IR 05000482/1989029
| ML20005E150 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 12/21/1989 |
| From: | Holler E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20005E147 | List: |
| References | |
| 50-482-89-29, NUDOCS 9001030407 | |
| Download: ML20005E150 (12) | |
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b APPENDIX
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q U.S. NUCLEAR REGULATORY C0lWISSION
REGION IV
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NRC Inspection Report: 50-482/89-29 Operating License:
NPF-42 yx,
f Docket: 50-482
' Licenseef Wolf Creek Nuclear Operating Corporation (WCNOC)
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P.O. Box 411-i
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Burlington, Kansas 66830 m
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Facility Name: Wolf Creek. Generating Station (WCGS)
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1 Inspection At: WCGS, Coffey County, Burlington, Kansas-
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Inspection Conducted:' November l-30, 1989
Inspectors:. B. L. Bartlett, Senior Resident Inspector
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Project Section D, Division of Reactor Projects
M. E. Skow, Senior-Resident Inspector; Project Section D, Division of Reactor Projects
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D..V. Pickett, Senior Project Manager
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Project Directorate IV, Division of Reactor
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M; Projects III, IV, V, and Special Projects, NRR L
H. G. Ashar,' Senior Civil Engineer
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Structures and Geosciences Branch
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Division of Engineering Technology, NRR.
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Approved:
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E. J.FHoller, Chief, Project Section D Date
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' Division of Reactor Projects
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~ Inspection Summary
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-Inspection Conducted November 1-30, 1989 (Report 50-482/89-29)
Areas Inspected:
Routine, unannounced inspection including plant status, operational safety verification, monthly maintenance observation, monthly surveillance observation, preparation for refueling, performance of a complex l
surveillance, followup on a previously identified NRC item, and followup of written reports of nonroutine events at power reactor facilities.
Results: Within the areas inspected, one noncited violation was identified.
The violat'on involved a failure to meet a 1-hour reporting requirpent (paragraph 4). The licensee performed a modification to the containment coolers to provide additional structural support. The licensee received 9001030407 891222
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PDR ADOCK 05000482 s
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-information regarding the need-to assess the adequacy of the cooler ' supports
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.more than 1 year before the matter was resolved. An unresolved item regarding timely action in response to the notification of this potential safety problem
. was ioentified (paragraph 4.a). A-safety-related circuit breaker failed to
operate manually (paragraph 4.b). The delay in making the 1-hour required and was discarded prior to the licensee performing a root cause analysis
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report, the absence of timely action in resolving the question regarding
adegJacy of-the containment cooler supports, and the discarding of the circuit
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breaker without performing a root cause analysis indicate the need to make
' personnel at all levels more aware of= the importance of recognizing potential-safety issues in a timely manner. -Notwithstanding this, the licensee's perfonnance during the inspection period was generally good regarding operation of-the plant and the performance of surveillances and maintenance items, i
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DETAILS
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Persons-Contacted
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Principal Licensee Personnel i
- R.. M. Grant, Vice_ President, Quality Assurance (QA)
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- R. Fl. Flannigan, Manager, Nuclear Safety Engineering (NSE)
- R. W. Holloway, Manager, Maintenance and Modifications.
- M. G.' Williams, Manager, Plant Support
- B. McKinney, Manager, Operations
- W. B. Norton, Manager, Technical Support
- W. M. Lindsay, Manager.. QA -
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- C W. Fowler, Manager, Instrumentation and Control (I&C)'
- H. K. Chernoff,- Supervisor, Licensing
- S. Wideman, Licensing Specialist III
- R.-Sims, Equipment Engineering Supervisor
- C. W. Russo Training A. B. Clason, Supervisor, Maintenance Engineering
.D. B. Goetz, Facilities and Modifications
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D. M. Gillespie, Facilities and Modifications The inspectors also contacted other members of the licensee's staff during
the inspection period to discuss identified issues.
- Denotes those personnel in attendance at the exit meeting held on December 1, 1989.
2.
Plant Status The plant operated in Mode 1 (100 percent reactor thermal power) during the inspection period. There were no reactor or turbine trips, 3.
Operational Safety Verification (71707)
The purpose of this inspection was to ensure that the facility was being operated safely and in conformance with license and regulatory requirements.
It also was to ensure that the licensee's management control system was effectively discharging its responsibilities for continued safe operation. The methods used to perform this inspection included direct observation of activities and equipment, tours of the
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facility, interviews and discussions with licensee personnel, independent verification of safety system status and limiting conditions for operation (LCO), corrective actions, and review of facility records.
