IR 05000219/1993005
| ML20044H253 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 06/02/1993 |
| From: | Rogge J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20044H249 | List: |
| References | |
| 50-219-93-05, 50-219-93-5, NUDOCS 9306080114 | |
| Download: ML20044H253 (15) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
93-05 Docket No.
50-219 License No.
DPR-16 Licensee:
GPU Nuclear Corporation 1 Upper Pond Road Parsippany, New Jersey 07054 Facility Name:
Oyster Creek Nuclear Generating Station Inspection Period:
April 6,1993 - May 17,1993 Inspectors:
Dave Vito, Senior Resident Inspector Richard Barkley, Project Engineer John Nakoski, ilesident Inspector w 8, /9D Approved By:
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Ipfin Rogge, Section Chief (/
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Date
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gReactor Projects Section 4B, D Inspection Summary: This inspection report documents the safety inspections conducted during day shift and backshift hours of station activities includeg: plant operations; radiological controls; maintenance and surveillance; engineering as:d technical support; security; and safety assessment / quality verification.
Results: Overall, GPUN operated the facility in a safe manner. However, radiological.
controls problems were encountered during work in the new radwaste building fill aisle on May 7 and May 11,1993. A special reactive inspection was conducted to evaluate these events (see Inspection Report 50-219/93-07). An unresolved item was identified regarding
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the assessment of fire loading caused by anti-contamination clothing located in' remote radiological worker dress out stations.
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9306080114 930603 PDR ADOCK 05000219 PDR G
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EXECUTIVE SUMMARY Oyster Creek Nuclear Generating Station Report No. 93-05 Plant Operations Overall, the plant was operated in a safe manner. A detailed walkdown of the 125 volt DC power system found the system to be installed and functioning as designed.
Radiolocical Contm13 Radiological controls problems were encountered during cleanup efforts in the new radwaste building fill aisle on May 7 and May 11, 1993. A special reactive inspection was conducted to evaluate these events (See Inspection Report 50-219/93-07).
Maintenance / Surveillance
Maintenance and surveillance activities observed during this inspection period were well j
conducted and appropriately controlled.
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Engineerine and Technical Support A walkdown of the 125 volt DC power system found that the fire loading caused by
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anti-contamination clothing located in remote radiological worker dress out stations had not i
been accounted for. The licensee was evaluating this issue at the end of the inspection period.
Safety Assessment and Ouality Verification A review of recent quality assurance audits and audit findings found that audit objectives had been accomplished and responsiveness to audit findings was reasonable. QA appropriately escalated an older quality deficiency report to higher management attention to expedite its resolution.
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TABLE OF CONTENTS Page
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EXEC UTIVE S UM M ARY....................................... ii
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1.0 OPERATIONS (71707, 71710)...............................
I 1.1 Operations Summary I
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1.2 Engineered Safety Feature System Walkdown
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1.3 Facility Tours.....................................
2.0 RADIOLOGICAL CONTROLS (71707).........................
3.0 MAINTENANCE / SURVEILLANCE (62703, 61726).................
3.1 Cleaning of Containment Spray Heat Exchangers 1-1 and 1-2........
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3.2 Restoration of the Redundant Fire Pump.....................
3.3 EDG No. 2 Fuel Oil Transfer System Repairs.................
- 3.4 Replacement of EPR Servo Valve.........................
3.5 Surveillance Observation
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4.0 ENGINEERING AND TECHNICAL SUPPORT (71707, 40500)..........
4.1 Assessment of Oyster Creek Susceptibility to Peach Bottom Pressure Temperature Limit Problem
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i 5.0 SECURITY (71707)
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6.0 SAFETY ASSESSMENT / QUALITY VERIFICATION (90712)...........
