ML19345H508
| ML19345H508 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 05/14/1981 |
| From: | Mayer L NORTHERN STATES POWER CO. |
| To: | NRC |
| References | |
| TAC-8731, TAC-8732, TACS-8731, TACS-8732, NUDOCS 8105210065 | |
| Download: ML19345H508 (2) | |
Text
\\s Mr North'et States Power Company 414 Nconet man Minneapons. Minnesota 55401 Telephone (612)330 5500 May 14, 1981 9
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Director of Nuclear Reactor Regulation 2
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U S Nuclear Regulatory Commission Washington, DC 20555
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42
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50-306 DPR-60 W
Status of Modifications and Analyses to Resolve Issue of Asymmetric Blowdown Loads on Primary System In our letters dated October 25, 1978 and December 17, 1979 Northern States Power Company outlined plant modifications and analyses planned to resolve the asymmetric LOCA loads issue at Prairie Island. The purpose of this letter is to provide P.he NRC Staff with information related to the status of this work.
As noted in the earlier correspondence, our plan to resolve this issue includes:
a.
Install restraints to limit the break area at the reactor vessel nozzles to less than one square foot.
b.
Perform plant specific structural analyses of the reactor vessel, vessel supports, nozzles, piping, i
fuel, ECCS piping, CRD mechanisms, reactors internals, l
and structural concrete.
l i
Replacement of vessel nozzle inspection port sand plugs u.
with redesigned low-mass pluge Pipe restraints were designed and f abricated by our architect-engineer, Fluor l
Power Services, for both units.
Installation was completed during the Autumn 1980 refueling of Unit No. I and the Spring 1981 refueling of Unit No. 2.
l A preliminary structural analysis report was submitted to the NRC in a letter dated April 11, 1980 from T M Anderson, Manager, Nuclear Safety Department, Westinghouse Electric Corporation, to Darrel G Eisenhut, Director, Division of Licensing, USNRC. The following evaluations have been completed:
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NORTHERN STATES POWER COMPANY e
Director of NRR May 14, 1981 Page 1.
Reactor pressure vessel and reactor coolant loop hydraulic forcing functions 2.
Reactor cavity pressure 3.
Reactor vessel motion 4.
Reactor coolant loop support 5.
Reactor internals 6.
Westinghouse fuel The remaining items to be addressed are:
1.
ECCS piping 2.
CRDM's 3.
Structural concrete 4.
Exxon fuel Pipe restraint design and fabrication as well as the structural analysis work has been performed jointly for the Prairie Island and Kewaunee plants as explained in our October 25, 1978 letter.
Following completion of the remaining analytical work, a final Westinghouse topical report will be submitted for NRC Staff review.
Seventeen questions relating to the preliminary structural analysis submitted on April 11, 1980 were forwarded to us by the Prairie Island NRC Project Manager in the Division of Licensing on November 10, 1980. A response to these questions will be prepared by Westinghouse following a meeting with the NRC Staff. A meeting was requested earlier this year to further define what must be supplied.
The schedule for addressing these questions and completing the final topical report is uncertain at this point.
The plant technical staff is now preparing a design change package to replace the existing reactor vessel nozzle inspection port sand plugs (total of six per unit) with a redesigned low-mass plug. This new plug will be designed to blow out in the event of a reactor coolant system pipe break in the reactor annulus area without presenting a missile hazzard. During operation, inspection port covers will be removed. This modification is now planned for the Autumn 1981 Unit No. I refueling outage and the Spring 1982 Unit No. 2 refueling outage.
Please contact us if you have any questions related to our plans for final resolution of this issue.
.l L 0 Mayer, PE Manager of Nuclear Support Services LOM/DMM/jh cc J G Keppler i
G Charnoff NRC Resident Inspector
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