ML17056C146
| ML17056C146 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 12/04/1992 |
| From: | Murley T Office of Nuclear Reactor Regulation |
| To: | Ridings B RIDINGS, B.L. |
| Shared Package | |
| ML17056C147 | List: |
| References | |
| 2.206, NUDOCS 9212100098 | |
| Download: ML17056C146 (12) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 December 4, 1992 Docket No. 50-220 Ben L. Ridings P.O.
Box 1101
- Kingston, Tennessee 37763
Dear Mr. Ridings:
On October 27,
- 1992, you filed a "Petition for Emergency Enforcement Action and Request for Public Hearing" (Petition) regarding Nine Mile Point Nuclear Station Unit No.
1 (NMP-1) with the Nuclear Regulatory Commission (NRC or Commission).
You requested that the Commission take direct review of the Petition.
The Commission has declined to take direct review of your Petition and has referred the Petition to me for consideration pursuant to 10 CFR 2.206.
The Petition requests that the NRC immediately order Niagara Mohawk Power Corporation (NMPC) to cease power operation of NMP-1 and place the reactor in a cold shutdown condition.
The Petition also requests that the Commission hold a public hearing before authorizing resumption of plant operation.
You seek relief based on allegations that:
(1)
NMP-1 does not meet NRC requirements for an engineered safety feature system (ESFS) grade high-pressure coolant injection (HPCI) system, (2) 45 percent of the containment isolation valves have administrative deficiencies, and (3)
- group, and the NRC have reviewed these safety concerns
- and, contrary to any practical justification, have remained silent.
With respect to the lack of an ESFS grade HPCI system, you had two concerns:
(1) you stated that the feedwater
- system, which can operate in an HPCI mode, is not an acceptable alternative system because it does not have a backup electrical power supply provided by an onsite emergency diesel generator and (2) you stated concern about using the feedwater system in an HPCI mode because some 44 out of 47 valves in the feedwater injection flow path are not included in the NMP-1 Inservice Test Program for pumps and valves.
Although NMP.-l does not have an ESFS grade HPCI system, the plant was designed and licensed by the NRC with other emergency core cooling system (ECCS) equipment that-provides adequate protection against all loss-of-coolant accidents.
The Commission's regulations in 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light water nuclear power reactors,"
require that licensees provide their plants with ECCS's designed to meet the criteria set forth in that section.
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Ben L. Ridings December 4, 1992 In response to your request for an immediate shutdown of NHP-1, my staff has reviewed the NRC's Safety Evaluation Report for NHP-1 dated December 27,
- 1974, and General Electric's (GE) loss-of-coolant accident analysis (NEDC-31446P) for the current fuel cycle (Cycle 10) concerning NHP-1's conformance to the requirements of 10 CFR 50.46.
The December 27, 1974, Safety Evaluation Report concluded that the NHP-1 ECCS satisfies the requirements of 10 CFR 50.46.
Furthermore, the Updated Final Safety Analysis Report (UFSAR) states that the HPCI is not an engineered safety feature system and, therefore, is not relied on in meeting the criteria of 10 CFR 50.46.
This conclusion was reaffirmed in GE's loss-of-coolant accident analysis for the current fuel cycle as well as in the previous reload cycles.
GE's analysis was prepared in response'to the requirements of NHP-1 Technical Specification 6.9.lf, "Reporting Requirements, Core Operating Limits Report."
The NHP-1 ECCS satisfies the requirements of 10 CFR 50.46 by utilizing the automatic depressurization system (ADS) and the core spray system (CSS),
both of which have redundancy and are supplied backup electrical power by the NHP-1 onsite emergency diesel generators.
The CSS in conjunction with the ADS is designed to accommodate the range of loss-of-coolant accidents from the smallest up to the largest line break.
For line breaks smaller than; 0.30 square foot, reactor pressure may not decrease rapidly enough to prevent clad overheating if ther e is no feedwater flow.
Therefore, the ADS is provided to depressurize the reactor so that the CSS can inject water into the reactor.
Because oper ation of the feedwater pumps in the HPCI mode is not required to meet the requirements of 10 CFR 50.46, an onsite emergency electrical power supply for the feedwater pumps is not required.
The NHP-1 Technical Specifications require the feedwater system to be operable in the HPCI mode as the normal means for core cooling; however., this system is not relied on to satisfy the requirements of 10 CFR 50.46.
Furthermore, the valves in the feedwater flow path are not required to be included in the NHP-1 inservice testing program because the feedwater system is not required to meet 10 CFR 50.46.
You also asserted that the NHP-1 feedwater system operating in the HPCI mode fails to 'meet GDC 33, 35, 36 and 37.
As stated in a Staff Requirements Memorandum dated September 18, 1992, the Commission has determined that the General Design Criteria in 10 CFR Part 50, Appendix A, do not apply to plants with construction permits issued prior to Hay 21, 1971.
At the time of promulgation of Appendix A to 10 CFR Part 50, the Commission stressed that the GDC were not new requirements and were prom'ulgated to more clearly articulate the licensing requirements and practice in effect at that time.
While compliance with the intent of the GDC is important, each plant licensed before the GDC were formally adopted was evaluated on a plant specific basis, determined to be safe, and licensed by the Commission.
Furthermore, current regulatory processes are sufficient to ensure that plants continue to be safe and comply with the intent of the GDC.
Plants with construction permits issued prior to Hay 21,
- 1971, do not need exemptions from the GDC.
Ben L. Ridings December 4, 1992 On the basis of the fore'going discussions, I have concluded that there is no basis to issue an immediately effective order to shut down NHP-1 because of the unavailability of an ESFS grade HPCI system.