6.1 Quality Assurance Audit Review
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6.2 Minor Maintenance Categorization........................ 10
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7.0 REVIEW OF PREVIOUSLY OPENED ITEMS (92701)
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8.0 EXIT MEETINGS (40500, 71707)
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8.1 Preliminary Inspection Findings.......................... 12 j
8.2 Attendance at Management Meetings le
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DETAILS 1.0 OPERATIONS (71707, 71710)
1.1 Operations Summary
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The plant operated at or near 100% power for the entire period, with one exception. On May 8,1993, power was reduced to 40% for a period of about one day to enable full closure testing of the main steam isolation valves. The testing was completed satisfactorily.
1.2 Engineered Safety Feature System Walkdown The inspector conducted a walkdown of the 125 volt vital DC power system and associated equipment. The inspector also reviewed the operating procedures, control room alarm response procedures, surveillance procedures, and technical specifications associated with the 125 volt DC system. The purpose of the walkdown was to independently verify the status and operability of the 125 volt vital DC system.
Walkdowns were conducted in those plant areas containing equipment related to the 125 volt
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DC system, including the control room, A and B battery room, C battery room, cable
spreading room,480 volt switchgear room,4160 volt switchgear room, and portions of the reactor building. Overall, system equipment was configured as designed and functioning properly. The control room panels appropriately reflected system status. The inspector reviewed the alarm response procedures for control room annunciator panels 8F and 9F and found them to be sufficiently detailed and easily referenced to the specific annunciator panel location. Current revisions of procedures associated with the 125 volt DC system were'in the control room. All of the procedures had been reviewed within the last two years. Area cleanliness was very good with the exception of an unsecured portable equipment hoist in the 4160 volt switchgear room that was later removed by the licensee. The following specific observations were made.
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A.
While assessing the structural support of equipment in the A and B battery room, the j
inspector noted that the backup static charger, which is shared by the A and B battery trains, was not seismically mounted to the floor to the same degree as the other B battery train equipment. The B batteries are safety-related, while the A batteries are not. The licensee noted that the static charger was not relied upon for safe shutdown in response to a seismic event but acknowledged the inspector's concern that a fall of
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the backup static charger cabinet could potentially affect the B 125 volt DC train equipment, particularly the B train distribution panel along the north wall of the A and -
B battery room. The licensee stated that the backup static charger was scheduled for evaluation during the seismic qualification walkdowns scheduled for the 15R refueling outage. The B train is normalty powered by a remotely controlled rotary motor-generator set that functions as a battery charge ~
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B.
While walking down the DC-2125 volt DC vital motor control center (MCC) on the j
75 ft elevation of the reactor building (RB), the inspector noted a large open metal j
cabinet containing several hundred pounds of anti-contamination clothing that was l
located within 10 ft of the vital MCC. The cabinet was part of a remote dress out area intended to facilitate radiological work in this part of the reactor building. The inspector asked if the fire loading of the anti-C clothing had been considered in the fire hazards analysis for the area. While the licensee felt that the evaluation had been done, they could not locate the associated documentation. During the inspection period, the licensee was performing an analysis considering the additional fire loading imparted by the contents of this anti-C clothing cabinet, as well as others on the RB 95 ft elevation and in the turbine building basement. Preliminary results indicated that existing fire protection features were adequate to mitigate the increased fire load and that the increased fire load did not impact the conclusions of the fire hazards analysis report. The licensee also agreed to respond to an inspector comment that more strategic location of the anti-C cabinets could be considered based on the proximity of safety-related or important-to-safety equipment and the availability of space. This issue is unresolved pending the licensee's completion of the fire hazards analysis evaluation and consideration of cabinet relocation. (Unresolved item 50-219/93-05-01).
C.
The inspector noted that there was no operator aid mounted on the wall beside the fire detection system and halon system alarm panels outside the A and B battery room.
