IR 05000483/1993003
| ML20035H176 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 04/27/1993 |
| From: | Jackiw I NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20035H170 | List: |
| References | |
| 50-483-93-03, 50-483-93-3, NUDOCS 9305030248 | |
| Download: ML20035H176 (16) | |
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U. S. NUCLEAR REGULATORY COMMISSION
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REGION III
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Report No. 50-483/93003(DRP)
~i Docket No.
50-483 License No. NPF-30
Licensee: Union Electric Company
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Post Office Box 149 - Mail Code 400
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St. Louis, MO 63166 Facility Name: Callaway Plant, Unit 1
i Inspection at: Callaway Site, Steedman, M0
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Inspection Conducted:
February 1 through April 8, 1993
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Inspectors:
B. L. Bartlett D. R. Calhoun J. A. Gavula j
L. R. Wharton l
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b 8.7 ~[8 Approved By:
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Reactor Pro' cts, Section 3A Date i
Inspection Summarv 1,
Inspection from February 1 throuah April 8.1993 (Report No. 50-483/93003(DRP)1 i
Areas inspected:
Routine unannounced safety inspections of engineered safety
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features (ESF) system walkdown, plant operations, maintenance and surveillance, inspection of licensee event reports, and follow up on previous NRC inspection
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findings was conducted.
A special inspection of a containment-cooling system water hammer transient was also conducted.
Results:
Of the areas inspected, one violation-was identified.
A summary
follows.
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Operations
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The operating crews continued to perform routine' activities and evolutions in a l
controlled and organized manner. Briefings were held for a number of maintenance
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activities to ensure all personnel involved were abreast of planned activities, I
aware of actions to take if problems arose, and informed to maintain open lines
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t of communication with the operations' staff.
A reactor operator was performing a number of surveillances when a water hammer
. transient was experienced on the 'C' containment cooler.
Failure of previous-corrective action to prevent the water hammer recurrence is a violation.
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9305030248 930427 I
PDR ADOCK 05000483 O
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Radiolooical Controls j
Radiological. conditions were maintained at an acceptable level.
The I
implementation of several licensee initiatives has resulted in reduced
contamination areas and has increased emphasis to repair leaks. The number of housekeeping deficiencies identified by the resident staff has decreased since the last reporting period.
Maintenance / Surveillance I
Routine maintenance and surveillance activities were effectively performed by knowledgeable electricians, mechanics and instrumentation and control (I&C)
technicians. Supervisory support and oversight were evident during craftsmen's
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performance of work activities. Also, documentation of completed work activities has improved.
- Enoineerino and Technical Sucoort There was active involvement by engineering personnel during a number of maintenance and surveillance tasks.
j Ongoing engineering support was provided while installing the monitoring cards
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for the logic power supplies. Also, good support was evident in addressing the turbine driven auxiliary feedwater (TDAFW) pump bearing wear issue.
Safety Assessment and Quality Verification
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Plant management continued to have an active presence in the resolution of daily
'l operational issues such as the TDAFW pump bearing wear and the water hammer
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events.
Management was aggressive in addressing the water hammer event that
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occurred in the containment cooling system.
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DETAILS 1.
Persons Contacted l
D. F. Schnell, Senior Vice President, Nuclear
- G. L. Randolph, Vice President, Nuclear Operations
- W. R. Campbell, Manager, Callaway Plant
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- C. D. Naslund, Manager, Nuclear Engineering
- J. V. Laux, Manager, Quality Assurance
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J. D. Blosser, Manager, Operations Support
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- M. E. Taylor, Assistant Manager, Work Control
D. E. Young, Superintendent, Operations M. S. Evans, Superintendent, Health Physics S. S. Sampson, Supervising Engineer, Site Licensing
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G. J. Czeschin, Superintendent, Planning and Scheduling
G. R. Pendegraff, Superintendent, Security
- C. E. Slizewski, Supervisor, Quality Assurance Program G. A. Hughes, Supervisor, Independent Safety Engineer Group i
- C. S. Petzel, Quality Assurance Engineer
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J. A. McGraw, Superintendent, System Engineering
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R. D. Affolter, Superintendent, Design Control r
- Denotes those present at one or more exit interviews.
