IR 05000285/1992029
| ML20125C049 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 12/07/1992 |
| From: | Harrell P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20125C044 | List: |
| References | |
| 50-285-92-29, NUDOCS 9212110021 | |
| Download: ML20125C049 (17) | |
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APPENDIX U.S. NUCLEAR REGULATORY COMMISSION REGI0fl IV NRC Inspeccion Report:
50-285/92-29 Operating License:
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Licensee: Omat i Public Power nistrict 444 Souti 16th Stre Mall Omaha, Nebraska 6f tw 2247 Facility Name:
Fort Calhoun Station Inspection At:
Blair, Nebraska Inspection Conducted: Oct:6er 11 through November 21, 1992 Inspectors:
R. Mullixin, Senior Resident Inspector R. Azu., Resident Inspector l2 $
Approved:
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P. H. IlgrreJL_.Ctridf, Technical Support Staf f Date Divisiortsef' Reactor Projects Inspection Summary Ar_eas Inspected:
Routine, unannounced inspection of onsite followup of events, operation'l safety verification, maintenance and surveillance vDservatiors, followup on corrective actions for a violation, and onsite followup of licensee event reports.
Results:
The licensee's actions in response to the loss of the safety Channel A
nuclear detector were proper and exh.Dited a high degree of concern for
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safety (Section 2).
Operations, radiological protection, and secarity personnel were
observed to be performing their duties in an excellent ma ner (Section 3).
The housekeeping in the vital areas was found to be very good.
However,
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in the turbine building, several areas require additional attention (Section 3.2).
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Maintenane* work activity to replace flange seals on a nonsafety-related l-
valve exceeded the skill of the craft when the activity began to affect
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l safety-related equipment in the vicinity. The work instruction provided l
9212110021 921207 PDR ADOCK 05000285 Q-PDR
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was general in nature and did not provide caution statements regarding the safety-related equipment located in the area (Section 4)'.
Prestaging of equipment by maintenance personnel prior to beginning work
in a contaminated area demonstrated good ladiological protection practices (Section 4).
Surveillance test procedure adequacy and procedural compliance were
found to be very good (Section 5).
Sramary of Inspection Findings:
. Violation 285/91?i-01 was closed (Section 6).
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91-030,92-009, 92-011,92-015, 92-019,92-021, 92-023, and 92-028 were closed (Sectio. 7).
Attachment:
Attachment - Persons Contacted and Exit Meeting
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-3-DETAILS 1 PLANT STATUS The licensee operated the Fort Calhoun Station at 100 percent power throughout this inspection period.
2 ONSITE RESPONSE TO EVENTS (93702)
^l Channel A Excore Detector Failure On October 25, 1992, the Channel A excore detector, which provided input to the reactor protection system, failed. The licensee initially suspected that a power supply had failed, but troubleshooting determined that the detector had failed inside containment.
This detector supplied inputs to the trip units for high power level, thermal margin / low pressure, and axial power distribution.
When the detector was determin2d as the cause, the licensee appropriately entered a 7-day shutdown action statement, as required by the Technical Specifications.
The replacement cf the detector would have required a plant shutdown; however, the licensee decided to swap the safety channels with the nonsafety control channel detectors. This would not require a plant shutdown nor a containment entry. The licensee developed the following action plan to proceed with the swapping of nuclear detectors:
Prepared Engineering Analysis EA-FC-92-78 and a 10 CFR Part 50.59
analysis to address swapping safety Channels A and D with nonsafety control Channels A and B, respectively.
Issued Operations Memorandum 92-10 to restrict c:ntrol rod movement
until testing was completed on the swapped safety channels.
Completed the swapping of the channels per an approved temporary
modification.
The licensee successfully completed the swapping of the safety channels with the nonsafety control channels an October 30.
2.2 Conclusions The licensee's actions in response to the loss of safety Channel A nuclear detector were proper and exhibited a high degree of concern for safety.
