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| {{#Wiki_filter:PRIORIT'Y Z REGULAT(<NAORMNEetH VTSVREMTION4STEM (RZDS)ACCESSION NBR:9505150014 DOC.DATE: 95/05/05 NOTARIZED: | | {{#Wiki_filter:PRIORIT'Y Z REGULAT(<NAORMNEetH VTSVREMTION4STEM (RZDS) |
| NO FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1 Rochester G AUTH i,NAME AUTHOR AFFILIATION MECREDY,R.C. | | ACCESSION NBR:9505150014 DOC.DATE: 95/05/05 NOTARIZED: NO DOCKET I FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1 Rochester G 05000244 AUTH i,NAME AUTHOR AFFILIATION MECREDY,R.C. Rochester Gas & Electric Corp. |
| Rochester Gas&Electric Corp.RECIP.NAME RECIPIENT AFFILIATION JOHNSON,A.R. | | RECIP.NAME RECIPIENT AFFILIATION JOHNSON,A.R. Document Control Branch (Doc ent Control Desk) |
| Document Control Branch (Doc ent Control Desk)JOHNSON,A.R. | | JOHNSON,A.R. Project Directorate I-1 (PDl-1) (Post 941001) |
| Project Directorate I-1 (PDl-1)(Post 941001) | | P |
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| ==SUBJECT:== | | ==SUBJECT:== |
| Forwards listed doucments for NRC review 6 approval prior submittal of improved TS consistent w/NUREG-1431. | | Forwards listed doucments for NRC review 6 approval prior to submittal of improved TS consistent w/NUREG-1431. |
| DISTRIBUTION CODE: A001D COPIES RECEIVED:LTR ENCL SIZE: TITLE: OR Submittal: | | DISTRIBUTION CODE: A001D COPIES RECEIVED:LTR ENCL SIZE: |
| General Distribution DOCKET I 05000244 to P NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). | | TITLE: OR Submittal: General Distribution NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244 P RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL 1D CODE/NAME LTTR ENCL PDl-1 LA 1 1 PD1-1 PD 1 1 JOHNSON,A 1 1 INTERNA E CE 1 1 NRR/DE/EMCB 1 1 DRCH/HICB 1 1 NRR/DSSA/SPLB 1 1 Y |
| 05000244 P INTERNA RECIPIENT ID CODE/NAME PDl-1 LA JOHNSON,A E CE DRCH/HICB NRR/DSSA/SRXB OGC/HDS3 COPIES LTTR ENCL 1 1 1 1 1 1 1 1 1 1 1 0 RECIPIENT 1D CODE/NAME PD1-1 PD NRR/DE/EMCB NRR/DSSA/SPLB NUDOCS-ABSTRACT COPIES LTTR ENCL 1 1 1 1 1 1 1 1 Y EXTERNAL: NOAC 1 1 NRC PDR 1 1 D U N NOTE TO ALL"RIDS" RECIPIENTS:. | | NRR/DSSA/SRXB 1 1 NUDOCS-ABSTRACT 1 1 OGC/HDS3 1 0 EXTERNAL: NOAC 1 1 NRC PDR 1 1 D |
| PLEASE HELP US TO REDUCE WASTE!CONTACTTHE DOCL'MENTCONTROL DESK, ROOII I PI-37 (EXT.504-20S3)TO ELIMIYATE YOUR NAME FROiI DISTRIBUTIOY. | | U N |
| LISTS FOR DOCUKIEYTS YOU DON"I'EED!TOTAL NUMBER OF COPIES REQUIRED: LTTR 12 ENCL 11 4ND ROCHESTER GAS AND ELECTRIC CORPORATION | | NOTE TO ALL"RIDS" RECIPIENTS:. |
| ~89 EASTAYENUE, ROCHESTER, N.Y Id6d9-0001 AREA CODE 716 5'-2700 ROBERT C.MECREDY Vice President kucteor Operations May 5, 1995 U.S.Nuclear Regulatory Commission Document Control Desk Attn: Allen R.Johnson Project Directorate I-1 Washington, D.C.20555 | | PLEASE HELP US TO REDUCE WASTE! CONTACTTHE DOCL'MENTCONTROL DESK, ROOII I PI-37 (EXT. 504-20S3 ) TO ELIMIYATEYOUR NAME FROiI DISTRIBUTIOY. LISTS FOR DOCUKIEYTS YOU DON "I'EED! |
| | TOTAL NUMBER OF COPIES REQUIRED: LTTR 12 ENCL 11 |
| | |
| | 4ND ROCHESTER GAS AND ELECTRIC CORPORATION ~ 89 EASTAYENUE, ROCHESTER, N. Y Id6d9-0001 AREA CODE 716 5'-2700 ROBERT C. MECREDY Vice President kucteor Operations May 5, 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:== |
| Technical Specification Improvement Program Rochester Gas&Electric Corporation R.E.Ginna Nuclear Power Plant Docket No.50-244 | | Technical Specification Improvement Program Rochester Gas & Electric Corporation R.E. Ginna Nuclear Power Plant Docket No. 50-244 |
| | |
| | ==Dear Mr. Johnson,== |
| | |
| | Rochester Gas and Electric (RG&E) currently anticipates submitting the proposed conversion to Improved Technical Specifications (ITS) consistent with NUREG-1431 later this month. Included within this submittal are several changes to the existing technical specifications (TS) that are supported by engineering calculations or other documents that will require NRC Staff review and approval separate from the conversion review. Based on recent conversations, the NRC has requested that RG&E provide these documents prior to the ITS submittal to allow the NRC Staff additional time for review. Therefore, attached are the following documents: |