In addition, a number of equipment operators, reactor operators,. senior reactor operators, and other members of the quality control, operations,
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maintenance, health physics, and engineering staffs were contacted.
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2.
Followup on Previous NRC Inspection Findinas (92701)
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(CLOSED 1 VIOLATION 483/92015-02: Failure to have locking device installed on normally closed valve BG V-0027.
While touring the auxiliary building the NRC inspectors identified that valve BG V-0027 in the chemical and volume control system was not in its
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required, locked closed, position.
i The licensee was notified and a locking device was immediately installed
on the valve.
Subsequently,. the licensee performed a walkdown of accessible components listed in the locked component control procedure (LCCP).
Appropriate actions were taken for all identified walkdown i
discrepancies.
. In - addition, the licensee discussed ' expectations. and procedural requirements for proper restoration with pertinent personnel.
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Also, the licensee will evaluate a modification to' the workman's
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protection assurance computer program which would identify the required restoration position for those valves in the LCCP.
The NRC inspectors verified that locked component discrepancies identified
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during the audit' had been properly addressed.
i This violation is closed.
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Enoineered Safety Features (ESF) System Walkdown (71710)
The inspectors performed a walkdown of the safety-related portions of the high head safety injection system (licensee system designator BG).
The purpose of the walkdown was to independently verify that a selected ESF system was operable as required by plant technical specifications (TS).
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The inspectors verified the positions of the BG valves in the plant (by direct observation), the positions of the associated electrical breakers in the various switchgear rooms, and that the associated switches on control panels were in their proper positions. Components, breakers, and compartments were found to be clean and appropriately maintained (no valve packing leaks, no bent valve stems or missing handwheels).
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Comments generated from the plant walkdown, comparison of procedures and prints, review of TS, review of engineering calculations, and review of the updated safety analysis report (USAR) are discussed below:
The centrifugal charging pumps (CCPs)
were clean and well maintained.
The "A" CCP had recently been decontaminated so that personnel access to the motor no longer required protective clothing.
However, there were pieces of duct tape attached to piping and other minor pieces of trash laying about the rooms.
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e The handwheel locking nut for normally locked valve, BG V096, was missing. This condition nullified the purpose of the locking device
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on the valve. The licensee immediately installed a new locking nut on the valve.
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Conclusions l
I The high head safety injection system was maintained at an acceptable level. While the housekeeping and cleanliness conditions in both the 'A'
and 'B' pump rooms had improved, a number of oil leaks were observed.
No violations or deviations were identified.
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Plant Operations (71707)
l The objectives of this inspection were to ensure that the facility was
being operated - safely and in conformance with license and regulatory requirements and that the licensee's management control systems were effectively discharging the licensee's responsibilities for continued safe
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operation.
The methods used to perform this inspection included direct
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observation of activities and equipment, tours of the facility, interviews and discussions with licensee personnel, independent verification of safety system status and limiting conditions for operation (LCOs),
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Areas reviewed during this inspection included, but were not limited to, control room activities, routine surveillances, engineered safety feature operability, radiation protection controls, fire protection, security,
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I plant cleanliness, instrumentation and alarms, deficiency reports, and corrective actions.
Conclusion i
The control room staff adequately performed routine operational tasks. No violations or deviations were identified.
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Containment Coolino System Water Hammer Event On March 23, 1993, operators were performing surveillance procedure OSP-l EF-P001A, "Section XI Essential Service Water Train A Operability," which
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demonstrates the operational readiness of the
"A" ESW pump.
They were also performing another surveillance procedure on the ESW system at the same time.
Because of this complication, the "A" ESW pump had to be stopped and restarted to verify the stroke time of the pump's automatic l
vent valve. Although not referenced in the surveillance procedures being
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performed, operators were aware of the precaution given in step 2.14 of l
normal operating procedure OTN-EF-00001, " Essential Service Water System."
The precaution stated, in part, that it was necessary to isolate the containment coolers prior to reestablishing flow to prevent water hammer due to draining of the ESW return lines.
On that basis, the operators
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closed the containment cooler return valves EF-HV-0047 and EF-HV-0049.
