ML20153B614
| ML20153B614 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 09/16/1998 |
| From: | Sorensen J NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9809230191 | |
| Download: ML20153B614 (4) | |
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Northern States Power Company Prairie Island Nuclear Generating Plant 1717 Wakonade Dr. East Welch, Minnesota 55089 September 16,1998 10 CFR 50.9 U S Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 l
PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 50.9 Report - Containment Heavy Load Weight issues On August 11 and 12,1998, NSP notified NRC Region til staff, in accordance with 10CFR50.9, of a case where information supplied to the NRC may not be complete and accurate in all material respects.
Background - Reactor Vessel Head Weight On December 9,1981, NSP (in a response to the NRC Generic Letter, " Control of Heavy Loads," dated December 22,1980) informed the Director, Office of Nuclear Regulation in Attachment lli, page 51, that the Vessel Head weighing 80,925 lbs. was one of the major loads handled by the Containment Polar Crane. While this is true, the reactor vessel head usually has the following additional loads included during a move between the reactor vessel flange area and the vessel head stand:
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Lift rig and platform e
Stud tensioner hoists p 00)
Personnel (usually 3)
The effective total weight, then, is nominally 177,000 lbs.
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l USNRC NORTHERN STATES POWER COMPANY September 16,1998 Page 2 In an analysis of the reactor vessel head lifting rig, Westinghouse used a design weight of 186,900 lbs, including coni..,gencies.
The December 9,1981 letter, Item 2.3.4 described analyses performed to demonstrate compliance with the NUREG-0612 Criteria i through 111 and guidelines of Appendix A for the case of an accidental drop of the reactor vessel head. The conclusion of that analysis, using conservative assumptions, was that both 4-inch diameter Si lines would remain functional. The analysis noted the reactor coolant loop may crack due to the load drop, which is likely an overly-conservative assumption for the current plant condition since the reactor coolant system loop restraints have been removed.
On June 6,1983, the NRC issued the Safety Evaluation Report for Phase I Control of Heavy Loads. The attached Franklin Research Institute Technical Evaluation Report TER-C5506-384/385, Table 2.1 reported the vessel head weight as 40.5 tons. The Safety Evaluation Report concluded that the Phase I requirements were met.
On May 31,1984, NRC transmitted a draft of TER-C5506-469/470 for the Phase Il program. Section 2.3.2 contained an evaluation of the December 9,1981 letter, item 2.3.4 response that concluded that the ability to keep the core adequately covered following imposition of the worst-case conditions had been demonstrated.
Related Issue - Upper Internals Weight The December 9,1981 NSP letter and NRC letter of June 6,1983 reported the upper internals weight as 50,000 lb. New upper internals have been installed which have a weight of approximately 60,000 lb.
The Westinghouse analysis for the ir,ternals lifting rig used a design weight of 202 kips, the weight of the lower internals. This analysis is bounding for movements of the upper internals.
Conclusion l
The reactor vessel head and upper and lower internals weights are well within the 230 l
ton capacity of the containment polar crane. The Operations Manual procedures for reactor disassembly and reactor assembly have used appropriate weights.
The reactor vessel head drop calculations in support of the December 9,1981 item 2.3.4 would have significant implication only if they are found to be non-conservative such that adequate core cooling could not be assured in the event of the worst-case reactor vessel head drop.
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USNRC NORTHERN STATES POWER COMPANY September 16,1999 Page 3 J
Action Taken to Date I
The original Fluor (the Prairie Island architect engineer) calculations, as available, have been reviewed. The calculation showing Si line integrity identified that a 0.55-inch deflection would be produced from the drop with an empty refueling pool. The original calculation used to determine the deflection, for the cases with the refueling pool empty and full, has not been found. This latter calculation would have intded the assumed reactor vessel head weight.
l Experienced personnel, familiar with the plant and the original methodology, have been employed to perform the calculations with the current plant configuration to provide a l
basis for future reactor vessel head movements.
l Action Plan i
The following steps are being taken prior to the next refueling outage:
- 1. If possible, locate reactor vessel head drop calculation to determine weight used.
l
- 2. If the reactor vessel head weight is less than the typical weight or the original j
calculations are unavailable, then calculate deflection due to dropping the head from l
an anticipated elevation.
- 3. If the deflection is less than previously calculated for a dry pool, no additional calculations are necessary.
- 4. If the deflection is greater than previously calculated for a dry pool, determine if the 4-inch SI lines will stay intact or if the RCS legs will remain intact. If either of these conditions are met, no additional calculations are necessary.
- 5. Alternatively, an acceptable deflection could be calculated working back from the allowables for the Si line.
- 6. Determine if the drop calculations provide sufficient proof that previous conclusions are still applicable.
- 7. Report results ofinvestigation and analysis to NRC.
l l
Relevant References related to the Control of Heavy Loads issue l
1.
Letter, L.O. Mayer (NSP) to Director Nuclear reactor Regulation (NRC), " Control of Heavy Loads (nine month submittal)," dated December 9,1981 2.
Generic Letter, Darrell G. Eisenhut (NRC) to All Licensees (PWR), " Control of Heavy Loads," dated December 22,1980 4
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USNRC NORTHERN STATES POWER COMPANY September 16,1998 Page 4 3.
Letter, Robert A. Clark (NRC) to D.M. Musolf (NSP), " Review of the Phase I Control of Heavy loads issue for Prairie Island Nuclear Generating Plant," dated June 6,1983 4.
Letter, James R. Miller (NRC) to D.M. Musolf (NSP), " Draft Technical Evaluation Report (TER) related to the Control of Heavy Loads Phase 11 for the Prairie Island Nuclear Generating Plant," dated May 31,1984 Additional Information Should you require additional information on this issue, please contact Joseph Gonyeau, Senior Nuclear Consultant, at 651-388-1121.
Commitment In this letter we have made one new Nuclear Regulatory Commission commitment: We will complete the items indicated in the Action Plan section prior to the next Reactor Vessel Head lift on either unit.
Joel P. Sorensen Plant Manager Prairie Island Nuclear Generating Plant c: Regional Administrator-Region Ill, NRC Senior Resident inspector, NRC NRR Project Manager, NRC J E Silberg i
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