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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plant cleanliness, instrumentation and alarms, deficiency reports, and
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corrective actions.
- Selected inspector observations are discussed below:
Daily operations were conducted in a professional manner with appropriate decorum demonstrated by operators in the execution of their duties. Crew turnovers observed were complete and necessary information for the safe operation of the plant appeared to be passed to the appropriate personnel.
During this report period, the licensee declared both trains of containment coolers out of service when it was determined that the cooler support structures did not meet seismic design requirements.
Paragraph 4 of this report discusses the details regarding the containment coolers.
No violations or deviations were identified.
4.
Monthly Maintenance Observation (62703)
The purpose of inspections in this area was to ascertain that maintenance activities on safety-related systems and components were conducted in accordance with approved procedures and Technical Specification (TS).
Methods used in this inspection included direct observation, personnel interviews, and records review.
Items verified in this inspection included:
Activities did not violate LCOs and redundant components were operable.
Required administrative approvals and clearances were obtained before initiating work.
Radiological controls were properly implemented.
- Fire prevention controls were implemented.
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Required alignments and surveillances to verify postmaintenance
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operability were performed.
Replacement parts and materials used were properly certified.
- Craftsmen were qualified to accomplish the designated task and additional technical expertise was made available when needed.
Quality control (QC) hold points and/or checklists were used and QC personnel observed designated work activities.
Procedures used were adequate, approved, and up to date.
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Portions of selected maintenance activities regarding the work
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requests (WRs)listedbelowwereobserved. The WRs and related documents
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were reviewed by the. inspectors:
- No.
Activity L
WR 05007-89 Add seismic supports per PMR 3168 to SGN01A WR 05008-89
/W seismic supports per PMR 3168 to SGN018 WR 05009-89 Add seismic supports per PMR 3168 to SGN01C
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WR 05010-89
' Add seismic supports per PMR 3168 to SGN01D
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Selected inspector observations are discussed below:
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a.
The licensee was notified through a problem investigation request (PIR) from Bechtel, dated July 29, 1988, that the containment cooler support structure seismic analysis at another facility was
questionable. The licensee reviewed the vendor's seismic analysis
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for WCGS containment coolers. The review identified some assumptions
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regarding cooler assembly supports that did not agree with the as-built configuration of the cooler assemblies. The vendor's analysis was based ontthe cooler assemblies-being supported at both
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the top and bottom. The actual configuration of the coolers was with supports at the bottom only. The licensee performed its own seismic analysis. This analysis determined that the coolers could meet the conditions.of,a postulated' earthquake with.09 g acceleration, but could not meet the conditions of the safe shutdown earthquake with
.12 g acceleration. The maximum acceleration expected for the operating basis earthquake is.08g.
The licensee declared both trains of containment coolers inoperable and entered TS Action Statement 3.5.2.3.
The action statement allowed the licensee 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to return one train of the containment coolers to operable status and 7 days to return both trains. At WCGS, the containment
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coolers are redundant with containment spray. Their redundancy is reflected in the extended action statement times for the coolers.
The licensee completed the design work on a modification to the containment cooler supports which provided support to the top of the cooler assemblies and completed the modifications to the "B" train L
coolers within the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ~. The licensee then completed the l
modifications to the "A" train containment coolers. Both trains of containment coolers were returned to operable status within the 7-day action statement time limit.
The licensee declared the containment coolers inoperable at 11:51 a.m. on November 17, 1989, and made a 1-hour 10 CFR 50.72 notification at 3:05 p.m.
Failure to meet the 1-hour notification time limit is a violation (482/8929-01). The violation is not cited L
because it is of minor safety concern, was corrected by the licensee,
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and meets the criteria specified in Section V.G. of the General Statement of Policy and Procedure for NRC Enforcement Actions. No
u written response to this violation is required. This noncited violation is closed.
The licensee appeared to manage the modification process well. The modification was made to the containment coolers inside the containment building. Certain areas were cordoned off inside
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containment and were marked with flashing lights to designate the high-high radiation areas. Work groups appeared coordinated in their efforts.
Fabrication of the repair parts was accomplished outside of containment.
Prior to performing the work, the licensee estimated that the total radiation exposure would be 10 man-rem with the actual total-radiation exposure being approximately 7 man-rem. The lower than estimated radiation exposures were the result of management actions discussed above. During performance of the modification, a licensee field engineer identified undersized field welds which were found acceptable for use after an engineering review and disposition.
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Following a review of the modification, the inspector observed weaknesses in the licensee's processing of the PIR. The July 29, 1988, PIR was not entered into the licensee's Industry Technical Information Program (ITIP) until approximately November 16, 1989.