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l The "A" ESW pump was stopped and restarted approximately 30 seconds later.
l Immediately after restarting the pump, a loud noise was heard in the control building and a seismic alarm was activated indicating that a possible water hammer had occurred.
The licensee's investigation into the extent of the water hammer event revealed very limited damage. A pin hole leak in one of the containment I
cooler coils was noted.
However, direct correlation to the event was I
difficult because there was no indication of any local deformation.
A bent indicating needle on a local pressure gauge was also found.
There were no other indications of damage and walkdowns found evidence that the piping had not moved during the event.
The licensee had previously evaluated the ESW system for prior water hammer events, and found that the piping was still within Level D stress limits even with several snubbers considered inoperable.
While this confirmed that the system was operable, it did not justify routine
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operation of the system with water hammer loads.
Because all of the
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snubbers were found to be operable and there was a general lack of damage, the licensee concluded that the current event was of less magnitude than the previous events and that the piping stresses would-have been within the allowable Level D stress limits. The NRC inspector concurred with the licensee's assessment of the event.
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Previous water hammer events in the ESW system have been documented in NRC Inspection Reports 50-483/92009 and 50-483/92016. The corrective actions from these previous events included a revision to the normal operating procedure that added caution statement 2.14 regarding the prevention of water hammer by isolating the containment coolers.
-Although the containment coolers were successfully filled and ' vented during normal i
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operations, the previous corrective actions were inadequate to prevent recurrence of water hammer during the latest performance of surveillances on the system.
The adequacy of the original caution statement is questionable in that it implies that only the return lines on the containment coolers need to be closed if flow is stopped.
To prevent column separation from occurring both the supply and return lines must be isolated.
In addition although the caution statement appears in the normal operating procedure, the surveillance procedures neither contain similar caution statements nor refer to the cautions statements in the
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normal operating procedure.
The failure to take adequate corrective
actions to preclude repetition of a significant condition adverse to quality is considered a violation of 10 CFR 50, Appendix B, Criterion XVI
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(483/93003-01).
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As a result of this event, the licensee issued further revisions to the normal operating and surveillance procedures with explicit directions on isolating both the inlet and outlet valves for the containment coolers.
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i The long term corrective action is to modify the pipe supports on the ESW
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system such that the anticipated water hammer loads will result in i
stresses less than allowable emergency stress levels. The acceptability l
of this approach has not been reviewed by appropriate NRC personnel.
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Conclusions j
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The recurrence of a water hammer transient, due to the implementatic. of
l inadequate corrective action for previous water hammer events, is a
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violation (50-483/93003-01).
One violation was identified.
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6.
Maintenance / Surveillance (62703) (61726)
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Selected portions of the plant surveillance, test, and maintenance
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activities on safety-related systems and components were observed or reviewed to ascertain that the activities were performed in accordance
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with approved procedures, regulatory guides, industry codes and standards,
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and the Technical Specifications.
The following items were considered during these inspections: the limiting conditions for operation were met
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while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using
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approved procedures and were inspected as applicable; functional testing and/or calibration was performed prior to returning the components or systems to service; parts and materials that were used were properly
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l certified; and appropriate fire prevention, radiological, and housekeeping conditions were maintained.
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Maintenance The reviewed maintenance activities included:
Vork Recuest No.
Activity W528526 Inspect and rework the lugs to valve AL HV-
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P516262 Inspect and service the limitorque operator
to valve AL HV-0030.
l P501569 Inspection and service of 'B' MDAFW pump.
W153098 Adjust position indicator for valve AL HV-0010.
W153099 Adjust positioner and I/P for valve AL HV-
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C527697 Installed monitoring cards for the logic power supplies.
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P529656 Annunciator logic power supply
functionality check.
W153049 Installed new relay in battery charger NK23.
W152187 Leak test of containment personnel _ hatch drive shaft seals.
C500721 Removed drain plugs and installed nipples
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on the 'B' MDAFW pump.
W528666 Replaced pinched wire for safety injection pump suction to refueling water storage tank isolation valve.
W514163 Inspection of safety injection pump 'B'
discharge to cold leg injection isolation valve.
