IR 05000369/1987019
| ML20235E397 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 07/02/1987 |
| From: | Guenther S, William Orders, Peebles T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20235E385 | List: |
| References | |
| 50-369-87-19, 50-370-87-19, NUDOCS 8707110038 | |
| Download: ML20235E397 (7) | |
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UMTED STATES NUCLEAR REGULATORY COMMISSION
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REGION 11 n
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,j 101 MARIETTA STREET, N.W.
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ATt. ANT A, GEORGI A 30323
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Licensee: Duke Power Company
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422 South Church Street Charlotte, NC 28242 Facility Name: McGuire Nuclear Station 1 and 2 Docket No(s): 50-369 and 50-370 License No(s):
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Inspection Conducted:
May 21, 1987 - June 20, 1987 i
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Inspectors:
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2[f,2 W. Orders, Seiifor Resident Ingector
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~7A dea. A 7Nr7 S.Guen'ther,RdsidentInspectf D' ate Signed j
Approved by:
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7/L[D T. A. Peebles, Se6 tion Chief p'
' Da'te Signed Division of Reactor Projects
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SUMMARY i
Scope: This routine unannounced inspection involved the areas of operations j
safety verification, surveillance testing, maintenance activitiss, follow-up of j
previous enforcement actions, refueling activities, design change activities, j
and independent inspection in the area of control room modifications.
j Results:
In the areas inspected, no violations or deviations were identified.
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8707110038 870702 PDR ADOCK 05000369 G
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REPORT DETAILS 1. Persons Contacted Licensee Employees
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- T. McConnell, Plant Manager B. Travis, Superintendent of Operations D. Rains, Superintendent of Maintenance B. Hamilton, Superintendent of Technical Services N. McCraw, Compliance Engineer M. Sample, Superintendent of Integrated Scheduling N. Atherton, Compliance G. Gilbert, Operations Engineer J. Snyder, Performance Engineer W. Reeside, Operations Engineer E. Estep, Project Engineer
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Other licensee employees contacted included construction craftsmen, technicians, operators, mechanics, security force members, and office personnel.
- Attended exit interview 2.
Exit Interview The inspection scope and findings were summarized on June 23, 1987, with those persons indicated in paragraph 1 above. The licensee did not identify as proprietary any of the information reviewed by the inspectors during the course of their inspection.
3. Unresolved Items An unresolved item (UNR) is a matter about which more information is required to determine whether it is acceptable or may involve a violation or deviation.
One unresolved item concerning a Unit I thermal power calculation is detailed in paragraph 4.
I 4. Plant Operations The inspection staff reviewed plant operations during the report period, to verify conformance with applicable regulatory requirements. Control room logs, shift supervisors' logs, shift turnover records and equipment removal and restoration records were routinely perused. Interviews were conducted with plant operations, maintenance, chemistry, health physics, and performance personnel,
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Activities within the control room were monitored during shifts and at shift changes. Actions and/or activities observed were conducted as prescribed in applicable station administrative directives. The complement of licensed personnel on each shift met or exceeded the minimum required by Technical Specifications.
Plant tours taken during the reporting period included, but were not limited to, the turbine buildings, Unit 2 reactor building, auxiliary j
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building, Units 1 and 2 electrical equipment rooms, Units 1 and 2 cable spreading rooms, and the station yard zone inside the protected area.
During the plant tours, ongoing activities, housekeeping, security,
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equipment status and radiation control practices were observed.
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Unit 1 Operations Unit 1 operated throughout the report period with no major problems.
Power was reduced briefly to about 70 percent on May 29, 1987, to allow ultrasonic testing of two elbows in the heater drain system, after significant wall thinning was detected in the equivalent Unit 2 elbows.
The licensee determined that the Unit 1 elbows were also somewhat degraded, but that the wall thickness was satisfactory for continued operation until the next refueling outage in September 1987.
As an added safety measure the licensee manufactured concentric, oversized elbow patches, which were welded over the degraded elbows during a second power reduction on June 6.