The purpose of ITIP is to review, analyze, and track industry technical information to ensure that lessons learned are translated into corrective actions, if required, to improve the safety and reliability of the plant. The technical information covered by ITIP includes NRC Information Notices, INP0 Significant Event Reports, Westinghouse Technical Bulletins, and other vendor equipment
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technical information that was not solicited from a vendor.
Although the licensee took action after receipt of the PIR, the initial action apparently took the form of a procurement matter regarding obtaining an improved containment cooler seismic analysis from the cooler vendor. The licensee did not recognize the potential significance of the PIR until November 1989, when the licensee performed its own seismic analysis of the cooler supports. The delay in completing action on this PIR may have been less serious in this case because of the previously discussed redundancy of the coolers and the sufficiency of the as-built condition of the cooler su) ports to withstand an acceleration from the operating basis earthqua<e.
The licensee stated that an internal review has been started to determine how the organization responsible for tracking this type of issue missed the PIR and if there are any other issues that should be in ITIP. This will remain an unresolved item pending the inspector's review of the licensee's findings and the inspector's review of the safety significance of the delay in resolving the adequacy of the containment cooler supports (482/8929-02).
b.
During a shift turnover on November 13, 1989, two shift supervisors discussed using an electrician to verify the status of molded case
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circuit breakers prior to hanging clearance tags. The inspector reviewed this item and determined thtt a molded case circuit breaker
. as found closed after an operator had manually ) laced the breaker
w operating handle in the open position. The breater, NG01A AF4, was a
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safety-related feeder breaker for SGK05A, a control building heating, i
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ventilation, and air conditioning unit. The licensee stated that the
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breaker status was found during a routine second check prior to performing work on the affected system. The circuit breaker was
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discarded and a replacement breaker installed on October 25, 1989, under WR 01664-89. - A root cause analysis was not perforned on the
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breaker. The licensee implemented temporary procedures to have electricians verify that other similar molded case circuit breakers actually opened when the circuit breakers were manually opened.
The licensee believed that the circuit breaker would have opened on demand from a circuit fault. However, this can not be stated conclusively because an adequate root cause analysis was not perfonned. The licensee stated that there had been five known fa !sres of similar breakers in the past. None of these five breaker failures was caused by a failure to manually open. However, interviews with various licensee personnel indicated that there may have been previous failures of similar circuit breakers to open manually which were never documented. The inspector discussed with licensee representatives, the need to make personnel at all levels aware of the importance of determining the root cause of equipment failures.
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c.
Maintenance work practices in gener+1 were reviewed during the inspection period. Some specific observations are discussed below:
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There was a comon prectice in the machine shop to make parts or components without QC inspection. However, that pract'ce was limited to non "Q" work. Safety related work appeared to have appropriate hold ptints and QC inspections.
Procedure ADM 01-034, Revision 10 " Internal / External System Cleanliness " required that susceptible parts be cleaned to avoid chemical contamination of compenents from machining process coolbnt and handling. The cleaning process included the use of acetone and demineralized water. Samplina also was required to ensure that any contaminants that could contribute to stress corrosion cracking were removed.
Preshift and prejob briefings were observed in the electric shop.
This was to verify that job content and safety precautions were discussed with the work crews.
Preshift briefings were held in the electric shop with all the oncoming electricians attending, and the day's jobs were discussed. The prejob briefing was a more informai discussion between the lead electrician and other personnel involved in the job, d.
The licensee's diesel generator (DG) vibration and oil leaks program was reviewed. During 24-hour runs of the DGs, the licensee performs
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a walkdown for connected fluid tubing nonconformances. An inspection checklist is included as supplemental work instructions to the WR, Two specific fuel oil tube questions were being addressed by the licensee.
First, the licensee found that fuel oil line tubes were rubbing against the injector pumps. Adjusting the fuel lines so that the tubig would not rub on the DG was observed by the inspector
during work required by WH 02867-89. That WR was listed as an item observed in.NRC Inspection Report 50-482/89-24 in September 1989.
i The licensee stated that replacement tubes were ordered and that their installation was anticipated during the next refueling outage.
Second, on September 6,1989, the inspector noted e fuel line
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vibrating during a DG's scheduled 24-hour run. : This was the same line that failed on the "B" diesel on November 27, 1988, and the "A"
diesel on December 4,1986. The observed vibration was pointed out
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to the licensee during the run and discussed again on September 7,
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1989. The licensee has issued Engineering Evaluation Request (EER) 89-KJ-08 concerning the vibration. The vibration was observed
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even though the tube material and support had been changed. A modification to replace the tube with a flexible hose was under consideration by the licensee as a potential disposition to the EER,
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Procedure ADM 01-201, Revision 3. " Control of Temporary Equipment,"
exempteJ the work table and tool chest in the diesel generator rooms from being immobilized. These items are larger and heavier than other exempt equipment listed in the procedure and appeared to have
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more potential for causing damage. The licensee stated that an EER was being considered for the condition that exists with the equipment in the diesel generator rooms.