Installation of Monitorina Cards for Loaic Power Supplies
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On March 10, 1993, the licensee was implementing modification, i
C527697, which installed monitoring cards (MCs) for the annunciator system's 14 logic power supplies. To maintain continuity, minimize the potential for any adverse impact, and better control the work activity, a decision was made to perform this work only on the day
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shift.
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After all MCs had been mounted, the I&C technicians began hard i
wiring the MCs on March 11, 1993. As part of this modification, a work request, to remove wires which connected all logic power i
supplies together, was to be performed. These wires (two) had been identified during a scheme check of the annunciator system, in response to the loss of annunciators event of October 17, 1992.
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This event was documented in inspection reports 50-483/92018 and 92020.
Licensee management decided to delay the removal of the wires until the modification was implemented.
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Both leads were disconnected prior to hard wiring the MCs.
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action was taken to minimize any adverse impact, stemming from
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personnel errors, on the operability of the annunciator system.
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One wire connected power between cabinets RK045El and E2; power
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between cabinets RK045E2 and E3 was connected by the other wire.
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Since the wiring configuration tied all the logic power supplies i
together this inherently increased the probability of losing reflash capability of the annunciator system.
By first removing these
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wires, if a short had occurred while working in cabinet RK045El, the i
number of annunciators lost would had been limited to that
particular cabinet (six logic power supplies).
The other two
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cabinets would not have been affected and would have maintained
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annunciator reflash operability.
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The lead connecting cabinets RK045E2 and E3 provided a parallel
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voltage source, because it had been incorrectly landed, to the
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J driver and reflash cards that are normally fed from logic power i
supply number four (LPS4).
When this second voltage source was i
removed, the LPS4 should have been providing its own output voltage
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J (-125 Vdc).
However, at some unknown time, the output fuse blew i
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Therefore, upon removing the second voltage source, the driver and reflash cards became inoperable; the reflash capability of 12 annunciators had been lost.
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operators during the loss the reflash capability of these non-safety-related annunciators.
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The nonfunctionality of the 12 annunciators was not determined when the wire was disconnected because the MC for LPS4, in cabinet
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RK045E3, had not been hardwired yet. By the end of the day shift, i
the Instrument and Control (I&C) technicians had only hardwired MCs for LPS#1, 2, and 3, in cabinet RK045El. Three MCs remained to be-hardwired in cabinet RK045El while cabinets RK045E2 and RK045E3 each
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had four MCs to be hardwired.
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As each MC was connected to each individual LPS, the modification
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only directed retesting those annunciators associated with that
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Therefore, the impact on system functionality of
the WR, which was implemented along with the modification, was left
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unchecked since a MC had not yet been connected to LPS4 in cabinet RK045E3.
The licensee's retest did not identify the failed logic power supply. If the entire annunciator system had been tested at the end of the day shift, this failure would have been discovered. However, i
retesting the whole system was not deemed necessary because the weekly annunciator test had verified system functionality.
In hindsight, the licensee felt the more prudent action to have taken would have been to test the entire system.
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t The failure of the LPS was detected at approximately 9:25 p.m. by I&C technicians troubleshooting an unrelated issue. Because it was
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unknown when the output fuse blew, the licensee could not determine how long the LPS was not functioning.
Upon notification, the control room staff entered off normal operating procedure OTO-RVs-0001, " Loss of Control Room Alarms." The system engineer reported
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to the site to assist in troubleshooting activities. A retest was satisfactorily performed after the failed output fuse for LPS4 was
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replaced.
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The licensee again verified system functionality, by performing additional self-checking actions which included verifying that the high output voltage (-125 Vdc) was present on all logic power l
supplies that had not been connected to MCs prior to continuing with the modification work on March 12, 1993.
Containment Personnel Hatch Local Leak Rate Test On February 18, 1993, the licensee performed barrel local leak rate test (LLRT), S523375, on the containment personnel hatch (CPH). The leakage rate was determined to be approximately 5,000 standard cubic centimeters per minute (SCCM) which met the acceptance limit of 21,000 SCCM.
Even though the surveillance was successful, the licensee performed W152187 which involved leak testing of the interior and exterior
handwheel shaft seals (0-rings).
