ML030970372
| ML030970372 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 03/25/2003 |
| From: | Solymossy J Nuclear Management Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-PI-03-020 | |
| Download: ML030970372 (3) | |
Text
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Committed to Nuclear ExcellenceJ" Prairie Island Nuclear Generating Plant Nuclear Management Company, LLC 1717 Wakonade Dr. East
- Welch MN 55089 March 25, 2003 L-PI-03-020 10 CFR 50.90 U S Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 License Amendment Request (LAR) dated March 25, 2003 Safety Analyses Transition Attached is a request for change to the Technical Specifications (TS), Appendix A for the Prairie Island Nuclear Generating Plant (PINGP) Units 1 and 2. The Nuclear Management Company (NMC) submits this request in accordance with the provisions of 10 CFR 50.90.
The Nuclear Management Company (NMC) has historically performed many of the safety analyses that support operation of the Prairie Island Nuclear Generating Plant (PINGP) with internal resources. In the future, NMC plans to have most of these safety analyses performed by Westinghouse. In the process of transitioning to Westinghouse safety analyses, NMC is making changes in plant operations and supporting documentation that involve Technical Specification changes. This license amendment request proposes TS changes which includes implementation of relaxed axial offset control (RAOC) of the reactor cores through changes in TS 3.2.1 and TS 3.2.3 and their Bases, relocation of selected operating parameters from TS 2.0 and TS 3.3.1 to the Core Operating Limits Report (COLR) and revision of the Pressurizer Pressure-Low Allowable Value. This proposed license amendment also includes, in TS 5.6.5, the appropriate references to the NRC approved methodologies which support relocation of operating parameters to the COLR.
In parallel with NRC review of this LAR, NMC will be performing an internal review of the Westinghouse methodologies under the requirements of 10 CFR Age o
USNRC NUCLEAR MANAGEMENT COMPANY L-PI-03-020 Page 2 of 3 50.59 and Westinghouse will be completing the applicable cycle specific safety analyses.
Exhibit A contains the description of the specific proposed changes, the reasons for requesting the changes, the supporting Safety Evaluation, and the proposed Determination of No Significant Hazards Consideration. Exhibit B contains Technical Specification and Bases pages marked up to show the proposed changes. Exhibit C contains the revised Technical Specification and Bases pages incorporating the proposed changes. Exhibit D provides Prairie Island instrument uncertainties for plant operating parameter inputs to the Westinghouse Revised Thermal Design Procedures (plant specific variances).
NMC requests that the NRC review and approve the proposed TS changes by December 31, 2003.
NMC is notifying the State of Minnesota of this LAR by transmitting a copy of this letter and attachments to the designated State Official.
This letter contains two new commitments in Exhibit A: 1) page 4 states, "The Al function input to the Overpower Delta-T Trip Function will be removed..."; and
- 2) page 8 states, "The resultant Design Limit [departure from nucleate boiling ratio] DNBR will be included in the Prairie Island Unit 1 and 2 Updated Safety Analysis Report." This letter does not contain revisions to existing commitments.
Please address any comments or questions regarding this LAR to Mr. Dale Vincent or myself at 1-651-388-1121.
I declare under penalty of perjury that the foregoing is true and accurate.
Executed on March 25, 2003.
(copies and attachments shown on page 3)
USNRC NUCLEAR MANAGEMENT COMPANY L-PI-03-020 Page 3 of 3 C:
Regional Administrator - Region 111, NRC Senior Resident Inspector, NRC NRR Project Manager, NRC Glenn Wilson, State of Minnesota
Enclosures:
Exhibit A, Description of Proposed Changes, Reasons for Requesting the Changes, Supporting Safety Evaluation and proposed Determination of No Significant Hazards Consideration Exhibit B, Marked Up Pages Exhibit C, Revised Pages Exhibit D, Instrument Uncertainties for Plant Operating Parameter Inputs to the Westinghouse Revised Thermal Design Procedure