On June 11, the licensee determined that a best estimate thermal power calculation performed on May 20, 1987, did not take into account the venturi fouling correction factor from the previous
venturi fouling test.
This resulted in a
non-conservative calculation of thermal power and apparent operation above the unit's rated thermal power limit of 3411 Megawatts thermal during the period from May 20 until power was reduced on June 11.
The licensee has initiated a Problem Investigation Report (PIR Serial No.1-M87-0116)
to review this incident.
It will be tracked as an Unresolved Item (UNR 50-369/87-19-01) pending completion and review of the licensee's PIR.
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Unit 2 Refueling i
Unit 2 began the report period in Mode 6, continuing the refueling outage begun on May 1, 1987. Core off load was completed on May 22
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and reload was begun on June 2.
The core reload and verification was completed on June 8 and Mode 5 (cold shutdown) was entered on June 17.
The report period ended with the unit still in Mode 5, anticipating entry into Mode 4 (hot shutdown) on June 24 and criticality on June 29.
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Major evolutions conducted during the outage included the following:
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upper head injection removal
steam generator tube shot peening, eddy current testing, and
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plugging and secondary side sludge lancing i
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emergency diesel generator overhaul
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nuclear service water pump impeller and small-bore piping replacement
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digital electro-hydraulic (DEH) control system replacement and
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main control board modifications No violations or deviations were identified.
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Surveillance Testing Selected surveillance tests were analyzed and/or witnessed by the inspector to ascertain procedural and performance adequacy and conformance with applicable Technical Specifications.
Selected tests were witnessed to ascertain that current written approved I
procedures were available and in use, that test equipment in use was calibrated, that test prerequisites were met, that system restoration was completed and test results were adequate.
Detailed below are selected tests which were either reviewed or witnessed:
PT-0-A-4350-28A 125 Volt Vital Battery Test i
PT-2-A-4403-03 RN Valve Stroke Timing PT-2-A-4209-03 NV Valve Stroke Test PT-2-A-4200-10C ND Auto Isolation PT-2-A-4209-01A NV Pump 1A Performance Test PT-2-A-4403-07 RN Train 2A Flow Balance j
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PT-1-A-4206-01A NI Pump 1A Performance Test PT-1-A-4209-01C Standby Makeup Pump Test i
PT-1-A-4401-05B KC Heat Exchanger-Performance Test PT-1-A-4206-01B NI Pump 1B Performance Test PT-1-A-4401-01 KC Train A Performance Test PT-1-A-4252-01A CA Pump 1A Performance Test PT-0-A-4150-05 PZR Safety Valve Setpoint Test No violations or deviations were identified.
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Maintenance Observations
Routine maintenance activities were reviewed and/or witnessed by the resident inspection staff to ascertain procedural and performance adequacy
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and conformance with applicable Technical Specifications.
The selected activities witnessed were examined to ascertain that, where applicable, current written approved procedures were available and in use, that prerequisites were met, that equipment restoration was completed and mainte1ance results were adequate.
An Unresolved Item (50-369,370/87-08-02), regarding the maintenance and testing of valve INM-26 as documented in a previous Inspection Report, has been resolycd as an apparent violation of Technical Specification 6.8.1.
This issue is discussed in detail later in this report.
No violations or deviations were identified.
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7.
Unit 2 Refueling (60705, 60710)
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Unit 2 commenced a refueling outage on May 1, 1987.
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reviewed a number of licensee procedures for the conduct of refueling operations to ascertain their adequacy. The procedures reviewed included, but were not limited to:
PT-2-A-4550-022 - Total Core Unloading PT-0-A-4550-024 - Fuel Assembly Examination OP-0-A-6550-11 - Internal Transfer of Fuel Assemblies / Inserts PT-2-A-4550-07 - Total Core Reloading During tte performance of PT-0-A-4550-24, Fuel Assembly Examination, evidence os baffle jet impingement damage was detected on one assembly.
Three fuel pins were loose in their grid straps, and one of the three had slid down to contact the bottom nozzle.
There was no evidence of gross cladding failure.