The decision to perform this testing was based on an alternating upward leakage trend; the last three LLRT surveillance results were 700, 19,000, and 5,500 SCCM,
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respectively. In anticipation of S523375 being unsatisfactory, the
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licensee prepared work request packages to troubleshoot the potential causes which included the 0-ring shaft seals and the
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equalizing valves.
The test consisted of pressurizing the area between the two rings to-48 psig. This was done for each upper and lower shaft. Of the four i
seals that were tested, the lower interior shaft seal failed at
1,545 SCCM indicating the inner seal had failed. The upper exterior
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shaft seal also failed at 1,500 SCCM indicating failure of the outer seal.
The failures of these 0-rings aided in identifying those
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components which were contributors in the barrel LLRT on the CPH.
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The remaining leakage was believed to have come from the equalizing valves and the door seals.
The issue was handled in a proactive and organized manner.
A briefing was held prior to commencing work activities. The testing was conducted with engineering personnel involvement. The control i
room shift supervisor was intmediately informed of test results to ensure timely entry of Limiting Condition for Operation Action Statement if necessary. Due to extensive disassembly required for repair, work request packages for the seals and the equalizing valves were pre-planned to minimize the potential of exceeding the LC0 action statement time limit. The licensee modified these work requests to replace the defective 0-rings during future maintenance activities.
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Surveillance l
l The reviewed surveillances included:
Procedure No.
Activity OSP-BG-00001 Boron injection flow paths verification.
OSP-SA-0017B Train 'B' safety injection and containment spray slave relay test.
OSP-Mi-00002 Periodic test of the
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stand-by diesel
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generator.
ISF-AE-OL518 Functional test of steam generator
'A'
narrow range level.
ISF-AE-0L528 Functional test of steam generator
'B'
narrow range level.
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ISF-AE-OL538 Functional test of steam generator
'C'
narrow range level.
I ISF-AE-OL548 Functional test of steam generator
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narrow range level.
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i ISF-GK-00R05 Functional test of control building supply air in radiation detection.
OSP-SA-0017B Train 'B' SIS /CSAS slave relay test.
OSP-KE-00003 Cask handling and spent fuel pool bridge crane excessive local interlock verification test.
OSP-HB-B0001 Section XI radwaste valve operability test.
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ISF-AE-OL517 Functional test of steam generator
'A'
narrow range level.
ISF-AE-OL527 Functional test of steam generator
'B'
narrow range level.
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ISF-AE-0L537 Functional test of steam generator
'C'
narrow range level.
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ISF-AE-OL547 Functional test of steam generator
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narrow range level.
ISF-SE-00N31 Functional test of nuclear instrumentation source range channel N31.
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MSE-ZZ-QS005 Functional test of 480V molded case circuit breaker.
OSP-SM-LLOL1 Containment personnel hatch door seal leak
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rate test.
Position Indication Discreoancies Durina Valve Strokes
On February 4, 1993 during the performance of OSP-AL-V001C, the licensee identified discrepancies between the light indications in
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the control room (CR) and the local valve position indicators (VPI)
on auxiliary feedwater system air operated valves AL HV-0010, and HV-0012.
Valves AL HV-0010 and HV-0012 were not indicating fully closed, by
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local VPI, but the closed light indications were received in the
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control room.
Work requests were written to repair the valves.
Valve AL HV-0010 was thought to be an indication problem only; and valve AL HV-0012 appeared not to fully stroke.
During troubleshooting and repair activities on AL HV-0012, the I&C technicians determined that the valve was stroking properly.
The technicians identified that the valve p citioner and the current to pressure controller were slightly out of adjustment.
The technicians made the necessary adjustments to both instruments l
resulting in the local VPI matching control room light indication.
As a retest, OSP-Al-V001C was performed satisfactorily.'
Due to valve AL HV-0010 not indicating full open upon actuation, the technicians made adjustments to the VPI stroke scale as well as the
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positioner. After making the appropriate realignments, engineering management and a maintenance engineer determined that valve AL HV-0010 was not traveling its full one and one-half inch stroke. The
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valve was only stroking one and three-eighths inches.