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-3 OPERATIONAL SAFETY VERIFICATION (71707)
3.1 Routine Control Room Observations The inspectors observed operational activitiet throughout this inspection period to verify that proper control room staffing and control room professionalism were maintained.
Shift turnover meetings were conducted in a manner that provided for proper communication of plant status from one shift to the other.
Discussions with operators indicated that they were aware of plant and equipment status and reasons for lit annunciators.
The inspectors observed that Technical Specification limiting conditions-for operation were properly documented and tracked.
Operators were-observed to properly control access into the control room operating area.
Plant management was observed'in the control room on a daily basis.
The inspector reviewed the control room log books for danger tags, caution tags, and locked valve deviations. All the logs were observed to be complete.
On November 11, 1992, the inspector selected examples from each log to verify that the component reflected the condition. stated in the. log book. The inspector _ verified that the correct tags were properly hung on the component for Tagouts 90-0068, 92-0253, 92-3255, 92-2187, 9I-2235, and 92-2314. The.
inspector also vnrified that twa valves (MS-103 and SI-342) were returned to their locked-c.losed position, as indicated in the locked-component deviation log.
3.2 Diant Tours The inspectors toured various areas of the plant to verify that proper
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housekeeping was being maintained.
The housekeeping in the vital areas was found to be very good. However, in the turbine building, many areas required attention. The licensee had also identified the housekeeping concerns and has formulated an action plan to solve the housekeeping problem in tne turbine.
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The inspectors verified, during plant tours, that various valve and switch positions were correct for the current plant conditions.. Personnel were observed obeying rules for personnel safety and rules for escorts, visitors, p
entry, and exits into and out of vital areas.
3.3 Radiological Protection Program Observations The inspectors verified that selected activities of the licensee's
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l radiological protection program were properly implemented.
Radiation and
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contaminated areas were properly posted and controlled. - Health physics personnel were observed routinely' touring the controlled areas.
The inspector observed on three occasions that licensee personnel performed the correct process when individuals alarmed the personnel contamination monitor while -
attempting to exit the radiologically controlled area.
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-5-3.4 Security Program Observations The inspectors observed various aspects of the licensee's security program.
Persnnnel and packages entering the protected area were observed to be properly searched.
Nondesignated vehicles entering the protected area were found to be properly escorted by armed security personnel, and security officers were ibserved performing their tours and/or manning their assigned posts.
Compensatory measures were observed to be properly performed whenever a security barrier was inoperable.
On November 17, 1992, the plant security system was unavailable due to hardware changes to the security computer system.
This required security officers to be posted at vital doors to unlock the doors and manually log all personnel entoring/ exiting vital areas.
The inspector noted that the officers maintained proper control over the vital area doors and continued to observe the doors for a period after the security computer was returned to service.
3.5 Conclusions Operations, radiological protection, and security personnel were observed to be performing their duties in an excellent manner.
4 MAINTENANCE OBSERVATIONS (62703)
4.1 Steam Generator Blowdown Control Valve Seal Replacement On November 10, 1992, the inspector witnessed the maintenance activity that was performed to replace the flange seals on Steam Generator B; Blowdown Control Valve HCV-1389.
This work activity was controlled by Maintenance Work Order 924494 and its associated work instruction. -The maintenance work order had been reviewed and approved, as noted by the appropriate signatures.
The inspector reviewed the maintenance work order and determined'that the information provided was accurate in identifying the item to be worked on, with specific postmaintenance testing requirements. The' work-instruction was found to be general in nature, relying more on the skill cf the craft.
The work was performed in the lower mechanical panetration room (Room 13),
i located in the auxiliary b ding.
Due to the potential for contamination when the valve was removeo, the area was required to be. roped off by radiological protection personnel.
Prior to this, thr licensee prestaged the equipment needed and removed all applicable piping insulation.
The maintenance personnel performing the maintenance activity signed in unden the appropriate radiation work permit and wore the required protective clothing as delineated in the radiation work permit.
