IR 05000334/1980024
| ML19343B478 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 10/28/1980 |
| From: | Caphton D, Troskoski W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19343B477 | List: |
| References | |
| 50-334-80-24, NUDOCS 8012240049 | |
| Download: ML19343B478 (6) | |
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D U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT
REGION I
Report No.
50-334/80-24 Docket No.
50-334 License No.
DPR-66 Priority Category C
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Licensee:
Duquesne Light Company 435 Sixth Avenue Pittsburgh, Pennsylvania 15219 Facility Name:
Beaver Valley Unit 1 Inspection At:
Shippingport, Pennsylvania Inspection Conducted:
August 25-29, 1980 Inspectors:
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Willian Troskoski, Reactor Inspector
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Approved by:
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D. L. Caphton, Chief, RO&NS 1
/da1(e Inspection Summary:
Inspection on August 25-29, 1980 (Report No. 50-334/80-24)
Areas Inspected:
Design Change approvals, reviews, implementation, and management controls; quality assurance and quality control interface with design changes; and, Onsite Safety Committee meeting minutes.
The inspection involved 32 inspector -
hours by one regionally - based inspector.
Results:
0# the three areas inspected, no apparent items of noncompliance or deviations were identified.
Region I Form 12 (Rev. April 1977)
80122400Y$
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DETAILS 1.
grsons Contacted J. Bacerek, Licensing
- M. Coppula, Results Coordinator B. Dawson, Schneider QA/QC
- S. Fenner, DL Co. QC Supervisor T. Lonnett, Adminstrative Assistant
- F. Lipchick, Sr. Compliance Engineer
- A. Mizia, DL Co. QA
- E. Rush, Schneider Site QA Manager
- J. Sieber, Superintendent Licensing and Compliance P. Slifkin, Coordinating Engineer
- G. Sovick, Compliance Engineer J. Starr, Coordinating Engineer
- P. Valenti, Station Engineering Group
- J. Werling, Station Superintendent
- H. Williams, Chief Engineer R. Zebowski, Technical Supervisor
- Denotes those present at the exit interview.
2.
Rreoperational and Startup Test Restart Program Status The inspector held discussions with representatives of Station Engineering, Station Licensing and Compliance, Duquesne Light Company Site Quality Assurance and Schneider Quality Assurance / Quality Control groups.
Areas l
discussed included:
Startup Testing Schedule;
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Approval and issuance of new Administrative Procedures to control
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Startup Testing; l
l Site Quality Assurance Surveillance Program; and,
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Design Control.
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3.
Preoperation and Startup Test Restart Program Implementation a.
Quality Assurance / Quality Control The inspector met with members of DL Company site Quality Assurance
group, and discussed implementation of the QA surveillance program.
Surveillance items related to category I design change packages (design changes that impact on nuclear safety) were reviewed on a sampling basis.
It was noted that items identified.by the QA program as possible deficiencies were forwarded to the QC group for resolution by qualified personnel.
Several surveillance items were selected at random and reviewed to verify that they were properly resolved.
The inspector had no further questions concerning the surveillance program at this time.
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b.
Design Changes and Modification (1) Design Change Packages in various stages of completion were reviewed to verify that:
The design change was properly reviewed and approved per
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Station Engineering Procedure 2.3, " Design Change Coordination,"
and Technical Specification 6.5.1.6.
The applicable documentation was being properly controlled
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per station administrative procedures.
The documentation included turnover checklists, open items lists, design output index, vendor documents, test specifications, engineering memoranda, station modification requests, and construction isometric drawings.
Jurisdictional controls over the design changes were being
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properly enforced.
(2) The following Design Change Packages were reviewed by the inspector:
(a) DCP-142, " Modify Control Room Annunciators for Diesel Generator Not Available."
(b) DCP-189, " Recirculation Spray and Low Head Safety Injection Pump NPSH."
(c) DCP-191, " Deletion of Reactor Trip Following Turbine Trip Below 50% Power".
(d) DCP-232, " Regenerative Heat Exchanger Relief Valve."
F (e) DCP-237, " Remove Part Length Control Rods."
(f) DCP-249, "4160 V Bus Undervoltage Relay."
(g) DCP-242, " Root Stop Valve for Main Steam and Steam Generator Feedwater System."
l (h) DCP-292, " Pressurizer Safety and Relief Valve Position Indication".