The operator aids for fire alarm panels had been installed plant-wide in response to two prior incidents in 1990 and 1991, resulting from misinterpretation of fire alarm
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panel indications. In both cases, licensee personnel interpreted panel indications as instrument malfunctions when they actually represented suppression system inoperability. The operator aids were intended to clarify immediate actions in response to fire alarm panel indications. The licensee installed the operator aids for the fire panels outside the A and B battery room prior to the end of the inspection period.
D.
While assessing the status and condition of vital MCC-1B2 in the 480 volt switchgear room, the inspector noted that the MCC cabinet label was completely obstructed by a seismic mounting bracket located horizontally across the top portion of the cabinet.
The only positive indication of the cabinet number is written in pencil on the side of the MCC cabinet. The licensee stated that this would be referred to the plant labeling group for resolutio.
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E.
In Inspection Report 50-219/93-03, dated March 10, 1993, the inspector reviewed the replacement of the A and B station batteries during the 14R refueling outage. While battery operability was not questioned, the inspector review brought out inconsistencies in the technical specification (TS) values for battery cell voltage and specific gravity. An unresolved item was identified at that time pending licensee clarification of TS surveillance criteria, procedural acceptance criteria, and other system related information describing the battery operating requirements for cell voltage and specific gravity.
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The inspector concluded that the 125 volt DC system equipment was installed in accordance with its design basis, functioning properly, well kept, and free of safety hazards. Procedures.
were adequate to operate and maintain system equipment with the exception of the functional test inadequacies noted during the electrical distribution system functional inspection (EDSFI)
performed in May 1992 (NRC Inspection Report 50-219/92-80). The EDSFI found that there was not a functional test program for the main battery breakers, static battery chargers, rotary battery chargers, and inverters to assure that they were meeting inservice operating requirements. The inspector verified that licensee actions in response to the EDSFI finding were being completed as scheduled.
1.3 Facility Tours The inspector observed plant activities and conducted routine plant tours to assess equipment conditions, personnel safety hazards, procedural adherence and compliance with regulatory requirements. Tours were conducted of the following areas:
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e control room
- intake area
- cable spreading room
- reactor building
- diesel generator building
- turbine building
- new radwaste building
- vital switchgear rooms
- old radwaste building
- access control points
- transformer yard Control room activities were found to be well controlled and conducted in a professional manner. The inspector verified operator knowledge of ongoing plant activities, equipment status, and existing fire watches through random discussions.
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2.0 RADIOIDGICAL CONTROIS (71707)
During entry to and exit from the radiologically controlled area (RCA), the inspector verified that proper warning signs were posted, personnel entering were wearing proper dosimetry, personnel and materials leaving were properly monitored for radioactive contamination, and monitoring instruments were functional and in calibration. Posted extended Radiation Work Permits (RWPs) and survey status boards were reviewed to verify that they were current and accurate. The inspector observed activities in the RCA and verified that personnel were complying with the requirements of applicable RWPs and that workers were aware of the radiological conditions in the area.
Performance in the radiological controls area was generally good during the inspection period. However, problems were encountered during cleanup work in the new radwaste building fill aisle. On May 7,1993, a worker entered a portion of the fill aisle area that had not been adequately surveyed. On May 11,1993, during continuation of the new radwaste fill aisle cleanup effort, workers encountered airborne contamination levels that were higher than anticipated, leading to questions about the adequacy of the respiratory protection equipment provided. A special inspection was conducted on May 17-18, 1993, to evaluate the causes and licensee actions in response to these events (see 3nspection Report 50-219/93-07).