Prior to initiating the work, the maintenance personnel noted that a hoist would be required to support the valve once it had been disengaged from the pipe. As a result, the maintenance employee, working within the r: ped-off l
area, began tying slings to the steam generator blowdown line snubber support and the high pressure safety injection pumps' (SI-2A and -2C) alternate
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suction line. Snubber SI-104.
The inspector questioned whether an engineering evaluation had been performed to determine if this was an acceptable practice since some of the equipment. involved was safety related. The licensee employee stated that none had been performed but that he did not think that it wculd be a problem. At this point, the licensee employee decided to halt the'
work in progress and contacted the system engineer for guidance.
Following the discussion with-the system engineer, note was added to the work instruction stating that the slings used to support the hoist could be hung off of the steam generator blowdown line snubber support and a component cooling water pipe located in the vicinity and that_this would have no impact on the performance of these systems.
The-inspector witnessed the-remainder of the maintenance activity and noted that the maintenance personnel performed this effort in an appropriate manner. Good radiological protection. practices-were noted in the removal of equipment, from the contaminated area, by the maintenance personnel.
Good health physics coverage was also noted throughout this activity.
4.2 Conclusions Overall, performance of maintenance personnel was found to be good, with good adnerence to adiological protection practices with regard to the removal of equipment from the contaminated area.
It must be noted, though, that the
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determination as to when and if it is appropriate to hang equipment from piping or piping supports was beyond the skill of the craft during this activity. Although the valve that was worked on was in a nonsafety system,
consideration upon writing the work instructions for the maintenance work order should-have taken into consideration the fact that safety-related equipment in the vicinity could be adversely affected by the maintenance activity.
5 SURVEILLANCE OBSERVATIONS (61726)
5.1 Reactor Anomalies On November 5,1992, the inspector witnessed the performance of Surveillance Test Procedure OP-ST-RX-0001, " Reactor Anomalics." This surveillance is performed weekly to satisfy the requiremeats of Technical Specification 3.10(1)b for the comparison of the overall core reactivity balance to predicted values.
This test can be performed by either a licensed operator or the shift
technical advisor.
The surveillance witnessed was performed by the shift technical advisor. The inspector verified that the latest revision was being used and that the procedure was being followed. The inspector performed the
test independently and the results matched those obtained by the licensee.
All. results were within the procedural acceptance criteria.
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e-7-5.2 Conclusions Surveillance test procedure adequacy and compliance'were found.to be very good.
E FOLLOWP ON CORRECTIVE ACTIONS FOR A VIOLATION (92702)
6.1 (Closed) Violation 285/9121-01: Failure to Take Adeauate Corrective Action This violation concerned the licensee's failure to promptly correct an identified condition in that a station battery jar was discovered cracked on July 1, 1991, but no corrective action was taken until a similar event.
occurred on September 11. The licensee declared both station batteries inoperable on September 12 and instituted a plant shutdown.
A root cause analysis was. completed on August 12, after the battery jar crack discovery on July 1.
This analysis concluded that the cracking was caused by stresses in the jar cover due to corrosion buildup around the positive post.
The root cause analysis recommended that the batteries be replaced with those having an improved terminal post seal design.
Thus, information was available to management on August 12 that a potential common mode failure existed.
However, the Plant Review Committee did not meat to discuss the battery cracking problem until August 29. At that. meeting, the Plant Review Committee Chairman directed that an operability determination be presented at the next scheduled meeting.
This operability determination was not performed.
The licensee attributed the failure to take prompt corrective. action on the-
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inadequate application of programs and procedures for identification and correction of adverse conditions. The level of significance after the July I cracking event was influenced by past experience, vendor guidance, and engineering judgment. The licensee had performed an engineering evaluation after a similar crack was discovered in March 1991.
This evaluation concluded that there was not an operability concern. Thus', the July I cracking was influenced by this previous evaluation, which had significant vendor input.
In addition, the Plant Review Committee-Chairman's directive to perform an operability determination was not performed since it was not tracked after the meeting.
The licensee's ccrrective actions included revising Nuclear Operations Division Quality Procedure N00-QP-19, " Root Cause Analysis Guideline."