Your Petition also stated that 45 percent of the primary containment isolation valves at NHP-1 had administrative deficiencies as indicated in Attachment 5
to your Petition.
- However, some of the valve identification numbers listed in Attachment 5 are not fully legible, and for these
- valves, we were unable to evaluate your concerns.
Note 17 applicable to valves listed on pages 1, 3, and 4 of Attachment 5 was not provided; for these
- valves, my staff reviewed the existing regulatory requirements and NHPC's procedures and programs for implementing those requirements and found no deficiencies.
The NRC staff had previously identified, through its inspection program, administrative deficiencies, similar to those identified in Attachment 5, with reactor coolant system isolation valves and containment isolation valves listed in the NMP-1 Technical Specifications and the UFSAR.
In a safety evaluation dated Hay 6, 1988, the NRC staff requested NHPC to resolve these administrative deficiencies.
Subsequently, by letter dated November 20,
- 1990, as superseded by letter dated February 7,
- 1992, NHPC submitted a request for a license amendment to update the NHP-1 Technical Specifications to resolve these administrative deficiencies.
Our review of this request is in progress and although we have not yet completed our review, we have reviewed your. concerns with respect to the current NHP-1 Technical Specifications, the
- UFSAR, and the most recent Inservice Testing Program for NHP-1 pumps and valves.
Our preliminary review indicated that NMPC is implementing adequate surveillance testing and leakage-rate testing procedures to verify valve and containment operability.
These procedures include functional testing required by the NHP-1 Technical Specifications to ensure that valves required to close during accident conditions function properly on receipt of a signal to close.
Furthermore, periodic valve exercising, stroke-time testing, and leakage-rate testing ensure that the inservice testing program and applicable technical specification requirements are met.
All the above testing provides reasonable assurance that NHP-1 can be operated without undue risk to the public health and safety in light of the described administrative deficiencies in isolation valves.
In addition, our preliminary conclusions are that the, current technical speci.fications,'he license amendment request previously discussed, or the UFSAR address most of these administrative deficiencies.
Based on the
- above, I have. concluded. that an immediately effective order to shut down NHP-1 on the basis. of'he identified administrative deficiencies with the containment isolation valves is not required.
As stated
- above, our review of the Petition has disclosed that some specific information in your Petition was not fully legible or not provided.
The NRC staff has been unable to contact you by telephone to obtain the missing information.
In order for the NRC.to provide a complete review of your
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Ben L. Ridings December 4, 1992
- concerns, we request that you provide the missing information promptly but, in any event, within 25 days of the date of this letter in order for us to consider it in our evaluation of your Petition.
You may provide the missing information by contacting Hr. Donald S. Brinkman, the NRC's Project Hanager for NHP-1 at (301) 504-1409.
With regard to your allegation that the NRC staff has previously reviewed these safety concerns and has remained silent, a copy of the Petition has been referred to the NRC Office of the Inspector Gene} al for whatever review and action the Inspector General deems appropriate.
The NRC staff will review your Petition in accordance with 10 CFR 2.206.
I will issue a final decision with regard to your Petition within a reasonable time.
A copy of the notice that is being filed for publication with the Office of the Federal Register is enclosed for your information.
This requirement affects one respondent and, therefore, is not subject to Office of Hanagement and Budget review under P.L.96-511.
Sincerely,
Enclosure:
Federal Register Notice cc w/enclosure:
See next page Thomas E. Hurley, Director Office of Nuclear Reactor Regulation
Niagara Mohawk... Power Corporation Nine Mile Point Nuclear Station Unit No.
1 CC:
Hark J. Wetterhahn, Esquire Winston
& St} awn 1400 L Street, NW Washington, DC 20005-3502 Supervisor Town of Scriba Route 8, Box 382
- Oswego, New York 13126 Hr. Neil S.
Cams Vice President - Nuclear Generation Niagara Mohawk Power Corporation Nine Mile Point Nuclear St'ation Post Office Box 32
- Lycoming, New York 13093 Resident Inspector U.S. Nuclear Regulatory Commission Post Office Box 126
- Lycoming, New York 13093 Gary D. Wilson, Esquire Niagara Hohawk Power Corporation 300 Erie Boulevard West
- Syracuse, New York 13202 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania 19406 Ms.
Donna Ross New York State Energy Office 2 Empire State Plaza 16th Floor
- Albany, New York'2223 Mr. Kim Dahlberg Unit 1 Station Superintendent Nine Mile Point Nuclear Station Post Office Box 32
- Lycoming, New York 13093 Hr. David K. Greene Manager Licensing Niagara Mohawk Power Corporation 301 Plainfield Road
- Syracuse, New York 13212 Charles Donaldson, Esquire Assistant Attorney General New York Department of Law 120 Broadway New York, New York 10271 Hr. Paul D.
Eddy State of New York Department of Public Service Power Division, System Operations 3 Empire State Plaza
- Albany, New York 12223 Hr. B. Ralph Sylvia Executive Vice President, Nuclear Niagara Mohawk Power Corporation 301 Plainfield Road
- Syracuse, New York 13212
NMP1 2.206 Acknowledgement Letter Date December 4, 1992 Distribution:
Docket Fi.le (50=220)--w/incoming -letter NRC/Local PDRs w/incoming-letter----'DO¹ 8255 THurley/FHiraglia, 12/G/18
12/G/18 PDI-1 Reading JGoldberg, OGC 15/B/18
- CCowgill, RGN-1
- RJones, 8/E/23 CHcCracken, 8/D/1
- JNorberg, 7/E/23