3.0 MAINTENANCE / SURVEILLANCE (62703, 61726)
3.1 Cleaning of Containment Spray Heat Exchangers 1-1 and 1-2 GPUN voluntarily entered a limiting condition for operation action statement (TS 3.4.C.3) to remove containment spray heat exchangers 1-1 and 1-2 from service for cleaning and replacement of sacrificial anodes in the heat exchanger inlet and outlet. The inspector observed the conduct of the job and discussed the work with the responsible maintenance supervisor. Plant maintenance personnel determined that the sacrificial anodes, while intact, were well depleted. in addition, a small buildup of mud and silt was detected on the upper tubesheet. The mud and silt was removed, the anodes replaced, the tubes checked for leaks and cleaned (as needed) and the heat exchangers closed and pressure tested before being returned to service. Operations returned the heat exchangers to service well prior to the
expiration of the TS action statement limits. The heat exchangers on the redundant train of containment spray remained in service and operable during this evolution. Overall, the job appeared to be well controlled and implemented. Based on the internal condition of the heat exchangers, the frequency of this preventive maintenance activity appeared appropriat i I
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3.2 Restoration of the Redundant Fire Pump in December 1992, the intake structure crane failed and struck the redundant fire pump i
house, damaging the fire pump piping and rendering the pump inoperable. GPUN continued the system restoration activities during this inspection period. The restoration is being conducted under mini-modification OCMM-996900-001 and is being performed in two i
phases. Phase I, which was completed, repaired sections of the pump piping and instaUed a connection to allow a fire truck to connect up to the fire water header and pressurize the header should the two main diesel fire pumps fail. Phase II, currently in progress, will restore the fire pump to its original configuration as well as add a new hydrant near the pump house and replace an air operator on a valve in the pump house.
The inspector observed portions of the work and reviewed the work package. The work package appeared to be in order, contained a limited number of field change notices and contained procedures (or procedural references) for all aspects of the work. Overall, the job was well controlled and implemented with QA attention to the quality aspects of the work.
The inspector considered the conduct of the modification in two phases to reflect an
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appropriate focus on plant safety.
3.3 EDG No. 2 Fuel Oil Transfer System Repairs On April 26,1993, the inspector observed maintenance performed to repair leaks in the fuel oil system for the No. 2 cmergenc diesel generator (EDG). The work included (1)
replacement of the seal between the day tank duplex level switch housing and the side of the day tank; (2) replacement of the fuel oil transfer pump filter 0-rings; and, (3) repair of pump casing leaks on both fuel oil transfer pumps. The work was performed as " skill of the craft."
The post-maintenance testing included an inservice leak test of the fuel oil transfer pumps as well as a diesel load test.
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The work was performed well with appropriate precautions taken for disposal of fuel oil and rags in compliance with hazardous waste requirements. After an initial problem, the inservice leak test of the fuel oil transfer pumps was performed successfully. When attempting to start the transfer pumps using the test switch in the generator cubicle, the pumps would not start. The EDG system engineer, who provided good oversight throughout the maintenance effort, then verified that the fuel oil pumps can only be started with the test
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switch when the diesel is running. Another equipment tagout was then implemented to jumper around the relay that controlled the pump start logic. The documentation, installation, and removal of the equipment tags and the jumper were done properly. The subsequent EDG load test was completed successfully. Good quality control oversight was provided during the maintenance effort.
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3.4 Replacement of EPR Servo Valve While preparing to reduce power for MSIV full closure testing on May 8,1993, a slight power perturbation occurred. During a brief malfunction of the turbine electrohydraulic pressure regulator (EPR), the turbine control valves opened from 83% to 94%, reactor pressure dropped about 10 psi, and a slight increase in reactor thermal power occurred. The EPR malfunction lasted for a few seconds and then plant conditions quickly returned to normal. A few minutes later, operators switched to the backup turbine mechanical pressure regulator (MPR) so that troubleshooting could be performed on the EPR.
The inspector observed the subsequent corrective maintenance that included cleaning and inspection of the EPR, replacement of the EPR servo valve, and replacement of the EPR final filters. The work package contained appropriate information for performane of the job, including current versions of the general troubleshooting procedure (A100 ADM-3660.41)
and the procedure for on-line service of the EPR (A100 SMI-3411.01). After removing the EPR final filter housing, the mechanics noted dirt in the outlet port of the filter head, The maintenance supervisor appropriately directed the technicians to take the filter head to the maintenance shop for cleaning to preclude the use of solvents in the radiologically controlled area. The used filter was tagged to preclude subsequent use. After EPR final filter replacement, the EPR servo valve was replaced, flushed, and the EPR was returned to service.