The revisions included the following:
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Vendor information, if critical to th; analysis, is required to have the
vendor analysis documented.
Potential common-mode failures should be discussed in the root cause
analysis.
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-8-Review of equipment history records for similar failures is required.
- The approved root cause analysis must be forwarded to the Plant Review-
Committee within 7 days.
In addition, Nuclear Safety Review Group Procedure NSRG-3, " Reviews and Investigations." was revised to enhance their review of root cause analysis reports.
The licensee's actions should provide a quicker and more thorough review of deficient conditions.
In addition, items requiring followup from a Plant Review Committee meeting are tracked and assigned the highest action priority.
These improvements in the licensee's program should be sufficient to minimize the possibility of further occurrences.
7 ONSITE REVIEW 0F LICENSEE EVENT REPORT 3 (92700)
7.1 1 Closed) Licensee Event Report 285/90-015:
Nonconservative Setpoints for the Low Temperature /0verpressure Protection S_ystem This report described how the variable setpoints for the two power-operated relief valves used for low temperature / overpressure protection were nonconservative.
The licensee determined that tt+ cause of this nonconservatism was deficiencies in the design process for the low temperature /oserpressure protection system.
The licensee determined that no historical conditions existed that would have had an impact on reactor coolant cystem integrity due to the nonconservative setpoints.
The licensee's engineering department designed the variable setpoint system for the relief valves in 1984.
However, no consideration was given on how the
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system would operate during a pressure transient such as the inadvertent operation of a reactor coolant pump at low temperatures.
The licensee determined that the cause of this event was a lack of adequate design review and an interface between the design group and the technical support department.
The licensee had reorganized the engineering department and improved procedural guidance on the preparation of design packages in 1988.
These changes improved the connunications problem that contributed to the event.
The licensee also established operational limits (pressure-temperature) to ensure reactor coolant system integrity.
Other corrective actions included proceduralized requirements for reactor coolant pump restart and high pressure safety injection pump operating criteria.
On June 1,1992, the licensee submitted an application for amendment of the operating license.
This application provided for improved controls on low temperature / overpressure protection.
This amendment had nct been approved at the end of this inspection period.
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-9-7.2 (Closeo) Licensee Event Report 285/91-015:
Radiation Monitor RM-060 Inoperable Due to Seismic Concerns
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This report described the licensee's conclusion, on February 28, 1991, that-the flow totalizer for Radiation Monitor RM-060 was nonseismically supported
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inside of the monitor's cabinet. Monitor RM-060 is the plant stack iodine monitor.
The licensee determined that the cause of the event was the failure to analyze for seismic considerations during the develooment of the modification for installation of the flow totalizer in 1977. The licensee determined the safety. significance of this event to be minimal since. Monitor RM-060 is one of five radiation monitors that can initiate a ventilation isolation actuation signal.
In addition, the operability of Monitor RM-060 is solely designed-for iodine monitoring and accountability.
It is also not required by_the Technical Specifications for initiation of a ventilation isolation actuation signal.
The licensee's corrective action was to seismically support the flow totalizer on the outside of the cabinet. This was compluted on July 13, 1991, and the radiation monitor was declared operable.
The inspector reviewed the licensee's completed work and found the flow totalizer to be securely installed on the ronitor cabinet. The licensee has made improvements to the modification control process since this flow totalizer was installed in 1977.
The e improvements should' minimize the possibility of a simiin occurrence of this type.
7.3 (Closed) Licensee Event Report 285/91-018:
Inoperable Station Batteries Due to Inadequate Desian of Terminal Post Seals This report documented the discovery, on September 11, 1991, of a crack in the front wall of a station battery cell. This crack, in addition to previous-cracks discovered, prompted the licensee to declare both -station batteries inoperable and chut down the plant.
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The licensec's root cause analysis found an inadequate design of the battery
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cell terminal post seals. This design did not adequately allow for the buildup of corrosion products at the positive terminal.
The growth of the corrosion radially outward created stresses on the plastic cell. jar, which resulted in the cracking.
