ML20147F915
| ML20147F915 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek, 05000000, Crane |
| Issue date: | 02/26/1988 |
| From: | Phyllis Clark GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| Shared Package | |
| ML20147F919 | List: |
| References | |
| C311-88-2025, NUDOCS 8803080070 | |
| Download: ML20147F915 (2) | |
Text
i..
s.
GPU Nuclear Corporation arsippany, New J ey 07054 201-316-7000 TELEX 136-482 Writer's Direct Dial Nurnber:
February 26, 1988 C311-88-2025 US Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 Gentlemen:
Oyster Creek Nuclear Station Docket No. 50-219 Provisional Operating License No. DPR-16 Three Mile Island Nuclear Station Unit 1 Operating License No. DPR-50 Docket No. 50-289 Safety Review Process GPU Nuclear Corporation Procedure 1000-ADM-1291.01, "GPU Procedure for Nuclear Safety and Environmental Impact Review and Approval of Documents" is the document which implements the safety review requirements of Section 6.5 of the Oyster Creek and Three Mile Island Unit 1 Technical Specifications.
On September 1,1986 GPUN implemented a major revision to the 1000-ADM-1291.01 Procedure, after documenting a detailed safety evaluation which established that the proposed changes did not require any change to the Technical Specifications, and that an Unreviewed Safety Question, as defined in 10 CFR 50.59, was not involved.
The new process requires an assessment of a proposed change relative to certain "screening criteria" which aid in the detemination of the necessity for a written safety evaluation.
This determination is the first step in the Safety Review Process, and requires a docur,'ent originator to assess the applicability of 10 CFR 50.59 and the potential impact of the change on nuclear safety.
If 10 CFR 50.59 is determined to be applicable to the change and/or there is a perceived potential impact on nuclear safety, then the originator must perfonn a written safety evaluation documenting the basis for his conclusion with respect to the existence of an Unreviewed Safety Question.
This constitutes the second step in the Safety Review Process.
This revision to the Safety Review Process evolved as a reelt of an in-house evaluation which concluded that many minor design and proceduiw changes with no impact on nuclear safety were undergoing documented safety evaluations, resulting in dilution of the overall Safety Review Process.
A mechanism was sought which would differentiate between changes to which 10 CFR 50.59 applied 8803080070 880226 PDR ADDCK 05000219
\\
P DCD 6365f/0170f GPU Nuclear Corporat on is a subsdary d General PutSc Utit es CorporatLon
I
{-
US Nuclear Regulatory Commission Page 2 and those beyond the scope of 10 CFR 50.59,.while not precluding a comprehensive assessment of impact on nuclear safety, i.e. a "graded approach *. The revision to Corpora :e Procedure 1000-ADM-1291.01 proceduralizes this "graded approact."
GPUN continues to assess the effectiveness of the "graded approach" to the Safety Review Process, and we anticipate that Corporate Procedure 1000-ADM-1291.01 will continue to undergo revision for enhancement.
One enhancement currently in progress is the incorporation of guidance provided to Responsible Technical Reviewers and Independent Safety Reviewers on Licensing Basis Documents.
A procedure revision for this enhancement is undergoing final review. We believe that our "graded approach" is sound and consistent with guidance provided by the Nuclear Safety Analysis Center of EPRI (Reference 1) which recently has been endorsed in the Nuclear Management and Resources Council "Draft Guidance on 10 CFR 50.59 Safety Evaluations",
(Reference 2). We also believe this "graded approach" is entirely consistent with the requirements and intent of 10 CFR 50.59.
A copy of the current revision of Corporate Procedure 1000-ADM-1291.01 is attached. We would appreciate your review of this document.
Sincerely, P. R. Clark President PRC:fg Ref. 1.
NSAC-105, "Guidelines for Design and Procedure Changes in Nuclear Power Plants", Final Report, July 1986.
2.
NUMARC/NSAC, "10 CFR 50.59 Guidance Document", transmitted for industry comment by NUMARC, November 12, 1987.
cc:
J. Stolz R. Hernan A. Dromerick S. Bryan L. Bettenhausen 6365f/017tf