ML20072L293
| ML20072L293 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 08/22/1994 |
| From: | Richard Anderson NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20072L296 | List: |
| References | |
| NUDOCS 9408310187 | |
| Download: ML20072L293 (2) | |
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Northem States Power Company 414 Nicollet Mall Minneapoks, Minnesota 55401-1927 Telephone (612) 330-5500 August 22, 1994 U S Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Request for Approval of Revision 6 of NSPNAD 8102-A, Prairie Island Nuclear
_fower Plant Reload Safety Evaluation Methods for Aeolication to PI Units The currently approved reload safety evaluation methods for Prairie Island are described in Revision 5 of NSPNAD-8102-A, " Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods For Application to PI Units". The purpose of this letter is to submit Revision 6 of NSPNAD-8102-A for NRC approval.
The minimum Departure from Nucleate Boiling Ratio (DNBR) limits used for accidents and transients are dependent on the Critical Heat Flux (CHF) correlations that are used. The two CHF correlations incorporated in Revision 5 are designated as WRB-1 and W-3. The changes proposed in Revision 6 are limited to modification of CHF correlations used to analyze the steam line break accidents only. These changes are as follows:
- 1. Remove reference to the WRB-1 correlation
- 2. Define the range of the W-3 correlation and the associated DNBR Limit as shown below:
Pressure Rance fosia)
DNBR Limit
>1000 1.30 500-1000 1.45 (The basis for this definition is provided by Reference 1.) provides a markup of the four affected pages of the topical (pages 184, 188, 189, and 261) and Attachment 2 provides typed versions of these pages, which constitutes Revision 6. It is noted that the new Reference 7, which has been added on page 261, is the letter that transmitted the NRC documents described in Reference 1 to this letter.
If you have any questions or comments please contact Mel opstad at 612-295-1653.
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USNRC August 22, 1994 Page 2 c: Regional Administrator III. NRC Senior Resident Inspector, NRC NRR Project Manager, NRC J E Silberg
Reference:
1.
NRC's safety Evaluation of the Westinghouse Electric Corporation Topical Report WCAP-9226-P/9227-NP, "Reector Core Response To Excessive Secondary Steam Releases" that was transmitted via a Jan 31, 1989 letter from Ashok C. Thadani to Mr. W. J. Johnson l
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