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shortened stroke distance was due to the adjustment of the actuator coupling rod (ACR), in that, the ACR was hitting against the
actuator stops prior to the valve stroking one and one-half inches.
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The licensee determined the shortened stroking distance would still
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provide adequate flow to the steam generator.
Thus, this as-left
stroking condition was considered acceptable.
"B" Motor-Driven Auxiliary Feedwater (MDAFW) Pump Bearino Wear
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On February 10, 1993, the licensee identified metal shavings in the oil from the "B" MDAFW pump.
The licensee later determined that pump operation was not adversely affected by the metal shavings.
As part of the 18-month inspection and cleaning surveillance, P501569, electricians were directed to drain the oil from both inboard and outboard motor bearings (IBMB/0BMB). The IBMB oil was determined not to be contaminated or discolored.
However, upon draining the OBMB, the oil was found to be discolored. After four additional flushes, the oil remained discolored. Subsequently, the licensee determined that the black discoloration was due to metal filings in the oil.
A sample of the oil was shipped off-site for analysis.
The licensee generated work request W528538 to remove the outboard bearing cover to inspect the bearing for damage.
The inspection revealed slight scratches and gouges to the bearing.
This damage had apparently been caused by the shaft lightly impacting the bearing during pump starts.
The bearing was removed and taken to the maintenance shop for repair.
After the bearing was reinstalled, a surveillance test was conducted on the pump.
No problems were noted during the pump retest.
The licensee intends to monitor other safety related pumps for bearing wear.
This issue does not present any long term pump operability
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concerns.
OSP-BG-00001 Procedure and Performance Comments.
The inspector noted that, while surveillance test OSP-BG-00001 did successfully verify that the boron injection flow paths were operable in accordance with Technical Specifications, the procedure needed additional improvements.
Specific comments are discussed below:
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While the acceptance criteria portion of the surveillance test specified that the flow paths from the refueling water storage tank and the boric acid storage tanks through the charging pumps to the reactor coolant system be verified, the only system called out in the body of the procedure was the charging system (BG). The' suction flow path to the charging i
pumps is referred to as the BN system and the discharge of the
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charging pumps goes through the EM system.
The operators
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verified the other systems were operable but the procedure did
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not require that they be verified operable.
Checkoff -list number two contained ambiguous valve position requirements. The equipment operator called the control room and requested guidance to ensure that the valves were left in
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the proper position.
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e Checkoff list number four contained two inaccurate notes. One note stated that only 1 of 4 boron injection tank (BIT) inlet
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valves were required for an operable flow path and the other note stated that only 1 of 2 BIT discharge valves were required for an operable flow path. The two BIT inlet bypass valves are too small to pass enough flow for charging system operability.
For single failure criteria both of the inlet
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and both of the outlet valves must be operable.
All inspector comments were given to the licensee for procedure modification.
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Conclusions Maintenance and surveillance activities continued to be well performed.
Timely follow up actions were demonstrated by electricians when draining oil from the TDAFW pump; their
questioning attitudes reflected a concern for system operability and l
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safety.
Proactive management was apparent during follow up actions for the discovery of a failed LPS, to ensure annunciator system operability and in actions taken to resolve an increasing leakage trend in the
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containment personnel hatch local leak rate test.
No violations or deviations were identified.
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Inspection of Licensee Event Reports (92700)
Through direct observations, discussions with licensee personnel, and a review of records, the following licensee event reports were reviewed to tietermine that reportability requirements were fulfilled, that immediate
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corrective action was accomplished, and that corrective action to prevent
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recurrence was accomplished in accordance with Technical Specifications.
The LERs listed below are considered closed.
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LER 92009:
Failure to adeouately test the manual reactor trip breaker control room handswitch contacts per Technical Specification 4.3.1.1-1 due to an incomplete procedure.
On August 7,1992, the licensee determined that the manual reactor trip breaker (RTB) handwheel switches, the undervoltage and shunt
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trip circuits, had not been properly tested.
Testing of these RTB features were required to meet surveillance requirements of TS 4.3.1.1-1.
Upon discovery of this testing deficiency, the licensee declared the RTBs inoperable.