The inspector will follow up the licensee's resolution of this matter.
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During a routine containunt entry, the inspector observed the
licensee checking the polar crane. The licensee was checking
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l clearances and accessibility to the wheels and a snubber assembly for j
work being planned for the next refueling outage. The work being planned for PMR 02664 involved the relocation of a support plate in
snubber assembly number four, replacement of a roller, and inspection i
I and refurbishment, if necessary, of the polar crane wheels. The licensee has experienced difficulties with the polar crane since the initial installation of the snubbers during the construction phase.
The difficulties included snubber roller and bearing wear and binding between the snubber support plate and the rail girder. The binding
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of the polar crane wheels on the rails was audible to workers and limited crane movement.
Inspection of the wheel assemblies was planned to ascertain any wear caused by stress generated by the snubber / girder binding.
Previous attempts to reselve the snubber bir, ding, including resetting the rail, and replacement of a roller with a different sized roller, have not been successful.
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previously performed appears to have been properly reviewed and documented.
One noncited violation and one unresolved item were identified in this
area of the inspection.
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5.
Monthly Surveillance Observation (61726)
The purpose of this inspection was to ascertain whether surveillance of safety-significant systems and components was being conducted in
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t accordance with 15. Methods used to perfom this inspection included direct observation of licensee activities and review of records.
Items inspected in this area included, but were not limited to, verification that:
Testing was accomplished by qualified personnel in accordance with an
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approved test procedure.
- The surveillance procedure was in conformance with TS requirements.
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The operating system and test instrumentation was within its current
calibration cycle.
Required administrative approvals and clearances were obtained prior
to-initiating the test.
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LCOs were met and the system was properly returned to service.
- The test data were accurate and complete and the test results met TS
requirements.
Surve111ances witnessed and/or reviewed by the inspectors are listed below:
STS GK-0018, Revision 8. " Control Room Emergency Yent System Train B
Operability Test," performed November 20, 1989 STS RE-006. Revision 3. "End of Life Core Moderator Temperature
Coefficient Measurement," performed November 20, 1989 Selected inspector observations are discussed below:
The end of life moderator temperature coefficient measurement was required by Surveillance Requirement (SR) 4.1.1.3.
The SR required that the measurement be perforned within 7 effective full power days of reaching an equilibrium boron concentration of 300 ppm. However, this SR requirement assumed that the isotopic ratio of boron-10 to boron-11 was that of natural boron (about twenty percent boron-10 and eighty percent boron-11).
Relatively little boron was physically removed from the core and replaced with natural boron; therefore, the boron-10 ratio to boron-11 decreased because of loss of boron-10 by neutron absorption.
The depletion of boron-10 required a higher overall concentration of boron to be as effective as a lower concentration of natural boron. The test was perfortred with a total boron concentration of 320 ppm. With the nuclear design and core characteristics data provided by the vendor, the licensee estimated that the effective boron (boron-10) concentration was equivalent
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-10-to 294 ppm of natural boron when the surveillance was performed. The plant operated with very few transients and shutdowns during this cycle and has had negligible coolant leakage. The surveillance results were satisfactory.
No violations or deviations were identified.
6.
Preparation for Refueling (60705)
The purpose of this inspection was to verif; the adequacy of the receipt, inspection, and storage of new fuel.
During this inspection period, the licensee was receiving and storing new fuel for the next (fourth) refueling outage, scheduled to begin March 8, 1990.
After the licensee had received two shipments of new fuel assemblies, the fuel vendor notified the licensee that some anonelies were found in other fuel assemblies that might also exist in the licensee's fuel assemblies.
The anomalies were " cocked dimples" in the grid assemblies. The grid assemblies provide lateral support to the individual fuel rodlets in the assen61y. Springs hold the fuel rods against the dimples with the grid assembly. The fuel assembly vendor provided a specialist to the licensee.
The specialist inspected the fuel assenblies that had been delivered. The licensee and the vendor found " cocked dimples" in fuel assemblies on site and in fuel assemblies at the vendor's facility that were designated for delivery to the licensee. The licensee stated that the vendor evaluated the condition for use as-is. The condition apparently has existed for a eliminate the occasional " cocked dimples." grid assembly design to long time and the vendor has modified the The licensee further stated that no fuel failures have been attributed to the " cocked dimple" condition.
