RG&E assessment includes the use of the results of the RPV surveillance capsule program. As the attached PTS determination shows, the Ginna RPV can be expected to operate past the end of license (EOL) of 2009 without exceeding the PTS screening criterion of 10CFR50.61 for the limiting case weld.
Very truly yours, Robert C. Mecredy REJtl,393 i.70jg J 95i0180239 'st5lOii PDR ADOCK 05000244 P PDR pyre+ p 8'K/ prtl.
xc Mr'. Allen R. Johnson (Mail Stop 14B2)
Project Directorate I-1 Washington, D.C. 20555 U. S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna Senior Resident Inspector
ATTACHMENT R.E. Ginna Reactor Vessel EOL PTS Assessment Discussion The purpose of this assessment is to show that the expected reference temperature PTS (RT~,) at EOL for the Ginna RPV limiting weld remains below the screening criteria when using surveillance program data. The limiting material for the Ginna RPV is the intermediate to lower shell weld having weld wire heat number 61782. The material used for the surveillance capsule program is weld wire SA-1036 having material heat number 61782 which has been determined to be acceptable for use as a surveillance material+. The determination of the reference temperature for RT~ for the RPV-PTS assessment uses the 10CFR50.61(b)(3) plant specific surveillance program method which includes the surveillance data adjustments per Regulatory Guide 1.99 revision 2 position 2, where:
(1) RTpgg = I + M + aRTggg (2) I = initial RT~~~
(4) ~RT~Y = reference temperature shift">
8 II. Material Values Used A. RPV material heat 61782 1~ I= the mean of the test values where: 1 o F(()
i=1 n
i1 n
n =
=
2
-38'F)
~ I= -19.5'F and o', = 18.5'F 2 ~ fluence 9 EOL = 3.68E19("
3 ~ Chemistry(')
Cu = ~ 25 Ni = .54 4 ~ CF('):
167.6oF 5 ~ M = 2ltr~ + e,~ where e, = 18.5'F (II.A.1) and 14 o F(4)
~ M = 46.4'F B. Surveillance Capsules Weld SA 1036 1~ ~RTND V 5.56E18 140oF R 1.15E19 165oF T 1.97E19 150oF S 3.87E19 205oF 2 ~ Mean Value Chemistry+(')+:
Cu = .214 Ni = .505 3 ~ CF>>:
150 'oF
IXX. Ca'lculations A. Chemistry factor ratio<4>:
CF of RPV (II.A.4 above) : CF of capsule (II.B.3 above)
~ Ratio = 167.6 = 1.11 150.9 B. Best fit CF<"
- 1. Summing adjusted capsule ~RTNDT Adjusted V 140 X 1.11 155.4 R 165 X 1.11 183.2 T 150 X 1.11 166.5 S 205 X 1.11 227.6 2 ~ Determine sum of fluence factor x adjusted aRT~~
where fluence factor (ff) = f< "
Adjusted
~Ca sule FF ~RT m Product V 0.836 155. 4 129.9.
R 1.039 183.2 190.3 T 1.185 166.5 197.3 S 1.349 227.6 307.
sum = 824.5 3 ~ Summing squares of fluence factors (FF) for capsules using fluence factor values from II.B.1:
~ca sule V 0. 836 0. 699 R 1. 039 1. 080 T 1.185 1. 405 S 1.349 1.820 sum = 5.004 4 Sum of adjusted aRTN T
- sum of squares of fluence 824.5 (XII.B.2) : 5.004 (III.B.3) = 164.8 F
C.'RT~~ for surveillance fit 9 EOL EOL gRT~~ (CF) frogs-oioi~o where f= 3.68
( I.B.4) x 3.68 8 F) (1. 338) = 220. '164.
5 D. RTrrs ~ EOL RT~ = I+ M + aRTNg~
-19 ~ 5 F (II.A.1) + 46.4 F (II.A.5) + 220.5 (III.C) 247.4oF IV. Results The RT~ value at EOL for the R.E. Ginna vessel is 247.4 F.
This is 52.64F below the PTS screening criterion described in 10CFR50.61.
REFERENCES:
(1) BAW 1803 Revision 1, "Correlations for Predicting the Effects of Neutron Radiation on Linde 80 Submerged Arc Welds."
(2) WCAP-13902, "Analysis of Capsule S from the Rochester Gas and Electric Corporation R.e. Ginna Reactor Vessel Radiation Surveillance Program," dated December 1993.
(3) Code of Federal Regulations, 10CFR50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events. "
(4) Regulatory Guide 1.99, revision 2, "Radiation Embrittlement of Reactor Vessel Materials."
(5) BAW 1920P, "Analysis of Capsule DB1-LG1," October 1986.
(6) Letter to Roger W. Kober from Morton B. Fairtile, "Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 15 to Facility Operating License No. DPR-18 Rochester Gas and Electric Corporation R.E.
Ginna Nuclear Power Plant Docket No. 50-244," dated June 12( 1986.
(7) BAW-1500, "Chemistry of 177-FA B& W Owner's Group Reactor Vessel Beltline Welds," September 1978.
(8) 10CFR50.61, "Fracture toughness requirements for protection against pressurized thermal shock events."
(9) BAW-2121P, "Chemical Composition of B&W Fabricated Reactor Vessel Beltline Welds," April 1991.