The assembly, which was to be reloaded, was removed from service. Also, during this outage the peripheral assemblies, which
are subject to baffle jetting, were modified by adding stabilizer clips to the outside faces.
The inconel clips function to alter the resonant l
frequency of the fuel pins and in so doing mitigate the consequences of baffle jet flow.
Refueling activities were monitored to ascertain Technical Specification compliance and procedural adherence.
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1 8.
Inadequate Post Maintenance Procedure.
As was reported in reports 369/87-08 and 369/87-14, on February 24, 1987, valve INM-26B, the Reactor Coolant (NC) Hot Leg Sample Header Outside Containment Isolation valve failed.to operate correctly during a slave relay test.
During a portion of the test, valve INM-26B should have closed and remained closed.
Instead, the valve cycled continuously, as long as the output relay was energized from the slave relay test device. With the output relay deenergized (normal), valve INM-26B operated correctly from the control board switch..In short, the valve would not have isolated on
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i an automatic engineered safety feature (ESF) signal.
Instrument and Electrical (IAE) personnel investigated the ' problem on February 25, 1987. When the-IAE technician removed the cover from the
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terminal block, he detected a short between terminals 24 and 15.
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short was cleared and the valve retested satisfactorily from the control
board switch.
In report 369/87-08, the issue of post maintenance testing was left unresolved due to ongoing investigations being performed by the licensee as well as the resident staff. The licensee subsequently concluded (PIR I
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M87-055) that testing of the automatic function of valves subsequent to replacement or major maintenance is not necessary and would not be performed. This conclusion was documented in inspection report 369/87-14.
The issue was maintained unresolved in that report due to the inspector's continuing concern over, and research of, the issue.
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On May 28, 1987, it was learned that on May 5, during the performance of a
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Unit 2 ESF actuation test, that valve 2KC3A, which is designed to isolate on an ESF signal, oscillated between open and closed as had INM26B.
j Subsequent inspection revealed an identical short. circuit in 2KC3A as had i
occurred in 1NM-26B.
I An additional review of the investigation of the INM26B event confirmed that the proposed corrective actions deal with precautions to prevent short circuits, but do not address a post maintenance test to detect inoperable automatic function.
Research also confirmed the following requirements pertinent to post maintenance testing.
10 CFR 50, Appendix B, Criterion XI, Test Control, states that a test program shall be established to assure that all testing required to demonstrate that systems and ' components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance units
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contained in applicable design documents.
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6 Section 17.2.11-(Test Control) of the Duke Power Company' Quality Assurance
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Program Topical Report which _ implements 10 CFR 50, Appendix B, i
Criterion. XI at the Duke sites, states that electrical tests, operational tests, proof. tests or other special tests are performed as required to verify the satisfactory performance of nuclear safety-related systems and components after maintenance.
The key words here are verify performance after maintenance.
Further, Section 17.2.14 (Inspection Test and Operating Status) of the Duke Power Company: Quality Assurance Program Topical, Report which
- implements Criterion XIV, states that the operability of an. item removed from operation for maintenance or testing be verified prior to returning the item to normal service. ANSI N18.7-1976/ANS-3.2, as committed to in the Duke Power Company Topical Report, Duke-1, in section 5.2.6 states that when equipment is ready to be returned to service, the equipment is to be placed in operation and verified to be functional.
Clearly, the intent of the foregoing is to ascertain that equipment placed-back in service after maintenance is fully operable.
Contrary to the above, testing required to demonstrate, that systems / components will perform satisfactorily-in service.was not performed in that both valves were returned to service, yet were inoperable.
These events, singularly and collectively, constitute a violation of the above requirements.
However, in a review of the most recent (March 30, 1987) Institute of Nuclear Power Operations (INPO)
evaluation results, it was determined that post maintenance testing was also on INPO finding (MA.3-1). In that vein, and in acknowledgment of the agreement between INPO and the NRC, a Notice of Violation will not be issued at this time.
Instead, the NRC will monitor Duke's response to the INPO finding to assure regulatory compliance is achieved.
Therefore, this item will remain unresolved, pending completion of Duke's j
corrective actions in response to the INPO finding, j
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