ML17059B943
| ML17059B943 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 03/19/1998 |
| From: | Hood D NRC (Affiliation Not Assigned) |
| To: | Mueller J NIAGARA MOHAWK POWER CORP. |
| References | |
| GL-88-20, TAC-M83645, NUDOCS 9804020012 | |
| Download: ML17059B943 (10) | |
Text
Mr. John H. Mueller
'hief Nuclear Officer
'Niagara Mohawk Power Corporation Nine Mile Point Nuclear Station Operations Building, Second Floor P.O. Box 63 Lycoming, NY 13093 March 19, 199
SUBJECT:
REQUEST FOR ADDITIONALINFORMATIONREGARDING INDIVIDUALPLANT EXAMINATIONOF EXTERNALEVENTS, NINE MILE POINT NUCLEAR STATION UNIT NO. 1 (TAC NO. M83645)
Dear Mr. Mueller:
The NRC staff, with assistance from Brookhaven National Laboratories, is reviewing your submittal of August 29, 1996, regarding the Individual Plant Examination of External Events performed in accordance with Generic Letter 88-20, Supplement 4, for Nine Mile Point Nuclear Station, Unit 1 (NMP1). We find that additional information is necessary to complete our review.
Your response to the enclosure is requested within 60 days of this letter to support our current review schedule.
Ifyou have questions regarding the enclosure or are unable to meet the requested response date, please contact me by phone at (301) 415-3049 or by e-mail at dsh@nrc.gov.
Sincerely, ORIGINAL SIGNED BY:
Docket No. 50-220 Darl S. Hood, Senior Project Manager Project Directorate l-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Enclosure:
Request for Additional Information ccw/encl: See next page
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Yiarch 19, 1998 Mr. John H. Mueller Chief Nuclear Officer Niagara Mohawk Power Corporation Nine Mile Point Nuclear Station Operations Building, Second Floor P.O. Box 63 Lycoming, NY 13093
SUBJECT:
REQUEST FOR ADDITIONALINFORMATIONREGARDING INDIVIDUALPLANT EXAMINATIONOF EXTERNALEVENTS, NINE MILE POINT NUCLEAR STATION UNIT NO. 1 (TAC NO. M83645)
Dear Mr. Mueller:
The NRC staff, with assistance from Brookhaven National Laboratories, is reviewing your submittal of August 29, 1996, regarding the Individual Plant Examination of External Events performed in accordance with Generic Letter 88-20, Supplement 4, for Nine Mile Point Nuclear Station, Unit 1 (NMP1). We find that additional information is necessary to complete our review.
Your response to the enclosure is requested within 60 days of this letter to support our current review schedule.
Ifyou have questions regarding the enclosure or are unable to meet the requested response date, please contact me by phone at (301) 415-3049 or by e-mail at
'sh@nrc.gov.
Sincerely,
/~4/6 Darl S. Hood, Senior Project Manager Project Directorate l-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Docket No. 50-220
Enclosure:
Request for Additional Information cc w/encl: See next page
I'
John H. Mueller Niagara Mohawk Power Corporation Nine Mile Point Nuclear Station Unit No. 1 CC:
Regional Administrator, Region I
U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Resident Inspector'.S.
Nuclear Regulatory Commission P.O. Box 126 Lycoming, NY 13093 Charles Donaldson, Esquire Assistant Attorney General New York Department of LaW 120 Broadway New York, NY 10271 Mr. Paul D. Eddy
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State of New York
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Department of Public Service Power Division, System Operations 3 Empire State Plaza Albany, NY 12223 Mr. F. William Valentino, President New York State Energy, Research, and Development Authority Corporate Plaza West 286 Washington Avenue Extension Albany, NY 12203-6399 Mark J: Wetterhahn, Esquire Winston 8 Strawn 1400.L Street, NW Washington, DC 20005-3502 Gary D. Wilson, Esquire Niagara Mohawk Power Corporation 300 Erie Boulevard West
- Syracuse, NY 13202 Supervisor Town of Scriba Route 8, Box 382 Oswego, NY 13126
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5.
Please describe the method for generating the IPEEE in-structure response spectra, and include some spectra that have been used for the component high-conflidence-of-low-probability-of-failure (HCLPF) calculations.
