IR 05000498/1993021
| ML20045F544 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 06/28/1993 |
| From: | Johnson W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20045F537 | List: |
| References | |
| 50-498-93-21, 50-499-93-21, NUDOCS 9307080014 | |
| Download: ML20045F544 (6) | |
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APPENDIX B
- i U.S. NUCLEAR REGULATORY COMMISSION i
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REGION IV
i NRC Inspection Report:
50-498/93-21 Il 50-499/93-21
Operating License: NPF-76 NPF-80 Licensee:
Houston Lighting & Power Company l
P.O. Box 1700
Houston, Texas 77251
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Facility Name:
South Texas Project Electric Generating Station, l
Units 1 and 2
E Inspection At: Matagorda County, Texas Inspection Conducted: May 10 through June 10, 1993
Inspectors:
J. 1. Tapia, Senior Resident Inspector Approved:
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W. D.
n' son, Chief, Project Section A
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Inspection Summary
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Areas inspected: A special inspection was conducted to determine the
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circumstances surrounding the inappropriate dispositioning of a service
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request that had identified deficiencies in the seismic qualifications of the
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. qualified display processing system. The inspection also reviewed a-previously identified unresolved item involving incorrect breaker setpoints
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for Class lE 480 VAC magnetic. adjustable molded case circuit breakers.
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i Results:
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A violation was identified concerning a potential operability issue that
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was not recognized and promptly resolved; as a result, the appropriate
!i Technical Specification (TS) Limiting Condition for Operations was not entered.
Personnel error also contributed to this TS violation when a i
request for a conditional release was incorrectly processed (Section 1).
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A previously identified unresolved item was reviewed concerning in-the l
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adjustment of magnetic trip settings of molded case circuit breakers.
This item will remain unresolved pending further NRC review (Section 2).
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9307080014 930630
PDR. ADDCK 05000498
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Summary of Inspection Findings:
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Violation 498/9321-01 was opened (Section 1).
e Attachment:
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Persons Contacted and Exit Meeting l
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DETAILS
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1 SEISMIC QUALIFICATIONS OF THE QUALIFIED DISPLAY PROCESSING SYSTEM (QDPS)
The Diagnostic Evaluation Team (DET) reviewed the backlog of service
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requests (SRs) to assess whether outstanding SRs had the potential to impact operability of plant systems. On April 28, 1993, SR AM-188190 was selected for further review because it involved Unit 1 QDPS card cages and power supply racks being discovered with missing seismic held-in screws. This SR was written by the System Engineer on January 4,1993, in order to resolve a long-standing problem associated with maintenance personnel failing to replace hold-in screws after work activities.
Neither the System Engineer or the Senior Reactor Operator that reviewed the SR identified the condition as a potential operability issue, due to affecting the seismic qualification of the QDPS cabinet, and the QDPS was not declared inoperable.
A conditional release
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authorization, which formally requires an operability evaluation to be performed by the Technical Support Engineering Group, was requested in the SR to obtain engineering concurrence that the missing screws did not affect the seismic qualification of the QDPS. The SR was subsequently delivered to the Maintenance Planning Department and was inadvertently filed rather than being forwarded to the Technical Support Engineering Group for completion of the conditional release authorization. On April 28, 1993, the DET discovered that the conditional release authorization for the QDPS had not been obtained.
On April 29, 1993, the SR was given to the Design Engineering Department to
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perform a conditional release on the effect of the missing hold-in screws.
The evaluation was performed using walkdown notes from the System Engineer
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that identified specific dates and locations for missing screws. This data was then provided to the vendor (Westinghouse) for seismic analysis and impact assessment.
The vendor's evaluation determined that the QDPS equipment could
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not be conditionally released and that the affected components were technically inoperable. The vendor's evaluation identified the specific equipment that was affected and considered inoperable. This equipment i
included all four steam generator power operated relief valves (PORVs) and both channels of the reactor coolant system (RCS) subcooling margin monitor.
The licensee initiated rework action on the missing hold-in screws on the QDPS cabinet.
The missing screws were replaced and the QDPS was declared operable
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on April 29, 1993.
The inspector determined that the condition concerning the seismic qualification of the QDPS was identified on January 4,1993, when the system engineer identified the missing hold-in screws and generated the SR.
However, the licensee f ailed to take prompt and effective actions to ensure that this identified deficiency was corrected or that an operability evaluation was performed until April 28, 1993, after the SR was reviewed by a member of the DET. This was a violation of the requirements of 10 CFR 50, Appendix B, Criterion XVI (498/9321-01).
In addition, as a result of the inadequate action taken to disposition the SR and conduct an operability evaluation concerning the seismic qualifications of the QDPS, the steam generator PORVs
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and the RCS subcooling margin monitor should have been declared inoperable -
from January 4, 1993, until April 29, 1993.
TS 3.7.1.6 requires that in
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Modes 1-4 (in Mode 4 when the steam generators are being used for decay heat
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removal), at least four steam generator PORV's be operable.
In addition,
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TS 3.3.3.6 requires that in Modes 1-3 the RCS subcooling margin monitor be
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Between January 4, 1993, and February 11, 1993, the reactor operated in modes 1-3 with both the steam generator PORVs and the RCS subcooling margin monitor inoperable, in potential violation of the requirements of TS.
i 2 INCORRECT BREAKER SETPOINTS
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On March 23, 1993, licensee personnel identified an issue concerning the setpoints for Class lE 480 VAC magnetic adjustable molded case circuit
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breakers being set incorrectly.
This concern was based on what appeared to be
a misapplication of the instructions contained in the Electrical Setpoint-
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Index for establishing the setpoints of reactor containment penetration
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breakers.