No violations or deviations were identified.
7.
Complex Surveillance (61701)
The purpose of this inspection was to ascertain whether testing of the more complex safety-related systems or subsystems was in conformance with regulatory requirements and industry guides or standards.
Portions of STS MT-044. Revision 2, " Containment Tendon Inspection," was observed. The licensee contracted with a vendor to perform the surveillance test.
The inspectors reviewed the vendor's procedure and found that it appeared adequate and in accordance with TS Surveillance Requirement 4.6.1.6.1.
The procedure included several significant steps.
A visual inspection was performed to locate any grease leaks or widespread concrete cracking or spalling. The as-found condition of the tested tendons were noted, which included grease condition, coverage, if water was detected inside the grease can, buttonhead condition, and the measured liitoff force. Two tendons, V-84 and 20-AC, were detensioned, and one
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-11-wire strand was removed from each of these two tendons. Tests of those
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individual strands were scheduled for accomplishment after other work was
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completed. The as-left retensioning liftoff data was also recorded. The l
as left grease replacement data was also recorded for those tendons
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inspected. The licensee was required to perforn this surveillance on a selected sample of the containment tendons. Special test equipment was utilized.
Personnel appeared knowledgeable in its use.
No violations or deviations were identified.
8.
Followup on a Previously identified NRC Item (92701)
(Closed) Unresolved Item (482/8924-02): Miniflow Valve Malfunctioning -
This item was opened pending licensee evaluation of safety significance of
the loss of injection flow with the miniflow valve open. This item
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concerned a valve installed downstream of the "A" centrifugal charging pump (CCP). During performance of a surveillance requirement, the valve shut as required and then opened again instead of remaining closed. The
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licensee's initial determination was that failure of the miniflow valve potentially affected the ability of the CCP to perform its safety function and the pump was declared inoperable. Further evaluation by the licensee found that the reduction of integrated safety injection flow resulted in a net peak clad temperature penalty of 68'F for the limiting small break case. The revised peak clad temperature is 1858*F (1790 + 68). This temperature is well below the 2200*F limit ~ stated in 10 CFR 50.46. This item is closed.
9.
Onsite Followup of Written Reports of Nonroutine Events at Power Reactor
Facilities (92700)
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The purpose of this inspection activity was to provide onsite inspection of events at operating power reactors. Specific inspection activities included:
Observing plant status.
- Evaluating the significance of the events, performance of safety
systems, and actions taken by the licensee.
Confirming that the licensee had made proper notification of the l
events and of any new developments or significant changes in plant l
conditions.
Evaluating the need for further or continued NRC response to the
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events.
The following items were considered during the followup:
Details regarding the cause of the event.
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Event chronology.
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Functioning of safety systems as reouired by plant conditions.
- Radiological consequences and persontel exposure.
- Proposed licensee actions to correct the cause of the event.
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Corrective actions taken or planned prior to resumption of facility
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operations.
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The inspectors reviewed the licensee's evaluation of Limitorque Corporation Part 21 (10CFR21) Notification of failures of Melamine torque
e switches on SMB-000 and SMB-00 actuators. The licensee received and
evaluated the report as part of its ITIP as No. 00830. The licensee had 52 Type SMB-000 actuators installed in operating systems. All 52 were inspected and 7 Melamine torque switches were found. All 7 Melamine
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torque switches were replaced with F1brite. The licensee also had 87 Type SMB-00 actuators installed in operating systems. Eighty-five of the Type SMB-00 actuators were inspected and the licensee found that the newer version switch was installed in all 85 of the actuators. Two
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applications, in valves EN HV1 and EN HV7, were not inspected. The licensee performed a documentation review of their 1986 motor operator valve actuator testing system (M0 VATS) testing of these valves on WR 3870-86 and 3869-86 and found that the torque switches were balanced and, thus, were the newer version. No problems on the SMB-00 type
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actuators were found. The licensee received and perfortned the reviews and inspections during their 1988 refueling outage. Operability assessments were, therefore, not necessary before the 7 Melamine torque switches were replaced. The licensee stated that they no longer purchase Melamine switches.
This item is closed.
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No violations or deviations were identified.
10.
Exit Meeting (30703)
The inspectors met with licensee personnel (denoted in paragraph 1) on December 1, 1989. The inspectors sunrnarized the scope and findings of the
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l inspection. The licensee did not identify as proprietary any of the information provided to, or reviewed by, the inspectors.
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