Please also. provide spectral comparisons between the IPEEE review-level-earthquake (RLE) input and the original design basis input.
Please provide detailed HCLPF calculations for battery boards 11 and 12, containment spray raw water pumps, the battery racks, and masonry block walls ¹23, ¹27 and ¹53.
Hi h Winds Floods a
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Even s 6.
NMP1 does not meet the criteria in the 1975 Standard Review Plan for floods. Therefore, according to NUREG-1407, the hazard frequency must be shown to be acceptably low (the probability of flooding any safety-related equipment is to be less than 1E-6 per year), or a flood probabilistic risk assessment (PRA) should be performed.
The maximum water height for which no safety-related equipment willbe failed is not explicitly given (although the grade level is given as 261 feet). Nor is any estimate given of the frequency of exceeding whatever water level the plant can withstand without failure of any safety-related equipment.
Please provide a determination of the maximum flood water level which will not fail any safety-related equipment, and estimate the frequency of exceeding this flood water level.
Alternatively, provide a PRA for the flood hazard.
7.
The revised probable maximum precipitation criteria given in hydrometeorological reports HMR-51 and HMR-52 were used to calculate a flood depth of 262.85 feet.
Because the diesel generators willfail ifflooding in the turbine building reaches 261'5", further analysis is required to determine ifthe plant can withstand a probable maximum precipitation event.
Your submittal appears to argue that the water in the turbine building would first go to lower elevations in the building, and by the time the water level would reach 261'5", the water level outside the buildings would have subsided.
However, no calculations are presented to justify this conclusion.
Please provide calculations to justify this conclusion. Also, what are
'he consequences of water levels less than 261'5" in the turbine building'7
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REQUEST FOR ADDITIONALINFORMATION REGARDING INDIVIDUALPLANT EXAMINATIONOF EXTERNALEVENTS NIAGARAMOHAWKPOWER CORPORATION NINE MILE POINT NUCLEAR STATION UNIT NO. 1 OPERATING LICENSE NUMBER DPR-63 DOCKET NO. 50-220 Seismic 1.
According to EPRI NP-6041-SL, "a preferred success path and an alternate success path based on operational and systems considerations" should be identified for seismic margin assessment, and "at least one of these paths must be able to cope with a small, seismically induced leakage unless such leakage can be ruled out by walkdown." (p. 3-1)
Although two success paths are identified in the NMP1 IPEEE submittal, the equipment in only one path was included in the safe shutdown equipment list (SSEL) for seismic margin assessment.
The reasons given in the submittal to exclude the other path from evaluation include statements that the systems in the path are "unlikelyto support success during LOCA conditions" and that "the additional effort involved in demonstrating low probability of no LOCA could be significant and may not be successful."
Your position is not consistent with the EPRI NP-6041-SL guidelines quoted above.
According to the guidelines, one should select two paths, with both paths able to cope under no LOCA condition, but only one of them needs to cope with a small LOCA. Therefore, both NMP1 success paths are needed to meet the EPRI NP-6041-SL guidance.
Please expand the SSEL to include the equipment in the second success path and provide information on the results of a,seismic margin evaluation of the expanded list.
2.
Table 7-1 of the NMP1 IPEEE submittal lists a number of plant improvements that were slated for implementation in the spring of 1997 during refueling outage 14. Please provide the current status of these improvements.
3.
The Cardox system mercury relays are known to be poor performers during seismic events and could inadvertently actuate fire protection systems.
Table 7-1 of the NMP1 IPEEE submittal calls for replacement of these relays or a procedure change.
Please clarify the final resolution for these mercury relays.
Ifa procedure change has been or will be implemented, please identify the specific operator actions involved and discuss the timing, stress, access and environmental concerns associated with operator actions following a seismic event.
4.
Please discuss the reliabilityof operator initiation of the Automatic Depressurization System after it has been manually disabled early in the accident.
You state that such reliabilitywould be high because the operators have disabled the system previously, and thus, they would be cognizant of its status.
However, there are issues of stress, timing, multiple demands on operator attention, etc., that are not discussed.
Please verify that such issues have been properly considered and that such operator action willindeed be executed with high reliability.
Enclosure
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