The licensee shut down the plant and temporarily replaced all of the battery L
cells with a newer post seal design. The new batteries were obtained from-L various sources, which-included other nuclear facilities, until completely new
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batteries could be ordered and installed.
This was completed and the plant was restarted.
Procurement of completely new batteries was begun and the new.
batteries were installed during the 1992 refueling outage. The inspectors
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problems.
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7.4 (Closed) Licensee Event Report 285/91-027:
Violation of Containment Integrity by Opening Valve WD-1060 During Sampling This report documented the licensee's discovery, on November 18, 1991, that containment integrity had been violated when samples were taken from the reactor coolant drain tank discharge line.
The sampling was done through Valve WD-1060, which is a drain valve between the two containment isolation valves.
Valve WD-1060 was used as a sampling point on 20 different occasioni during tbr investigation of abnormal increases in reactor coolant drain tank level, NRC Violation ?a5/9126-02 was cited as a result of this event. The corrective actions were reviewed during the closecut of this violation in NRC Inspection Report 50-285/92-14.
This licensee event report is closed.
7.5 (Closed) Licensee Event Report 285/91-028:
Unmonitored Release on loss of the 161-kV System This event concerned the licensee's discovery, on December 1,1991, that the sample pump for the exhaust stack gas,_ iodine, and particulate monitors in the laboratory and radioactive waste processing building was not running.
This constituted an unmonitored release since release from the building had been made while the pump was not in operation.
The cause was determined to be due
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to the previous day's loss of the 161-kV offsite power supply resulting from severe winter weather. The design of-the power supply to the sample pump and the exhaust _ fans, where the sample is taken from, is that when power is restored the f ans restart, but the sample pump had to be started locally. The licensee determined that the routine releases that occurred while the sample pump was inoperable had minimal-safety significance.
The licensee's initial corrective action was to alert operators to the need to reset the power supply to the sample pump upon a loss of power.
In additica, a temporary modification (TM-04) was installed, which would automatically restart the sample pump when power is restored._ The inspector reviewed this-temporary modification and found it would accomplish this purpose.
The licensee's long-term corrective action was to implement Engineering Change Notice 92-524, which made the temporary modification permanent.
In addition, it provided control room annunciation upon the loss of power to the sample pump.
The inspector reviewed the licensee's corrective actions and fcund them to be sufficient to address this licensee event report.
7.6 (Closed) Licensee Event Report 285/91-029:
Personnel Air Lock Leak Rate Test Deficiency This licensee event report described a condition that was discovered, on December 2,1991, by a special services engineer The engineer, while reviewing procedures for containment leak rate testing, identified that the
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-11-Type B test procedure for the personnel air lock did not adequately test-the inner personnel air lock equalizing valve, as required by Technical Specification 3.5.(3)d.
The root cause of this event was attributed to an inadequate procedure change review process that was used when a change was made to the testing procedure in 1974.
Contributing causes to this event included: (1) the lack of understanding or knowledge of the regulations regarding the design basis involved, by all individuals who reviewed and/or approved the procedure change, as well as, by those individuals who have performed the biennial reviews for this procedure; and (2) the fact that in 1974 no procedural requi - ent existed for performing documented safety evaluations for procedure-changes.
The following are corrective actions that were implemented by the licensee immediately following the discovery of the test deficiency and the corrective actions that were implemented to preclude recurrence of this event:
On December 4,1991, the inner personnel air lock door equalizing valve
was declared inoperable due to the lack of proper leak rate testing.
Administrative controls were established to ensure that containment integrity was maintained by danger tagging the outer door closed.
A procedure (IC-3T-AE-0006) was developed to leak test the inner door
equalizing valve. This test was approved and incorporated into station procedures on December 6.
On December 7 this test was performed with acceptable test results.
The safety evaluation and review process for procedt e changes has been
substantially upgraded since 1974 and is documented in Nuclear Operations Division Quality Procedure N00-QP-3, "10 CFR 50.59 Safety Evalut ions."