The GPUN technical functions turbine group systems engineer provided oversight for most of the repair effort. The engineer concluded that some type of foreign material, possibly from the filter gasket of the replaced EPR fuel filter, had caused a momentary restriction in the double acting piston EPR servo valve, causing the brief abnormal system response. The inspector concluded that the work was performed well and that good oversight was provided by the maintenance supervisor and the technical functions system engineer.
3.5 Surveillance Observation The inspector observed the performance of Surveillance Procedure 636.2.002, "Six Month Diesel Generator Inspection," Revision 22, dated March 3,1993, on emergency diesel generator #2. The inspector noted that a properly approved procedure was in use, approval was obtained and prerequisites satisfied prior to beginning the test, test instrumentation was properly calibrated and used, technical specifications were satisfied, and personnel performing the tests were qualified and knowledgeable about the test procedure. During the performance of the surveillance test, the maintenance technicians properly employed the temporary procedurc change process to correct an improper equipment reference. Step 6.7.3 of the
. pr xedure referred to a lube oil pump switch located in the engine cabinet. The technicians were aware that this switch had been relocated to the generator compartment during the 14R refueling outage but did not continue the procedure until the equipment relocation was verified by the system engineer and the temporary procedure change was properly approve.
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4.0 ENGINEERING AND TECHNICAL SUPPORT (71707,40500)
4.1 Assessment of Oyster Creek Susceptibility to Peach Bottom Pressure Temperature Limit Problem The Peach Bottom incident occurred on October 15,1992, after an automatic scram following closure of the main steam isolation valves (MSIV) Reactor vessel water level dropped to the low level setpoint and the high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) systems subsequently injected. Operators had difficulty controlling vessel level, and a subsequent scram signal was received on high pressure, shortly 'after the initial scram signal was reset. During the time frames while the scram signals were not reset, cold water was being injected through the control rod drive mechanisms to the bottom vessel head.
This cold water injection plus a lack of forced circulation for nine hours caused thermal stratification and resulted in the technical specification pressure / temperature limits being exceeded.
The inspector reviewed this issue as it related to Oyster Creek and also referred the issue to the licensee for assessment. The initial conclusion was that Oyster Creek is susceptible to a reactor vessel bottom head pressure / temperature problem such as the one that occurred at Peach Bottom. Oyster Creek does not have a HPCI or RCIC system but does employ isolation condensers (IC) to effect reactor depressurization/cooldown after an " isolation" scram. A particularly susceptible plant condition would be a post-scram evolution that disabled the use of the ICs. This plant condition actually recently occurred after the May 2,1992, isolation scram caused by a loss of offsite power due to forest fires in the
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vicinity of the plant. The recirculation pumps tripped following the scram due to the anticipated transient without scram (ATWS) recirc pump trip signal on high reactor pressure.
For about 20 minutes during the post-scram cooldown, pressure was being controlled by the electromatic relief valves (EMRV) due to the unavailability of the ICs. The ICs had been rendered inoperable due to a vessel overfeed from the feedwater system. Operators tried to use the feedwater pumps for level control but did not realize that feedwater regulating valve position had been locked in place when the loss of offsite power occurred (control air pressure had been lost). Pressure / temperature limits were not approached during this time frame.
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Reducing the susceptibility to this problem can be approached in several ways:
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Ensuring that the scram and post-scram cooldown procedures clearly address this subject. Current Oyster Creek procedures call for resetting the scram signal as soon as possible but do not provide much amplifying information. GPUN is considering adding a caution to the scram procedure alerting operators that lower head cooling occurs if the scram signal is not reset during a reactor isolation. Following an isolation scram, the scram signal cannot be reset until the MSIV closure scram signal can be bypassed, i.e., when the reactor is below 600 psig and the reactor mode switch is in the shutdown or refuel positio.