| | (a) "Criticality Analysis of the R.E. Ginna Nuclear Power Plant Fresh and Spent Fuel Racks, and Consolidated Rod Storage Canisters," dated June 1994. |
| | (b) WCAP-14040, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Revision 1, December 1994. |
| | (c) RG&E Methodology for Determining the Low Temperature Overpressure Protection System (LTOPS) Setpoints. |
| | The spent fuel pool criticality study is required to support an increase in allowed fuel enrichment necessary to support a conversion ,to 18 month fuel cycles which is planned to begin following the 1996 refueling outage. An increased in fuel enrichment affects current Ginna Station TS 5.3.1.b and TS 5.4. |
| | ~ GgQQ oi 95OS<S00<4 9S0505 PDR ADGCK 05000244 P PDR,'; |
| | |
| | The remaining two documents support the incorporation of a Pressure Temperature Limits Report (PTLR) which is planned to be implemented during the conversion to ITS. Since Ginna Station utilizes the LTOPS for preventing overpressurization of the residual heat removal system, Section 3.0 of WCAP-14040 does not apply to the installed system. Therefore, RG&E specific methodology is provided for your review. Both the RG&E specific methodology and a red-line markup of the methodology provided in WCAP-14040 are provided. |
| | All items proposed to be relocated from the current Ginna Station TS to the PTLR are addressed by these two documents. |
| | The items RG&E anticipates to be relocated to the Core Operating Limits Report (COLR) are provided in attached Table 1. As can be seen from this table, NRC approved methodology exists for all items proposed to be relocated such that no documents require submittal to the NRC at this time. |
| | The current schedule for implementation of the ITS for Ginna Station indicates NRC Staff approval of the proposed new TS by November 1995. Therefore, RG&E requests that NRC review of these three documents be coordinated to support this schedule. |
| | Ver truly yours, Robert C. Mec edy MDFK677 Attachments xc: U.S. Nuclear Regulatory Commission Mr. Allen R. Johnson (Mail Stop 14B2) |
| | PWR 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 |
| | |
| | Table 1 R.E. Ginna Proposed COLR Parameters Ginna ITS COLR Parameter NRC Approved Methodology 3.1.1 Shutdown Margin Limits WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985 3.1.3 Moderator Temperature WCAP-9272-P-A, Westinghouse Coefficient BOL limits Reload Safety Evaluation Methodology, July 1985 3.1.3 Moderator Temperature WCAP-9272-P-A, Westinghouse Coefficient EOL Limits Reload Safety Evaluation Methodology, July 1985 3.1.5 Shutdown Bank Insertion WCAP-9272-P-A, Westinghouse Limit Reload Safety Evaluation Methodology, July 1985 3.1.6 Control Bank Insertion, WCAP-9272-P-A, Westinghouse Sequence, and Overlap Reload Safety Evaluation Limits Methodology, July 1985 3.2.1 Heat flux hot. channel WCAP-9272-P-A, Westinghouse factor (F<(Z) limits), K(Z) Reload Safety Evaluation curve, and equation Methodology, July 1985 WCAP-9220-P-A, Westinghouse ECCS Evaluation Model 1981 Version, Rev. 1, February 1982 3.2.2 Nuclear Enthalpy Rise Hot WCAP-9272-P-A, Westinghouse Channel Factor (F~ limit), Reload Safety Evaluation power factor multiplier and Methodology, July 1985 equation 3.2.3 AFD CAOC limits and target, WCAP-9272-P-A, Westinghouse band Reload Safety Evaluation Methodology, July 1985 WCAP-8385, Power Distribution Control and Load Following Procedures Topical Report, September 1974 |
| | |
| | COLR Parameter NRC Approved Methodology 3.3.1 OTaT WCAP-8745-P-A, Design Bases OPaT for the Thermal Overpower ~T and Thermal Overtemperature |
| | ~T Trip Functions, September 1986 3.4.1 DNB pressurizer pressure WCAP-8567-P-A, Improved limit, RCS Tavg limit, and Thermal Design Procedure, RCS total flow limit February 1989 WCAP-11397-P-A, Revised Thermal Design Procedure, April 1989 3.5.1 Accumulator boron WCAP-9272-P-A, Westinghouse concentration Reload Safety Evaluation Methodology, July 1985 3.5.4 RWST boron concentration WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985 3.7. 12 Spent Fuel Pool Boron WCAP-11596-P-A, PHOENIX-P/ANC Concentration Nuclear Design System for Pressurized Water Reactor Cores, June 1988 3.9.1 MODE 6/Refueling Boron WCAP-9272-P-A, Westinghouse Concentration Reload Safety Evaluation Methodology, July 1985 |
