IR 05000269/1982025
| ML20058C700 | |
| Person / Time | |
|---|---|
| Site: | Oconee, McGuire, 05000000 |
| Issue date: | 06/28/1982 |
| From: | Brinkley R, Bryant J, Burnett P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20058C675 | List: |
| References | |
| 50-269-82-25, 50-270-82-25, 50-287-82-25, 50-369-82-22, NUDOCS 8207260425 | |
| Download: ML20058C700 (4) | |
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NUCLEAR REGULATORY COMMISSION
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REGION 11 101 MARIETTA ST., N.W., SUITE 3100
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g ATLANTA, GEORGIA 30303
Report Nos. 50-269/82-25, 50-270/82-25, 50-287/82-25, and 50-369/82-22 Licensee: Duke Power Company P. O. Box 2178 Charlotte, NC 28242 Facility Names: Oconee Units 1, 2, and 3 and McGuire 1 Docket Nos. 50-269, 50-270, 50-287, and 50-369 License Nos. DPR-38, DPR-47, DPR-55, and NPF-9 Inspection at Duke Power Company Corporate offices in Charlotte, NC Inspectors:
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uut?/ f* 7 6--2[&2 R.1. Br'ickley Reactor Erp)in~eer, Region IV Date Signed
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P~. T' Burnett, Reactor Inspector, Region II Date Signed Accompanying Personnel:
J. L. Carter, NRR; T. A. Mcdonald, Argonne National F
Laboratory Approved by:
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J.' Br}6nt, S'ectjon Chief, Division of Project D4te Signed and Resident Programs SUMMARY Inspection on June 7-9, 1982 Areas Inspected This special announced inspection involved 44 inspector-hours on site in the area of validation of the computer code RETRAN.
Results In the areas inspected, no apparent violations or deviations were identified.
8207260425 820629 PDR ADOCK 05000269 G
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DETAILS 1.
Persons Contacted Licensee Employees
- K. S. Canady, Manager, Projects and Licensing
- J. F. Norris, Assistant Engineer, Licensing
- G. B. Swindlehurst, Associate Engineer, Reactor Safety
- P. M. Abraham, System Engineer J. Lee, Junior Engineer S. Nesbit, Junior Engineer
- Attended exit interview 2.
Exit Interview The inspection scope and findings were summarized by the inspector on June 9, 1982, with those persons indicated in paragraph 1 above. The senior management representative present committed to the actions identified in paragraphs 5.b.1 and 5.b.2 below. Additional management comments were for clarification or acknowledgement of the statements of the inspector.
3.
Licensee Action on Previous Inspection Findings Not inspected.
4.
Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, violations, or deviations.
The unresolved item disclosed during the inspection is discus-sed in paragraph Sa.
5.
Validation of Computer Codes This special inspection is part of a series of inspections being conducted at the request of NRR regarding the development, verification, and use of the RETRAN computer code by the nuclear industry. The basic objectives of these inspections are to assure that RETRAN computer code has been developed, maintained, and modified in accordance with the design control measures contained in ANSI N45.2.11-1974; and to verify that the organi-zation's computer code activities address the requirements of 10 CFR Part 21, 10 CFR Part 50.55e and 50.59 as applicable.
The Steam Production Department's Administrative Policy Manual for Nuclear Stations and the Reactor Safety Unit's Safety Analysis Procedures Manual were reviewed to determine that those design control measures contained in ANSI N45.2.11-1974 that are applicable to a code user were adequately addressed. The licensee's procedures for control and verification of inputs, interfaces and changes to the code were inspected against that standard.
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i The licensee agreement between the Electric Power Research Institute and Duke Power Company; computer code certificate for RETRAN-02,
M00002; code verification document for RETRAN-02, M0D002; Analysis No.
C-0SA-SA-82-004-0 (0vercooling Transient Evaluation for Reactor Vessel Thermal Shock); Analysis No. C-0SA-SA-82-001-0 (0vercooling Transient Evaluation for Reactor Vessel Thermal Shock); Analysis No. SRC-053-i SA-79-001-0 (HPIS Flow Split-LOCA in Injection Line); Anal C-0SA-SA-81-004-0 (HPIS Flow Split-LOCA in Injection Line)ysis No.
- and Analysis No. SRC-0SA-SA-81-001-0 (Emergency Feedwater System Minimum Flow Requirement) were examined for implementation of procedural requirements. No violations or aeviations were identified; however, the following unresolved and followup items were noted
a.
Unresolved Item It is not clear that the criteria used in the " Safety-Related Analysis Evaluation Checklist," to determine the classification of an analysis, satisfies the NRR definition of safety-related. Duke uses the following criteria in determining if an analysis is to be processed as safety-related:
(1)
if it determines the presence or absence of an unreviewed safety question or technical specification change, (2)
if it justifies a technical specification change, (3)
if it justifies a change in the performance or design of safety-related structures, systems, or components, or (4)
if it modifies the licensing basis of the safety-related analysis report.
The criteria above many be too limiting since the response to an NRC request for information may have the effect of forestalling an NRC request to change the technical specification or to reanalyze an accident, system or component. The issue to be resolved then is -
should every analysis performed to respond to an NRC request for information be treated as safety-relateo analysis.
This item will be referred to NRR for clarification and resolution and is identified as Unresolved Item 50-269/82-25-01.
b.
Followup Items 1.
Duke Power Company intends to revise Section 4.8.10.3(b) of the Administrative Policy Manual for Nuclear Stations to clarify their intention that the justification for use of a supervisor as an independent reviewer be documented.
(Ref. Table 17.0-1 of "QA ProgramDuke-1", Amendment 4.).
This action is to be completed within 90 days from the completion of this inspection (IFI 50-269/82-25-02).
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2.
Duke Power Company intends to revise Section 4.8 of the Admini-strative Policy Manual for Nuclear Stations to incorporate the Safety Analysis Unit (SAU) policy on reporting of errors in safety analysis (SAU Memo of May 10,1982) as a departmental policy.
This action is to be completed within 90 days from the completion of this inspection.
(IFI 50-269/82-25-03).
3.
The results of audits of the Steam Production Department by Duke Power Company QA personnel will be reviewed during a future inspection.
(IFI 50-269/82-25-04).