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Clarificatica/ refinement of requirements and guidance for starting recirc pumps and/or shutdown cooling pumps, depending on plant conditions, to effect needed forced circulation. As noted above, Oyster Creek does have the ATWS recirc pump trip on high reactor pressure or low-low reactor water level and cannot currently restart a recirc pump if they are tripped for this or any other reason after an isolation scram until cold shutdown is achieved. Also, a recirc pump cannot be started unless the temperature of coolant within the idle loop is within 50 degrees of overall coolant temperature. The shutdown cooling system cannot be used for forced circulation until reactor pressure is less than 150 psig and reactor coolant temperature is less than 350 degrees F on all five recirc loops.
Oyster Creek does not have direct bottom head coolant temperature indication, i.e..
off the drain to the reactor water cleanup system. Such an instrument would provide a viable means of accurately determining temperature differentials to permit recirc pump starts for forced circulation. Since this indication does not exist at Oyster Creek, the post-scram cooldown procedure (Procedure 203.4, Step 4.10) specifies that
..." If cooldown is a result of a scram accompanied by a trip of all recirc pumps, do not attempt to restart any recirc pump until the reactor has been depressurized to atmospheric pressure. This will prevent the sweeping away of subcooled water in the lower head region, which can result in unacceptable thermal stresses in the stub tube area (GE SIL 251, Control of RPV Bottom Head Temperatures)."
Oyster Creek does have reactor vessel metal temperature indications in the bottom head area. These metal temperature indications could provide control room personnel with additional information but whether or not they could be used to justify recirc pump restart above the cold shutdown condition would have to be subsequently evaluated.
3.
Analysis of how Oyster Creek could be affected by events similar to Peach Bottom and how much margin to vessel pressure / temperature limits exists for similar events.
GPUN is evaluating the effectiveness of natural circulation cooling at Oyster Creek based on past data and analyses. A recent GPUN assessment of shutdown cooling system (SDC) capability during an isolation cooldown indicated that SDC can very effectively and rapidly cool down the bottom head area. The results of the analysis showed that within an hour of SDC initiation (2 of 3 SDC pumps at full flow), a differential temperature of about 100 degrees F was established between the core outlet temperature and the Recirculation Imop E (SDC) outlet temperature. Input of water from the CRD system would contribute, to a small degree, to further cooling of
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the below core region. Operations personnel and particularly the shift technical advisors were required to become familiar with this analysis for reference in reactor isolation condition.
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The licensee stated that they are considering possible actions related to this issue and will make initial recommendations at the end of June 1993, subject to management approval. The licensee did acknowledge that recent events at Oyster Creek have particularly sensitized them to thermal stratification issues and that they are aware of the need to assure that these issues are appropriately addressed.
5.0 SECURITY (71707)
During routine tours, the inspector verified that access controls were in accordance with the Security Plan, security posts were properly manned, protected area gates were locked or guarded, and isolation zones were free of obstructions. The inspector examined vital area
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access points and verified that they were properly locked or guarded and that access control was in accordance with the Security Plan.
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6.0 SAFETY ASSESSMENT / QUALITY VERIFICATION (90712)
6.1 Quality Assurance Audit Review The inspector reviewed several recently ccmpleted quality assurance audit reports and the status of selected findings to assess accomplishment of audit objectives and the responsiveness to and results of corrective actions. The following documents were reviewed.
