ML17263B045

Forwards Listed Doucments for NRC Review & Approval Prior to Submittal of Improved TS Consistent w/NUREG-1431

ML17263B045
Person / Time
Site:	Ginna
Issue date:	05/05/1995
From:	Mecredy R ROCHESTER GAS & ELECTRIC CORP.
To:	Andrea Johnson NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
Shared Package
ML17263B047	List: ML17263B045 ML17263B048
References
RTR-NUREG-1431 NUDOCS 9505150014
Download: ML17263B045 (8)

Text

PRIORIT'Y Z REGULAT(<NAORMNEetH VTSVREMTION4STEM (RZDS)ACCESSION NBR:9505150014 DOC.DATE: 95/05/05 NOTARIZED:

NO FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1 Rochester G 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)

SUBJECT:

Forwards listed doucments for NRC review 6 approval prior submittal of improved TS consistent w/NUREG-1431.

DISTRIBUTION CODE: A001D COPIES RECEIVED:LTR ENCL SIZE: TITLE: OR Submittal:

General Distribution DOCKET I 05000244 to P NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72).

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:.

PLEASE HELP US TO REDUCE WASTE!CONTACTTHE DOCL'MENTCONTROL DESK, ROOII I PI-37 (EXT.504-20S3)TO ELIMIYATE YOUR 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 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/