IR 05000280/1986019
| ML18149A266 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 08/20/1986 |
| From: | Belisle G, Runyan M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18149A265 | List: |
| References | |
| 50-280-86-19, 50-281-86-19, NUDOCS 8609100065 | |
| Download: ML18149A266 (9) | |
Text
Report Nos.:
UNITED STATES NUCLEAR REGl)LATORY COMMISSION
REGION II
101 MARIETTA STREET, ATLANTA, GEORGIA 30323 50-280/86-19 and 50-281/86-19 Licensee:
Virginia Electric and Power Company Richmond, VA 23261 Docket Nos.:
50-280 and 50-281 Facility Name:
Surry 1 and 2 License Nos.:
DPR-32 and DPR-37 Inspection Conducted:
July 28 - August 1, 1986 l
'tf[;:>
Inspector:
F'"; )~v\\,,.r::-t.----. Runyan Approved by: ({i:/~I(.:;;;;
G.LK. Befis'fe;vActing Section Chief Division of Reactor Safety SUMMARY Date Signed Date Signed Scope: This routine, unannounced inspection was conducted in the areas of design control and tests and experiment Results:
No violations or deviations were identified *
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8609100065 860826
PDR ADOCK 05000280 G
REPORT DETAILS Persons Contacted Licensee Employees
- D. Benson, Assistant Station Manager S. Burgold, Operations Coordinator D. Christian, Superintendent, Operations
- N. Clark, Quality Assurance (QA) Manager R. Coupe, Design Control Engineer
- W. Craft, Licensing Coordinator C. Duong, Associate Engineer
- W. Grady, Supervisor, Quality
- E. Grecheck, Superintendent, Technical Services S. Hurst, Quality Specialist E. Kopanski, Supervisor, Quality Surveillance
- J. Logan, Supervisor, Site Engineering Services E. May, Supervisor, Plant Engineering D. McClure, Quality Specialist D. Nolan, Supervisor, Construction Engineering J. Rayno, Engineer D. Sanderson, Senior Instructor, Nuclear D. Schaffer, Supervisor, QA Engineering P. Tucker, Site Engineering Officer J. Wroniew*icz, Supervisor, Nuclear Project Engineering Other licensee employees contacted included office personne NRC Resident Inspectors
- W. Holland, Senior Resident Inspector
- M. Davis, Resident Inspector
- Attended exit interview Exit Interview The inspection scope and findings were summarized on August 1, 1986, with those persons indicated in paragraph 1 abov The inspector described the areas inspected and discussed in detail the inspection findings listed belo No dissenting comments were received from the license Inspector. Followup Item:
Emergency Vent Damper Environmental Qualification (EQ), paragraph The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspector during this inspectio *
3 Licensee Action on Previous Enforcement Matters This subject was not addressed in the inspectio.
Unresolved Items Unresolved items were not identified during the inspectio.
Design Program (37702)
References:
(a)
10 CFR 50 Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Pl ants, Criterion II I (b)
10 CFR 50.54(a)(l), Conditions of Licenses (c)
VEPCO Quality Topical Report, VEP 1-5A (d)
Regulatory Guide 1.64, Quality Assurance Requirements for the Design of Nuclear Power Plants (e)
ANSI N45.2.11-1974, Quality Assurance Requirements for the Design of Nuclear Power Plants (f) Regulatory Guide 1.33, Quality Assurance Requirements (Operations)
(g)
ANSI NlB.7-1976, Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants (h)
10 CFR Part 50.59, Changes, Tests and Experiments (i) Technical Specifications, Section 6 The inspector reviewed the licensee's design change program required by references (a) through (i) to determine if these activities were conducted in accordance with regulatory requirements, industry guides and standards, and Technical Specifications (TS).
