ML20113H515
| ML20113H515 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 01/18/1985 |
| From: | Musolf D NORTHERN STATES POWER CO. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| RTR-REGGD-01.150, RTR-REGGD-1.150 GL-83-15, NUDOCS 8501250186 | |
| Download: ML20113H515 (2) | |
Text
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Northem States Power Company 414 Nicollet Malt Minneapons. Minnesota 55401 Telephone (612) 330-5500 January 18, 1985 Director Office of Nuclear Reactor Regulation U S Nuclear Regulatory Commission Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Reactor Vessel Inservice Inspection The purpose of this Letter is to provide, for the information of the NRC Staff, a description of methods to be used during the performance and evaluation of reactor vessel inservice inspections at Prairie Island.
In an effort to avoid potential delays involved with the evaluation of reactor vessel inservice inspection indications, NSP contracted with Westinghouse to develop methodology which could be utilized to expedite this process.
In response, a detailed ASME Section XI IWB-3600 fracture mechanics analysis has recently been completed by Westinghouse for the entire Prairie Island Unit 1 and 2 reactor vessels. These analyses would be used if any indications are found which exceed the IWB-3500 acceptance standards. The analyses reflect the latest frac:ure mechanics technology and utilize the more accurate and valid criteria incorporated in Section XI of the ASME Code in recent years.
Our plans in this area were recently discussed with the NRC Licensing Project Manager for Prairie Island. This information is being provided, at his request, to document our intent to utilize the latest ASME Section XI fracture mechanics criteria during the Prairie Island reactor vessel inservice inspections being performed in 1985 under both our 1st and 2nd 10-year ASME Code Section XI Inservice Inspection Programs.
Specifically, indications found by volumetric examinations will be characterized, and then be accepted or evaluated, in accordance with ASME Section XI, 1983 Edition thru and including the Winter 1983 Addenda; except, the IWB-3640 criteria for austenitic steel will not be used, until approved by the NRC. The extent of volumetric examination and reportability are not affected. Also, as required by IWB-3610, if analytical evaluation is required for any indications which exceed the IWB-3500 acceptance standards, the evaluation details would be submitted for review by the NRC.
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Dirscter, NRR January 18, 1985 Page 2 Northem States Power Company The reactor vessel inservice inspections will be performed by Westinghouse using methodology which is based on Appendix A to Revision 1 of Reg.tlatory Guide 1.150 transmitted by the Commission in Generic Letter 83-15. 'the Westinghouse mathodology is documented by a Westinghouse Position Paper dated January 23, 1984.
Please contact us if you have any questions related to this information.
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David Musolf g
Manager - Nuclear Suppert Services DMM/EFE/ dab c: Regional Administrator-III, NRC NRR Project Manager, NRC Resident Inspector, NRC G Charnoff O
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