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Oyster Creek Audit S-OC-92-09, Plant Operations (Audit Period 8/20/92-12/31/92)
Oyster Creek Audit S-OC-92-16, Plant Training (Audit Period 11/25/92-1/29/93)
Quality Deficiency Report (QDR)91-053
QDR 93-005
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QDR 93-013
Audit S-OC-92-09 assessed overall control room performance and professionalism, shift turnover, switching and tagging, procedural compliance, organizational interface, and training. The audit noted five instances where surveillance tests were performed by
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instrumentation and control (I&C) technicians who had not completed a related on-the-job training (OJT) assignment beforehand. The auditors found that the administrative procedure for conduct of installed instrument surveillance, calibration, and maintenance did not include the OJT requirement for unsupervised work which supports accreditation of the training program. Acceptable corrective actions were taken by the I&C department.
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Audit S-OC-92-09 provided some objective assessment of operations department performance but was not as performance-based as other recent QA audits. The inspector noted one specific case in which the auditors reviewed the control room panel tag process but did not
attempt to assess operator knowledge. In November 1991, the licensee implemented a new
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system for logging and display of control room panel deficiency and information tags. The change involved the replacement of the control panel tags with uniquely identified magnetic
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circular markers. Tags which contain information necessary for the operators to respond to
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plant events remain on the control panel and are not replaced by a marker. While the auditors assessed the process and its documentation, they did not question operator knowledge of the markers currently on the control panel. The auditors acknowledged that questioning the operators on specific markers may have provided a more performance-based look at the effectiveness of the process.
Audit S-OC-92-16 assessed training program consistency with 10 CFR 55 requirements, performance of job performance measures (JPM), control of examination materials, instructor qualification, implementation of remedial training or reexamination requirements, and operator and shift technical advisor (STA) training requirements. The audit was primarily compliance-based and verified that the assessed training programs were completed as required and provided for instruction in appropriate areas to support the successful performance of assigned tasks.
QDR 91-053, dated July 16,1991,.noted the lack of procedural instructions for the operation and response to alarms on the plant computer system (PCS). The proposed corrective action was to develop an alarm response procedure, similar to that provided for the control panel annunciators, to address PCS alarms outside the control room operators general knowledge and not readily identified from other control room indications. The operations department has delayed the completion of the procedure several times, citing operations support staffing problems. QA appropriately escalated the QDR to the attention of the Director, Operations and Maintenance in August 1992. The current target date for completion of the procedure related corrective actions is May 31,1993.
QDR 93-005 noted that written verification of a welder's qualifications were not checked before being allowed to perform work. It was later discovered that the appropriate training had been completed but the documentation had not been finalized. QDR 93-013 noted that VT-2 examination of the Class 2 portion of the shutdown cooling system was not performed during the 14R outage. Appropriate corrective actions were proposed for these QDRs and are either completed or in progress.
6.2 Minor Maintenance Categorization During a plant tour for the engineered safety feature system walkdown of the 125 volt DC system (see Section 1.2), the inspector noted that the fire door separating the A and B trains of the 480 volt switchgear was unlatched and being held open by the 480 volt switchgear room ventilation air flow. After reporting this to the control room, the inspector was
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informed that equipment operators had noted the fire barrier impairment six days earlier and that an hourly fire watch had been posted as a compensatory measure. While the compensatory measure was appropriate, the inspector questioned why the defective door knob
had yet to be repaired. Control room personnel had written a job order as soon as the fire
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impairment was identified, and operations support personnel entered the job order into the computerized work management system (GMS2) as a high priority item. Maintenance i
i scheduling personnel then responded electronically via GMS2 that the repair of the door knob i
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could possibly be oerformed as " minor maintenance" under the guidance of Procedure 2000-WMS-1220.15, " Minor Maintenance," Revision 1, dated August 30,1991. Operations i
support personnel did not object to this characterization, and the repair effort was placed under minor maintenance. At this point, prioritization of the effort was controlled by maintenance scheduling. A subsequent walkdown of the defective fire door found that not only was the door knob defective, but the doorjamb was loose. Since repair of the door i
jamb may involve alteration of the structural wall to which the jamb is attached, the repair effort no longer fits the definition of minor maintenance and must be reprioritized under the
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formal work request system. The door has yet to be repaired, and the compensatory measure remains in place.