(3) References l
(a) Technical Specification 6.5.1.6,
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(b) Station Engineering Procedure 2.3, " Design Change Coordination,"
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(c) Onsite Safety Committee (OSC) M2eting Minutes:
BV-OSC - 84-76 BV-OSC - 77-77 BV-OSC - 97-77 BV-OSC - 65-79 BV-OSC - 22-80 BV-OSC - 24-80 BV-OSC - 28-80 BV-OSC - 47-80 BV-OSC - 73-80 BV-OSC - 81-80 (4) Findinas (a) The cover sheet for DCP-189, " Recirculation Spray and Low Head Safety Injection Pump NPSH", indicated through an approval signoff that the design concept had been reviewed during OSC meeting No. 47-80.
The cover sheet also designated that no technical specification changes were required.
The inspector confirmed that OSC meeting No. 47-80 reviewed DCP-189; and the committee concurred that "... the design concept did not require changes to the Safety Analysis Report or Technical Specifications nor do they result in any unreviewed safety questions." Thus, the DCP cover sheet accurately reflects the OSC safety analysis of the design change.
Review of the Technical Specification Change Request Log revealed that change request No. 1A-27, Permanent LHSI and Recirculation Spray Modifications, affects the following Appendix A Technical Specifications:
l 3.6.3.1 *LCO - Containment Quench Spray System (1A-27, l
rev. 0);
3.6.2.3 LCO - Chemical Addition System (1A-27, rev. 2);
3.6.2.3.a LCO - Chemical Addition System (1A-27, rev. 1);
3.5.5 LCO - Refueling Water Storage Tank (1A-27, rev. 0);
4.6.2.3 - Surveillance Requirements for Chemical Addition I
System (IA-27, rev. 1);
4.6.2.1 - Surveillance Requirements for Containment Quench Spray Subsystem (1A-27, rev. 3).
- Where LCO is the Limiting Condition for Operation.
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This Technical Specification change request and three subsequent revisions, do not support the OSC position that DCP-189 "... did not require changes to...the Technical Speci ficati on... "
(b) Appropriate approval signoffs on the cover sheet for DCP-191, " Deletion of Reactor Trip Following Turbine Trip Below 50% Power", indicated that the safety analysis had been completed by the OSC, and that a technical specification change would be required.
OSC meeting 22-80 performed the required review of DCP-191.
The OSC concurred that the design concept... "did not require changes to procedures as described in the SAR (safety analysis report), nor do they result in any unreviewed safety questions." No technical specification change
- request was addressed by the OSC.
Reactor trip setpoints are defined in Technical Specification 2.2.1, Reactor Trip System Instrumentation Setpoints.
The Reactor Trip following a Turbine Trip is one such setpoint.
Any design change that would delete or alter this trip setpoint, requires prior review and approval by both the NRC (as specified in 10 CFR 50.59(a)) and appropriate licensee groups (as specified in the technical specifications).
The OSC position on DCP-191 is not substantiated by Technical Specification 2.2.1.
Items (a) and (b) above, demonstrate an inadequate OSC review of proposed changes or modifications to plant systems or equipment that affect nuclear safety, as required by Technical Specification 6.5.1.6.
The licensee stated that he shared this concern, and would develop and implement a system whereby all Design Change Packages would be reviewed and any required technical specification changes obtained prior to returning any equipment or systems to service.
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Items ~(a) and (b) are unresolved pending inspector followup l
of the program implementation (80-24-01).
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(c) CFR 50.59 (b) requires... "a written safety evaluation which provides the basis for the determination that the change, test, or experiment does not involve an unreviewed safety question." The inspector questioned the licensee on the adequacy of the bases provided for several design changes.
The licensee informed the inspector that in the future, guidance provided by IE Circular 80-18, Safety Evaluations for Changes to Radioactive Waste Treatment
Systems, dated August 22, 1980, would be employed during j
the preparation of all bases used tp determine whether or
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not an unreviewed safety question is invalved in the
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design change.
This is an inspector follow up item (80-24-02).
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4.
Exit Interview The inspector swet with licensee representatives (denoted in section 1),
at the conclusion of the inspection on August 29, 1980.
The Inspector summarized the scope and findings of the inspection, and discussed the concerns denoted in Paragraph 3.
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