|
| |
|
| ==Dear Mr.Johnson,==
| | Attachment A Criticality Analysis of the R.E. Ginna Nuclear Power Plant Fresh and Spent Fuel Racks, and Consolidated Rod Storage Canisters |
| Rochester Gas and Electric (RG&E)currently anticipates submitting the proposed conversion to Improved Technical Specifications (ITS)consistent with NUREG-1431 later this month.Included within this submittal are several changes to the existing technical specifications (TS)that are supported by engineering calculations or other documents that will require NRC Staff review and approval separate from the conversion review.Based on recent conversations, the NRC has requested that RG&E provide these documents prior to the ITS submittal to allow the NRC Staff additional time for review.Therefore, attached are the following documents: (a)"Criticality Analysis of the R.E.Ginna Nuclear Power Plant Fresh and Spent Fuel Racks, and Consolidated Rod Storage Canisters," dated June 1994.(b)WCAP-14040,"Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Revision 1, December 1994.(c)RG&E Methodology for Determining the Low Temperature Overpressure Protection System (LTOPS)Setpoints.
| |
| The spent fuel pool criticality study is required to support an increase in allowed fuel enrichment necessary to support a conversion ,to 18 month fuel cycles which is planned to begin following the 1996 refueling outage.An increased in fuel enrichment affects current Ginna Station TS 5.3.1.b and TS 5.4.~GgQQ 95OS<S00<4 9S0505 PDR ADGCK 05000244 P PDR,';oi The remaining two documents support the incorporation of a Pressure Temperature Limits Report (PTLR)which is planned to be implemented during the conversion to ITS.Since Ginna Station utilizes the LTOPS for preventing overpressurization of the residual heat removal system, Section 3.0 of WCAP-14040 does not apply to the installed system.Therefore, RG&E specific methodology is provided for your review.Both the RG&E specific methodology and a red-line markup of the methodology provided in WCAP-14040 are provided.All items proposed to be relocated from the current Ginna Station TS to the PTLR are addressed by these two documents.
| |
| The items RG&E anticipates to be relocated to the Core Operating Limits Report (COLR)are provided in attached Table 1.As can be seen from this table, NRC approved methodology exists for all items proposed to be relocated such that no documents require submittal to the NRC at this time.The current schedule for implementation of the ITS for Ginna Station indicates NRC Staff approval of the proposed new TS by November 1995.Therefore, RG&E requests that NRC review of these three documents be coordinated to support this schedule.Ver truly yours, MDFK677 Attachments Robert C.Mec edy xc: U.S.Nuclear Regulatory Commission Mr.Allen R.Johnson (Mail Stop 14B2)PWR 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 Table 1 R.E.Ginna Proposed COLR Parameters Ginna ITS COLR Parameter NRC Approved Methodology 3.1.1 Shutdown Margin Limits WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985 3.1.3 3.1.3 3.1.5 3.1.6 3.2.1 3.2.2 3.2.3 Moderator Temperature Coefficient
| |
| -BOL limits Moderator Temperature Coefficient
| |
| -EOL Limits Shutdown Bank Insertion Limit Control Bank Insertion, Sequence, and Overlap Limits Heat flux hot.channel factor (F<(Z)limits), K(Z)curve, and equation Nuclear Enthalpy Rise Hot Channel Factor (F~limit), power factor multiplier and equation AFD CAOC limits and target, band WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985 WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985 WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985 WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985 WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985 WCAP-9220-P-A, Westinghouse ECCS Evaluation Model-1981 Version, Rev.1, February 1982 WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985 WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985 WCAP-8385, Power Distribution Control and Load Following Procedures
