ML17265A560
| ML17265A560 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 03/03/1999 |
| From: | Bajwa S NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML17265A561 | List: |
| References | |
| NUDOCS 9903100013 | |
| Download: ML17265A560 (6) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON' c
2055&400'i ROCHESTER GAS AND ELECTRIC CORPORATION DOCKET NO. 50-244 R. E. GINNA NUCLEAR POWER PLANT AMENDMENTTO FACILITYOPERATING LICENSE Amendment No.
License No. DPR-18 The Nuclear Regulatory Commission (the Commission or the NRC) has found that:
A. The application for amendment filed by the Rochester Gas and Electric Corporation (the licensee) dated November 24, 1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facilitywilloperate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance:
(i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities willbe conducted in compliance with the Commission's regulations; D. The issuance of this amendment willnot be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-18 is hereby amended to read as follows:
9'P03i000i3 990303 PDR ADQCK 05000244 P
PDR (2) Technical S ecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 73, are hereby incorporated in the license. 'The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORYCOMMISSION S. Singh Bajwa, Director Project Directorate 1-1 Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
March 3, 1999
ATTACHMENTTO LICENSE AMENDMENTNO. 73 FACILITYOPERATING LICENSE NO. DPR-18 DOCKET NO. 50-244 Replace the following pages of the Appendix A Technical Specifications with the attached pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Remove 4.0-1 5.0-20 5.0-21 Insert 4.0-1 5.0-20 5.0-21
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Design Features 4.0 A
4.0 DESIGN FEATURES 4.1 Site Location The site for the R.E.
Ginna Nuclear Power Plant is located on the south shore of Lake Ontario, approximately 16 miles east of Rochester, New York.
The exclusion area boundary distances from the plant shall be as follows:
Direction Distance m
N (including offshore)
NNE NE ENE E
WNW NW NNW 8000 8000 8000 8000 747 640 503 450 450 450 503 915 945 701 8000 8000 4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 121 fuel assemblies.
Each assembly shall consist of a matrix of zircaloy or ZIRLO clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO~) as fuel material.
Limited substitutions of
- zircaloy, ZIRLO, or stainless steel filler rods for fuel rods, in accordance with NRC approved applications of fuel rod configurations, may be used.
Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or cycle specific analyses to comply with all fuel safety design bases.
A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.
(continued)
R.E.
Ginna Nuclear Power Plant 4.0-1 Amendment No. Pg 73
p eporting Requirements 5.6 P
5.6 Reporting Requirements 5.6.5 COLR (continued) b.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.
(Hethodology for LCO 3. 1. 1, LCO 3.1.3, LCO 3. 1.5, LCO 3. 1.6, LCO 3.2. 1, LCO 3.2.2, LCO 3.2.3, and LCO 3.9.1.)
2.
3.
4.
5.
6.
7.
WCAP-13677-P-A, "10 CFR 50.46 Evaluation Model Report:
'COBRA/TRAC Two-Loop Upper Plenum Injection Model Updates to Support ZIRLO Cladding Option,"
February 1994.
(Methodology for LC0-3.2. 1.)
WCAP-8385, "Power Distribution Control and Load
'ollowing Procedures
- Topical Report," September 1974.
(Methodology for LCO 3.2.3.)
WCAP-12610-P-A, "VANTAGE + Fuel Assembly Reference Core Report,"
April 1995.
(Methodology for LCO 3;2. 1).
WCAP 11397-P-A, "Revised Thermal Design Procedure,"
April 1989.
(Methodology for LCO 3.4. 1 when using RTDP.)
WCAP-10054-P-A and WCAP-10081-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code,"
August 1985.
(Methodology for LCO 3.2. 1)
WCAP-10924-P-A, Volume 1, Revision 1,
"Westinghouse Large-Break LOCA Best-Estimate Methodology, Volume 1:
Model Description and Validation Responses to NRC guestions,"
and Addenda 1,2,3, December 1988.
(Methodology for LCO 3.2. 1)
(continued)
R.E.
Ginna Nuclear Power Plant 5.0-20 Amendment No. g 73
eporting Requirements
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5.6 Reporting Requirements 5.6.5 COLR (continued) 8.
WCAP-10924-P-A, Volume 2, Revision 2, "Westinghouse
- Large-Break LOCA Best-Estimate Methodology, Volume 2:
Application to Two-Loop PWRs Equipped with Upper Plenum Injection," and Addendum 1, December 1988.
(Methodology for LCO 3.2. 1) 9.
WCAP-10924-P-A, Volume 1, Revision 1,
Addendum 4, "Westinghouse Large-Break LOCA Best-Estimate Methodology, Volume 1: Model Description and Validation, Addendum 4: Model Revisions,"
March 1991.
(Methodology for LCO 3.2. 1) c.
The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDH, transient analysis limits, and accident analysis limits) of the safety analysis are met.
d.
The COLR, including any midcycle revisions or supplements, shall be'rovided upon issuance for each reload cycle to the NRC.
5.6.6 Reactor Coolant S stem RCS PRESSURE'AND TEMPERATURE LIMITS REPORT PTLR a ~
RCS pressure and temperature limits for heatup,
- cooldown, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:
LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits" b.
The power operated relief valve lift settings required to support the Low Temperature Overpressure Protection (LTOP)
- System, and the LTOP enable temperature shall be established and documented in the PTLR for the following:
MODE 4";
HODE 5, Loops Filled";
LCO 3.4. 10, "Pressurizer Safety Valves";
and LCO 3.4. 12, "LTOP System."
(continued)
R.E.
Ginna Nuclear Power Plant 5.0-21 Amendment No. g 73