The biennial review process has been upgraded as part of an overall
enhancement and is documented in Standing Order S0-G-36, " Biennial-Review."
By June 15, 1992, all Type B leak rate test _ procedures were reviesed,
along with the current configuratior.
- all Type B penetrations, to ensure that the penetrations were beb., 'ested in accordance with
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10 CFR Part 50, Appendix J, criteria.
The inspector reviewed the documentation for the completion of the corrective-actions. As a rcsult of the completed actions, this licensee event report is closed.
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7.7 (Closed) Licensee Event Report 285/91-030:
Radiation Monitors Out of Sarvice With Containment Pressure Reduction in Progress This report-documented the December 10, 1991,_ event when_a containment pressure reduction was in progress while the auxiliary building ventilation stack iodine, particulate, and gas radiation monitors (RM-060, -061, and -062, respectively) were removed for filter replacements. The shift supervisor directed, during the shift briefing, a licensed operator to terminate'the containment pressure reduction early before filter replacements began.
However, the licensed operator became busy on another task and forgot.to terminate the pressure reduction.
Another licensed operator, who was unaware of the directive to terminate the pressure reduction,' removed the radiation monitors from service for filter replacements.
It wasn't until 6 minutes-later that an operator realized the Technical Specification requirement was not being met and terminated the containment pressure reduction.
The licensee's evaluation determined that this event resulted from a breakdown in verbal and written communications and work practices. The examples listed by the licensee included:
The operator that was directed to terminate the containment pressure
reduction early did not relay the fact-that he was unable-to 'do so.
The surveillance test procedure used for filter replccements did not
require the shift supervisor's approval prior to removing the radiation-
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monitors from service Operating Instruction 01-RM-1, " Radiation Monitoring Normal and Accident
Operation," did not ensure that releases had been secured prior to removing the radiation monitors from service.
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The operators involved failed to verify that the containment pressure
reduc + ion had been termlaated prior to filter replacements.
l The corrective actions taken by the licensee included the discussion of this -
event in licensed operator requalification training and procedural revisions.
Surveillance Procedure CH-ST-VA-0001, " Auxiliary Building Exhaust Stack; Alpha, Iodine and Particulate Sampling and Analysis," was revised to require a shift supervisor's signature prior to removing the radiation monitors from service.
In addition, Operating Instruction 01-RM-1 was revised to require the. securing.
of releases prior to removing gas and particulate monitors from service for filter replacement.
The inspector reviewed the licensee's corrective actions and found them to be sufficient to close this licensee event report.
t 7.8 (Closed) Licensee Event Report 92-009:
Unplanned Actuation of the Ventilatior. Isolation Actuation Signal This event occurred, wbtle in a refueling outage, when electricians replaced
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fuses for Channel B of the ventilation isolation actuation signal after i
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-13-completing maintenance.
The maintenance that being performed was the cleaning and repair / replacement of the lockout relays for the five radiation monitors -
for Channel B of containment radiation high signal.
If any one of the five radiation monitors reaches its setpoint, a containment' radiation high signal-is generated, which produces a ventilation isolation actuation signal. 'In this event, one of the five lockout relays was in the tripped condition when'
fuses were reinstalled.
All equipment operated as designed when the inadvertent actuation signal was recei,ed.
The licensee determined the root cause of this event to be an inadequate
verification and validation of Procedure EM-RR-EX-0201, " Repair / Replacement of Lockout Relays." This procedure did not require that the lockout relays tre :in a particular position (reset) prior to reinstalling fuses. Also, the maintenance work order instructions did not specify this requirement. An over reliance on the technical adequacy of the procedure and maintenance work order was determined to be a contributing cause.
The licensee's corrective actions ir !nded personnel training on this event
and the revision of Procedure EM-RR-EX-0201.
The inspector verified that the procedure was revised to include the proper lockout relay condition prior to energizing the relays.