| |
| -Topical Report, September 1974
| |
|
| |
|
| 3.3.1 OTaT OPaT COLR Parameter NRC Approved Methodology WCAP-8745-P-A, Design Bases for the Thermal Overpower~T and Thermal Overtemperature
| | / |
| ~T Trip Functions, September 1986 3.4.1 3.5.1 DNB pressurizer pressure limit, RCS Tavg limit, and RCS total flow limit Accumulator boron concentration WCAP-8567-P-A, Improved Thermal Design Procedure, February 1989 WCAP-11397-P-A, Revised Thermal Design Procedure, April 1989 WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985 3.5.4 RWST boron concentration WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985 3.7.12 3.9.1 Spent Fuel Pool Boron Concentration MODE 6/Refueling Boron Concentration WCAP-11596-P-A, PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores, June 1988 WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985 Attachment A Criticality Analysis of the R.E.Ginna Nuclear Power Plant Fresh and Spent Fuel Racks, and Consolidated Rod Storage Canisters 0/}}
| | 0}} |
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Category:CORRESPONDENCE-LETTERS
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ML17265A5361999-02-0202 February 1999
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1999-09-30
[Table view]
Category:INCOMING CORRESPONDENCE
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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
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Text
PRIORIT'Y Z REGULAT(<NAORMNEetH VTSVREMTION4STEM (RZDS)
ACCESSION NBR:9505150014 DOC.DATE: 95/05/05 NOTARIZED: NO DOCKET I FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1 Rochester G 05000244 AUTH i,NAME AUTHOR AFFILIATION MECREDY,R.C. Rochester Gas & Electric Corp.
RECIP.NAME RECIPIENT AFFILIATION JOHNSON,A.R. Document Control Branch (Doc ent Control Desk)
JOHNSON,A.R. Project Directorate I-1 (PDl-1) (Post 941001)
P
SUBJECT:
Forwards listed doucments for NRC review 6 approval prior to submittal of improved TS consistent w/NUREG-1431.
DISTRIBUTION CODE: A001D COPIES RECEIVED:LTR ENCL SIZE:
TITLE: OR Submittal: General Distribution NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244 P RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL 1D CODE/NAME LTTR ENCL PDl-1 LA 1 1 PD1-1 PD 1 1 JOHNSON,A 1 1 INTERNA E CE 1 1 NRR/DE/EMCB 1 1 DRCH/HICB 1 1 NRR/DSSA/SPLB 1 1 Y
NRR/DSSA/SRXB 1 1 NUDOCS-ABSTRACT 1 1 OGC/HDS3 1 0 EXTERNAL: NOAC 1 1 NRC PDR 1 1 D
U N
NOTE TO ALL"RIDS" RECIPIENTS:.
PLEASE HELP US TO REDUCE WASTE! CONTACTTHE DOCL'MENTCONTROL DESK, ROOII I PI-37 (EXT. 504-20S3 ) TO ELIMIYATEYOUR NAME FROiI DISTRIBUTIOY. LISTS FOR DOCUKIEYTS YOU DON "I'EED!
TOTAL NUMBER OF COPIES REQUIRED: LTTR 12 ENCL 11
4ND ROCHESTER GAS AND ELECTRIC CORPORATION ~ 89 EASTAYENUE, ROCHESTER, N. Y Id6d9-0001 AREA CODE 716 5'-2700 ROBERT C. MECREDY Vice President kucteor Operations May 5, 1995 U.S. Nuclear Regulatory Commission Document Control Desk Attn: Allen R. Johnson Project Directorate I-1 Washington, D.C. 20555
Subject:
Technical Specification Improvement Program Rochester Gas & Electric Corporation R.E. Ginna Nuclear Power Plant Docket No. 50-244
Dear Mr. Johnson,
Rochester Gas and Electric (RG&E) currently anticipates submitting the proposed conversion to Improved Technical Specifications (ITS) consistent with NUREG-1431 later this month. Included within this submittal are several changes to the existing technical specifications (TS) that are supported by engineering calculations or other documents that will require NRC Staff review and approval separate from the conversion review. Based on recent conversations, the NRC has requested that RG&E provide these documents prior to the ITS submittal to allow the NRC Staff additional time for review. Therefore, attached are the following documents:
(a) "Criticality Analysis of the R.E. Ginna Nuclear Power Plant Fresh and Spent Fuel Racks, and Consolidated Rod Storage Canisters," dated June 1994.