In addition, the licensee requested, and was granted, a Temporary Waiver of Compliance (TWOC). This TWOC allowed for continued operation of the plant until an emergency TS amendment was issued.
The TS amendment required that testing of the RTB handwheel switches be performed upon the next entry into Mode 4 or the next scheduled refueling outage.
The NRC issued the emergency TS amendment on August 21, 1992. On September 20, 1992, an unplanned reactor trip occurred resulting in the reactor entering Mode 4.
The RTB handwheel switches were tested at that time.
Enforcement action for this event is addressed in Inspection Report No. 50-483/92012(DRP).
Licensee's Evaluation of Root Cause and Corrective Action Root Cause The root cause of this event was personnel error.
The licensee determined that a maintenance engineer inadvertently deleted the procedural steps of MSE-SB-QS001 which met the requirements of TS 4.3.1.1-1.
As part of a consolidation effort to departmentalize RTB
testing, procedure MSE-SB-QS001 was revised in 1988.
As such,
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section 6.5 was deleted from the procedure, but was not incorporated i
into another surveillance procedure.
Also TS 4.3.1.1-1 was not i
removed from the list of requirements to be met for satisfactory performance of the procedure.
- These errors were not identified by the preparer or qualified reviewer during the procedural revision process.
In addition, the licensee determined that the previous procedure revisions adequately tested the undervoltage circuitry, but not the shunt trip circuitry.
Corrective Action
Procedure MSE-SB-QS003, " Manual Reactor Trip Switch, TAD 0T,"
was written to meet TS surveillance requirement 4.3.1.1-1.
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Applicable steps in procedure MSE-SB-QS003 addressed previcus shunt trip testing deficiencies.
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Fifty percent of the individuals involved in the procedure
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revision process have received retraining.
All remaining personnel will be trained by June 1,1993.
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Required that a comparison be performed between the revised
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surveillance procedure and its associated surveillance task sheet (STS) to ensure all procedural changes have been
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accurately incorporated into the STS.
Inspector's Review The licensee's root cause analysis and corrective actions were i
thorough and effective.
The inspector verified that a - new j
procedure, MSE-SB-QS003, had been written to independently verify-l the operability of the UV and ST circuits.
This procedure i
satisfactorily tests the RTB features required to neet surveillance
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requirements of TS 4.3.1.1-1.
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This LER is closed.
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b.
LER 92007:
Reactor trio on low level steam oenerator 'C' due to a i
failed relav driver cara which caused a feedwater isolation valve to i
fast close.
l On May 23, 1992, the reactor tripped on low level in the
'C'
steam
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generator (SG). The SG low level was caused when SG 'C' feedwater
.i isolation valve, AE FV-0041 fast closed.
The control room staff
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took appropriate actions in response to the trip.
f Licensee's Evaluation of Root Cause and torrective Action The licensee later determined that inadvertent valve closure was due l
to the failure of a relay driver card in the main steam and
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feedwater isolation (MSFIS) train
'A'
cabinet.
The card was j
subsequently replaced and satisfactorily tested.
In addition, the
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licensee is evaluatinc upgrades to the MSFIS system to improve
reliability.
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Inspector's Review l
The inspector reviewed two previous failures of input buffer cards j
in the MSFIS cabinet (LERs in 1985 and 1986) to verify whether the l
May 23, 1792, card failure could have been prevented.
The j
inspectors determined the failures were unrelated.
The earlier
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failures were attributed to capacitor failures. Since upgrading the
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capacitors, the cards have not experienced any additional failures.
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The failure of an and gate chip was the-cause of the May 23, 1992,
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card failure.
The licensee's actions taken for root cause determination and to prevent recurrence were adequate.
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i This LER is closed.
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8.
Exit Meetina (71707)
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The inspectors met with licensee representatives (denoted in paragraph 1)
i at intervals during the inspection period. The inspectors summarized the
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scope and findings of the inspection.
The licensee representatives
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acknowledged the findings as reported herein.
The inspectors also i
discussed the likely informational content of the inspection report with
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regard to documents or processes reviewed by the inspectors during the j
inspection. The licensee did not identify any such documents / processes as
proprietary.
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