The inspector discussed this issue with operations and maintenance department personnel and reviewed the current version of Procedure 2000-WMS-1220.15, " Minor Maintenance." After reviewing the procedure, the inspector noted that the definition of minor maintenance could be broad?y interpreted and could result in the unnecessary delay of the repair of equipment important to safety, such as fire doors. The licensee acknowledged the inspectors comments and noted bat a revision to the minor maintenance procedure is currently underway that will address this issac ss well as incorporate aspects of the preventive maintenance program.
7.0 9 EVIEW OF PREVIOUSLY OPENED ITEMS (92701)
(Closed) Violation 50-219/92-08-01. This violation dealt with the improper implementation of Procedure 604.4.007, " Containment Spray and Emergency Service Water System
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(CS/ESW) 1 Pump Operability and Inservice Test," that resulted in the inadvertent spray of approximately 825 gallons of water into containment on April 20, 1992.
The inspector evaluated the licensee's post transient review and human performance critique of this event in Inspection Report 50-219/92-08. The corrective actions intended to preclude
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recurrence were noted in the licensee's response to the violation, dated July 2,1992, and
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included procedure changes to upgrade the affected surveillance procedures to the procedure writer's guide and implementation of training in the crew self-checking concept for operating crews. The inspector reviewed Procedure 607.4.0N, Revision 19, dated May 16,1993, and
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Procedure 607.4.005, Revision 15, dated May 16, 1993, the current operability and inservice test procedures for CS/ESW system 1 and 2 pumps, respectively. The procedures have been properly upgraded to the procedure writer's guide format and have distinctively separated and clarified instructions to the operator for occasions when pump inservice test data is or is not being taken. A multi-action procedure step in Procedure 607.4.004 had been a contributing factor to the April 20,1992, event.
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- 12 Crew self-checking training was provided to all operating crews between June 1992, and September 1992, during simulator training at the Nine Mile Point simulator. The concept of crew self-checking focuses on understanding the consequences of an individual action on others who may be affected by that action, particularly when several members of a crew are sequentially performing different actions to accomplish a desired result. Other points of the training included maintaining open crew communication and resisting " tunnel vision" on an individual action. Based on the completion of these actions, this violation is closed.
(Closed) Unresolved Item 50-219/92-25-02. This unresolved item dealt with the assessment of the cause of a reactor scram signal that occurred on December 11, 1992, whileeplacing the reactor mode switch in the refuel position from the shutdown position. The control room operator (CRO) had difficulty moving the mode switch into the refuel position and removing the key from the keylock switch within a prescribed two-second time delay. The scram signal occurred when the two-second time delay was exceeded. Shortly after the scram signal was received, it was reset and the reactor mode switch was kept in the refuel position. The licensee did not want to troubleshoot the mode switch for a potential equipment problem at that time since it was locked in the refuel position. The licensee's safety review group (SRG)
has since reevaluated the facts related to this event and concluded that the scram signal was caused when the CRO took longer than two seconds to turn the switch and remove the key from the keylock switch and that there was not an equipment problem. This item is closed.
8.0 EXIT MEETINGS (40500, 71707)
8.1 Preliminary Inspection Findings A verbal summary of preliminary findings was provided to the senior licensee management on May 21,1993. During the inspection, licensee management was periodically notified verbally of the preliminary findings by the resident inspector. No written inspection material was provided to the licensee during the inspection. No proprietary information is included in this report.
The inspection consisted of normal, backshift and deep backshift inspection; 30 of the direct inspection hours were performed during backshift periods, and 16 of the hours were deep backshift hours.
8.2 Attendance at Management Meetings The resident inspectors attended exit meetings for other inspections conducted as follows:
April 30,1993 Report No. 50-219/93-06 May 7,1993 Report No. 50-219/93-08 At these meetings the lead inspector discussed preliminary findings with senior GPUN management.
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