The following criteria were used during the review to assess the overall,acceptability of the established program:
Procedures have been established to control design changes which include assurance that a proposed change does not involve an unreviewed safety question or a change in TS as required by 10 CFR 50.5 Procedures and responsibilities for design control have been established including responsibilities and methods for conducting safety evaluation Administrative controls for design document control have been established for the following:
0
4 Controlling changes to approved design change documents Controlling or recalling obsolete design change documents such as revised drawings and modification procedures Release distribution of approved design change documents Administrative controls and responsibilities have been established commensurate with the time frame for implementation to assure that design changes will be incorporated into:
0
Plant procedures Operator training programs Plant drawings to reflect implemented design changes and modifications Design controls require that implementation will be in accordance with approved procedure Design controls require assigning responsibility for identifying post-modification testing requirements and acceptance criteria in approved test procedures and for evaluation of test result Procedures assign responsibility and delineate the method for reporting design changes to the NRC in accordance with 10 CFR 50.5 Controls require review and approval of temporary modifications in accordance with Section 6 of the TS and 10 CFR 50.5 The documents listed below were reviewed to determine if these criteria had been incorporated into the licensee's design change program:
NODPS-ADM-07 NODPS-ENG-01 NODPS-ENG-02 NODPS-ENG-03 NODS-ADM-06 NODS-ENG-01 NODS-ENG-07 NODS-ENG-08 Policies, Standards, and Procedures History, Revision 0 Configuration Control, Revision 0 Engineering, Revision 0 System Engineering, Revision 0 Operations, Responsibilities, and Interfaces, Revision 0 Control of Modifications, Revision 1 Design Control Process, Revision O Design Control Document, Revision O Engineering and Construction Nuclear Design Control Manual 1.1 Design Organization, Revision 2 2.1 Nuclear Design Control Program, Revision 0 2.2 Indoctrination and Training, Revision 1 3.1 Design Inputs, Revision 3 3.2 Design Change Package (DCP) Preparation Guidelines, Revision 4 3.3 Design Verifications, Revision 0 3.4 Engineering Change Request (ECR), Revision 0 3.5 Design Document Revision, Revision 1 3.6 A/E Interface Control, Revision 1 5.1 Document Control, Revision 0
_ ____J
- NDCIM 3 Nuclear Design Control Policy and Interface Procedure, Revision 2 NDCIM 5 Nuclear Document and Drawing Control Interface Procedure, Revision 3 ADM-2 Function Bypass and Temporary Modification Control, dated 10-8-85 ADM-39 Design Charges, dated 4-17-86 STN-GN-0001 SOP 8.2.3S SOP 8.2.4S SOP 8.2.7S Instructions for DCP Preparation, dated 5-30-86 Handling of Approved Design Change Packages, Revision 0 Validation and Closeout of Design Change Packages, Revision 0 Design Change Package Supporting Documentation, Revision 0 The inspector reviewed procedures and interviewed personnel to determine the structure of the design control progra Engineering and Construction (E&C)
is designated the design authority for the license E&C is a corporate department with a percentage of personnel located on sit The site division is led by the Site Engineering Officer (SEO).
A small portion of engineering design takes place on site; a greater portion is accomplished at the corporate office The licensee also frequently employs architect/
engineers (A/E) and consultants, contracted on an ongoing or job-specific basi The hired design organizations are responsible to implement the design control program as delineated in the licensee's procedure The Nuclear Operations Department (NOD) has overall responsibility for the operational and safety elements of the design control progra NOD reviews design outputs to ensure that plant safety and operability are not adversely affecte NOD's site-based Design Control Engineer coordinates this effor The inspector reviewed the following QA audit reports to assess licensee-identified problems in this area:
S84-15 85-33 86-27 Design Control, February 19, 1985 Civil and Nuclear Engineering, February 25, 1986 Mechanical and Electrical Engineering, July 17, 1986 Audit S84-15 of site E&C identified numerous documentation problems such as reviews exceeding procedural time 1 imits, drawings showing inaccurate or deleted references, and discrepancies between drawing indexes and drawings located in working copies of DCP Documented corrective action appeared adequat Audit 85-33 of corporate engineering identified several significant program nonconformances, one of which was a failure to prepare an independent verification of specifications prepared by an A/ Audit 86-27, also a corporate inspection, identified problems in the area of indoctrination trainin All audit findings appeared to have been closed on a justifiable basi Audit 86-27 was conducted April 9-25, 1986, and not issued until July 17, 198 This exceeds the 30-day limit imposed by the T A violation is not issued on the basis of an interview with a corporate QA official who stated that this was an isolated incident and was the result of
the QA inspector responsible for writing the report being given additional field inspection dutie The inspector reviewed the most recent QA surveillance of site E&C entitled
"Surveillance of Design Changes,U February 12, 198 This surveillance of design change implementation in the field identified several significant findings including:
American Concrete Institute standards not me American Institute of Steel Construction standards not me Use of untrained fire watche Latest revisions of drawings not use ANSI N45.2.2 requirements not me Although surveillances as such are not required by the TS, the nature of these findings warrant immediate corrective actio The licensee response to the above findings was acceptable, but not aggressiv Each individual item was addressed, though a higher quality-conscience attitude could have resulted in a more complete reassessment of potential generic issues and root causes. A finding is not issued for this item; however the NRG expects aggressive corrective action whenever significant findings are identified by any mechanis The inspector randomly selected several DCPs and conducted a detailed, indepth review of the design change program to verify procedural compliance with the accepted QA progra In addition, a partial review of the technical adequacy of the DCPs was conducte All of the DCPs selected had been installed in the field and documentation was either complete or in the final stages of completio The following is a listing of the DCPs inspected during this review:
84-37 84-60 84-75 84-84 85-14 Emergency Vent Damper Position Indication, Unit 2 Emergency Diesel Generator No. 3 Radiator Cooling Louvers, Units 1 and 2 Appendix R CCW and RHR Circuit Isolation Reactor Vessel Head Shielding, Unit 2 Fuel Assembly Reconstitution, Unit 1 Each of the DCPs above appeared to contain sufficient documentation to verify that all elements of the design control program were addresse The files provided enough information to reconstruct the chronological history of each DC The inspector placed special emphasis on the engineering review and safety analysis (ER&SA) report for each packag The ER&SA addressed the following items in separate sections:
fire hazards, seismic concerns, environmental qualification (EQ), ALARA, recent NRG concerns, TS review, Final Safety Analysis Report (FSAR) review, design basis document, unreviewed safety question, and materi a 1 s of constructio This format assured that these important design considerations would not be overlooke Each ER&SA reviewed by the inspector appeared adequate in scope and depth, although the unreviewed safety question determination (USQD) tended to be unidimensional,
- brief, and unimaginativ Most USQDs consisted of one or two sentences for each part without the appearance of an in-depth examinatio Although the USQDs appeared to meet minimum acceptable standards, future NRC inspections will focus on their overall adequac Each DCP contained a detailed review of plant procedures and drawings affected by the modificatio The licensee apparently conducted plant training on the more significant procedural changes for personnel who had a need to kno This conclusion was based on documentation in the DCP and discussions with site training personne To evaluate the quality of the DCP at the time it was sent for construction in the field, the inspector performed a detailed review of all field changes issued against the original DC Among the five DCPs reviewed, the number of field changes ranged from six to fiftee Only a very small percentage of the field changes resulted from configuration control problem A much larger number resulted from problems with the modification installation procedure Another 1 arge contributing factor was drawing error Fie 1 d changes may be considered design control program performance i ndi ca tors which, in this case, point to a potential weakness in DCP procedures and drawings.
Within this area, one inspector followup item was identifie While DCP 84-37, Emergency Vent Damper Position Indication, was undergoing final Quality Control {QC) closeout, it was discovered that safety-related dampers modified for EQ considerations were not able to be verified on the Environmental Qualification Maintenance List (EQML) or the Qualification Data Report (QDR).
This discrepancy was identified on June 3, 1986 and as of August 1, 1986, remained ope Until the EQML and QDR reflect the EQ treatment of the emergency vent dampers installed per DCP 84-37, this item will be tracked as Inspector Followup Item 280, 281/86-19-01, Emergency Vent Damper EQ Documentatio Tests and Experiments (33703)
References:
(a)
10 CFR 50.54(a)(l), Conditions of Licenses (b)
VEPCO Quality Topical Report, VEP 1-5A (c)
10 CFR 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants (d)
10 CFR 50.59, Changes, Tests and Experiments (e) Technical Specification, Section 6 (f) Regulatory Guide 1.33, Quality Assurance Program Requirements {Operations)
(g)
ANSI N18.7-1976, Administrative Controls and Quality Assurance for the Operations Phase of Nuclear Power Plants
The inspector reviewed the licensee's test and experiment program required by references (a) through ( g) to determine.if the program was in conformance with regulatory requirements, industry guides and standards, and T The following criteria were used during this review to assess the overall acceptability of the established program:
A formal method was established to handle all requests or proposals for conducting plant tests involving safety related component Provisions assured that all tests will be performed in accordance with approved written procedure Responsibilities were assigned for reviewing and approving test procedure A formal system, including assignment of responsibility, was established to assure that a*ll proposed tests will be reviewed to determine whether they are as described in the FSA Responsibilities have been assigned to assure that a written safety evaluation required by 10 CFR 50.59 will be developed for each test to assure that it does not involve an unreviewed safety question or a change in T The documents listed below were reviewed to determine if the previously listed criteria had been incorporated into the licensee's test and experiments progra VEPCO QA Topical Report, VEP 1-5A ADM-35 Special Tests, 3-11-86 The inspector verified that licensee program documents ensured that, prior to conducting special tests or experiments involving safety-related systems, a safety analysis per 10 CFR 50.59 will be performed and detailed procedures will be written and followe The licensee's test and experiment program is governed by the use of special test (ST) procedure STs are controlled by the same preparation, review, and approval requirements applicable to all plant procedure The inspector reviewed documentation supporting the fol lowing ST perfor-mances:
ST-162 ST-163 ST-182 Heat Rate Test, June 27, 1984 "
Reactor Core Analysis and Testing, Monitoring the Effects of*
Partially Inserted Rod B-6 During Core Operation on Surry Unit 1, June 28, 1984 Boric Acid Transfer Pumps Operability Test, September 23, 1985 Each ST above was properly approved and presented cautions, prerequisites, and step-by-step guidance which assured that the system in question would be
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...
returned to its original conditio All procedural steps were properly signed of CFR 50.59 safety evaluations for each of the above STs were reviewe The evaluations appeared to address the safety issues required by 10 CFR 50.59 in a satisfactory and understandable manner, but lacked a broad perspective as discussed in paragraph Within this area, no violations or deviations were identified.