Engineering personnel identified this issue after discussions with (
electrical maintenance personnel that were responsible for performing calibrations on a new molded case circuit breaker. This new breaker supplied r
power for the motor-operated containment supplemental purge exhaust inside-
containment isolation valve.
The methodology for adjusting the magnetic trip settings provided in the
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Electrical Setpoint Index consisted of two steps.
The first step was the
determination of the setting range, which was a calculated value resulting
from multiplying the running current by a value of _from 9 to 14, depending on the type of breaker. This calculated value does not include a tolerance
value.
The second step involved utilizing the vendor manual for the selection of a setting mark on the breaker. This step, the selection of a mark which
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corresponds to the desired setting range, included a tolerance value which was i
supplied by the vendor. The issue identified by licensee engineering
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personnel was the inclusion by maintenance personnel of the tolerance value in
the first step of the trip setting process.
This misapplication of the
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tolerance value was determined to have the potential of reducing circuit
overcurrent protection, i
A walkdown inspection of all Class lE 480 VAC motor control center breakers e
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(562 in number) was subsequently performed to determine the extent of the l
probl em.
The Unit I walkdown was completed on March 30 and Unit 2 on i
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March 31, 1993.
The licensee identified 7 Class IE breakers in Unit I with i
setpoints below the required setting, 4 Class 1E breakers above the required t
setting, and 3 Class lE nonpenetration breakers above the required setting.
l Of these 14 breakers, 5 of the breakers below the required setting were
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considered to be inoperable. Unit 2 contained 7 Class IE breakers below the required setting,1 Class lE penetration breaker above the required setting, and 1 Class 1E nonpenetration breaker above the required setting. Of these 9 breakers, only the 7 breakers below the required setting were considered to be inoperable.
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All of the inoperable breakers were associated with motor-operated valves (MOV) in various systems.
Because the supply breakers were considered to be inoperable, the associated MOVs that were powered from these breakers were also inoperable.
The licensee conducted a review to determine when the setpoint changes last
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occurred for each affected breaker in an attempt to envelope the period that
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these valves were inoperable. The results of the review were indeterminate; the licensee was not able to identify when the last setpoint change had been made to the affected breakers and, therefore, the valves were considered inoperable since the units were licensed. The inspector determined that multiple failures to comply with the requirement of TSs had or curred.
In Unit 1, the affected inoperable valves and the associated TSs were:
Valve JS Supplemental Containment Purge Supply Isolation (MOV0003)
3.6.1.7 Auxiliary FW to Steam Generator 1B Isolation (MOV0065)
3.7 1.2, 3.6.3 Containment Normal Purge Isolation (MOV0083)
3.6.1.7 Fire Protection Containment Isolation (MOV0756)
3.6.3 Accumulator 1C Discharge Isolation (M0V0039C)
3.5.1 In Unit 2, the affected inoperable valves and the associated TSs were:
Valve JS Residual Heat Removal Inlet Isolation (MOV0060A)
3.4.1.4.1, 3.4.1.4.2 3.4.1.3, 3.5.6 RCFC Chilled Water (MOV0070)
3.6.3
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RCFC Chilled Water (MOV0137)
3.6.3 i
Accumulator 2B Discharge Isolation (H0V0039B)
3.5.1 CCW RCDT Heat Exchanger Isolation (MOV0392)
3.7.3 Fire Protection Containment Isolation (MOV0756)
3.6.3 CCW Containment Isolation (MOV0404)
3.6.3, 3.7.3 At the conclusion of this inspection, the licensee was conducting an l
engineering evaluation to determine if the magnetic trip settings for the i
affected circuit breaker were excessively conservative and that these breakers in their as-found condition would have performed their function of supplying
power to their loads and maintaining the safety-related MOVs operable.
This issue will remain unresolved, pending completion of these engineering
evaluations and further review by the NRC.
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ATTACHMENT 1 PERSONS CONTACTED 1.1 Licensee Personnel H. Bergendahl, Manager, Technical Services J. Blevins, Supervisor, Procedure Control J. Calloway, Consultant, Participant Services M. Chakravorty, Executive Director, Nuclear Safety Review Board
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K. Christian, Manager, Plant Operations W. Cottle, Group Vice President M. Coughlin, Senior Licensing Engineer J. Groth, Vice President, Nuclear Generation A. Harrison, Supervising Engineer, Nuclear Licensing H. Hesidence, Independent Safety Engineer Group Acting Director T. Jordan, General Manager, Nuclear Engineering W. Jump, General Manager, Nuclear Licensing M. Kanavos, Manager, Mechanical Nuclear Engineering W. Kinsey, Vice President, Plant Support D. Leazar, Plant Engineering Manager M. Ludwig, Manager, Nuclear Training G. Parkey, Plant Manager P. Parrish, Senior Specialist, Licensing
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R. Rehkugler, Director, Quality Assurance S. Rosen, Vice President, Nuclear Engineering The personnel listed above attended the pre-exit and/or exit meeting.
In addition to the personnel listed above, the inspectors contacted other personnel during this inspection period.
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2 EXIT MEETING A pre-exit statusing meeting was conducted on May 25, 1993.
Because of additional findings that the DET had identified, the final exit meeting with members of the licensee's Nuclear Licensing Department was conducted on June 10, 1993, following the inspector's review. These findings were consistent with, and did not alter the information provided at, the pre-exit
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meeting on May 25, 1993.
During both the pre-exit and the exit meetings, the inspectors reviewed the scope and findings of the report. The licensee did not identify as proprietary any information provided to, or reviewed by, the inspectors.
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