(b) WCAP-14040, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Revision 1, December 1994.
(c) RG&E Methodology for Determining the Low Temperature Overpressure Protection System (LTOPS) Setpoints.
The spent fuel pool criticality study is required to support an increase in allowed fuel enrichment necessary to support a conversion ,to 18 month fuel cycles which is planned to begin following the 1996 refueling outage. An increased in fuel enrichment affects current Ginna Station TS 5.3.1.b and TS 5.4.
~ GgQQ oi 95OS<S00<4 9S0505 PDR ADGCK 05000244 P PDR,';
The remaining two documents support the incorporation of a Pressure Temperature Limits Report (PTLR) which is planned to be implemented during the conversion to ITS. Since Ginna Station utilizes the LTOPS for preventing overpressurization of the residual heat removal system, Section 3.0 of WCAP-14040 does not apply to the installed system. Therefore, RG&E specific methodology is provided for your review. Both the RG&E specific methodology and a red-line markup of the methodology provided in WCAP-14040 are provided.
All items proposed to be relocated from the current Ginna Station TS to the PTLR are addressed by these two documents.
The items RG&E anticipates to be relocated to the Core Operating Limits Report (COLR) are provided in attached Table 1. As can be seen from this table, NRC approved methodology exists for all items proposed to be relocated such that no documents require submittal to the NRC at this time.
The current schedule for implementation of the ITS for Ginna Station indicates NRC Staff approval of the proposed new TS by November 1995. Therefore, RG&E requests that NRC review of these three documents be coordinated to support this schedule.
Ver truly yours, Robert C. Mec edy MDFK677 Attachments xc: U.S. Nuclear Regulatory Commission Mr. Allen R. Johnson (Mail Stop 14B2)
PWR 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
Table 1 R.E. Ginna Proposed COLR Parameters Ginna ITS COLR Parameter NRC Approved Methodology 3.1.1 Shutdown Margin Limits WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985 3.1.3 Moderator Temperature WCAP-9272-P-A, Westinghouse Coefficient BOL limits Reload Safety Evaluation Methodology, July 1985 3.1.3 Moderator Temperature WCAP-9272-P-A, Westinghouse Coefficient EOL Limits Reload Safety Evaluation Methodology, July 1985 3.1.5 Shutdown Bank Insertion WCAP-9272-P-A, Westinghouse Limit Reload Safety Evaluation Methodology, July 1985 3.1.6 Control Bank Insertion, WCAP-9272-P-A, Westinghouse Sequence, and Overlap Reload Safety Evaluation Limits Methodology, July 1985 3.2.1 Heat flux hot. channel WCAP-9272-P-A, Westinghouse factor (F<(Z) limits), K(Z) Reload Safety Evaluation curve, and equation Methodology, July 1985 WCAP-9220-P-A, Westinghouse ECCS Evaluation Model 1981 Version, Rev. 1, February 1982 3.2.2 Nuclear Enthalpy Rise Hot WCAP-9272-P-A, Westinghouse Channel Factor (F~ limit), Reload Safety Evaluation power factor multiplier and Methodology, July 1985 equation 3.2.3 AFD CAOC limits and target, WCAP-9272-P-A, Westinghouse band Reload Safety Evaluation Methodology, July 1985 WCAP-8385, Power Distribution Control and Load Following Procedures Topical Report, September 1974
COLR Parameter NRC Approved Methodology 3.3.1 OTaT WCAP-8745-P-A, Design Bases OPaT for the Thermal Overpower ~T and Thermal Overtemperature
~T Trip Functions, September 1986 3.4.1 DNB pressurizer pressure WCAP-8567-P-A, Improved limit, RCS Tavg limit, and Thermal Design Procedure, RCS total flow limit February 1989 WCAP-11397-P-A, Revised Thermal Design Procedure, April 1989 3.5.1 Accumulator boron WCAP-9272-P-A, Westinghouse concentration Reload Safety Evaluation Methodology, July 1985 3.5.4 RWST boron concentration WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985 3.7. 12 Spent Fuel Pool Boron WCAP-11596-P-A, PHOENIX-P/ANC Concentration Nuclear Design System for Pressurized Water Reactor Cores, June 1988 3.9.1 MODE 6/Refueling Boron WCAP-9272-P-A, Westinghouse Concentration Reload Safety Evaluation Methodology, July 1985
Attachment A Criticality Analysis of the R.E. Ginna Nuclear Power Plant Fresh and Spent Fuel Racks, and Consolidated Rod Storage Canisters
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