7.9 (Closed) Licensee Event Report 92-011: Unacceptable Valve Arrangem:nt for Service Air System Containment Penetration H-74 ibis event resulted from the discovery, during the:1992 refueling outage, that
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the compressed air containment penetration valve arrangement did not meet the isolation criteria required for a containment atmosphere-expused system.
Containment Penetration M-74 consisted of an outboard automatic isolation-valve (HCV-1749) and an inboard normally-open manual valve (CA-555).
For this
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arrangement to be acceptable, the compressed air system pressure would have to
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be greater than the maximum containment design-pressure. Huwever, since the air compressort would not automatically load onto the emergency diesel generators under accident conditions, the compressed air system pressure would be less than the maximum containment pressure. Thus, the valve arrangement was not. acceptable.
The licensee determined that the original Final Safety Analysis Report stated
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A revision to this report, in 1971, stated that a compresser would not automatically start..However,'the containment penetration for the compressed air system was not modified.
1-The licensee's initial currective action was to remove Valve CA-555-and
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install a blank flange before starting up from the refueling outage. This-
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alleviated any immediate safety concern.
In addition, the. licensee reviewed other valve arrangements for' containment penetrations and found no other similar problems. The licensee plans to install a qualified inboard isolation
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valve during the next refueling occage. This licensee: event repor: i s closed
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based on the licensee's completea and proposed corrective actions.
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-7.10 (Closed) Licensee Event Report 92-015:
Loss of Shutdown Cooling Flow Contrel and Flow Indication This report documented the April 12, 1992, event of power being lost to the shutdown cooling. flow control valve controller and.hutdown cooling flow indication.
fhe plant was in a refueling outage at the time of the event and in an abnormal electrical alignment.
The control rocm operators determined the cause of the loss of power and restored shutdown cooling within 7 minutes.
During the event, the reactor coolant system temperature increased 6*F.
The NRC issued Violation 285/9209-01 based on this event for an inadequate procedure that allowed the plant to be in the abnormal electrical lineup.
This licensee event report is closed based upon the review to be performed of the licensee's corrective actions for the violation.
7.11 (Closed) Licensee Event Report 92-019:
Control Element Assembly Drop and Plant Shutdown,j},1 to Clutch Coil Failure This report documented the May 31, 1992, event when a control element assembly dropped into the core while at 100 percent power.
The licensee reduced power
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to 70 percent to attempt to recover the rod but was unsuccessful. A plant
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shutdown was initiated per Technical Specification requirements and a Notification of Unusual Event was declared.
The control element drive mechanisms at the fort Calhoun Station use a rack and pinion mechanism to perform vertical movement of the assembly.
The control rod is held in place by an electromagnetic clutch. The licensee i
investigated the cause of the control rod drop and found that the clutch coil had failed. The coil was replaced and resistance readings on the clutch coils for the other control rods were taken.
No problems were noted and the plant was restarted.
The licensee determined the root cause of this event to be the material failure of the clutch coil in Control Element Drive Mechanism 35.
A laboratory analysis of the coil concluded that the failure was caused by an electrical short, possibly due to a manufacturing defect or an induced overstress such as a power surge.
The licensee's immediate corrective actions to replace the faileu coil and l'
test the remaining clutch coils, prior to plant startup, was conservative.
I r, addition, the licensee has proposed to evaluate potential means of perfarming predictive maintenance on the clutch coils.
The licensee will perform a l
further examination of the Control Element Assembly 35 clutch coil circuit
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during the 1993 refueling outage to look for signs of overstress in a series
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l resistor. The completed and proposed corrective actions are sufficient to
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close this licensee event report.
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. Closed)' Licensee Event report-92-021:
Failure to Initiate a Fire Watch for an Inoperable Fire Door This report resulted from the discovery, on June 11, 1992,_ that the fire door to the charging pump valve room would close but not latch.
The broken latch was discovered by a nonlicensed operator Caring the performance of monthly Surveillance Test Procedure OP-ST-FP-0001, " Fire Protection System Inspection and Test." The operator who made the discovery noted that +he latch was broken but titled to notify the shift supervisor or initiate a maintenance wors request.
Procedure OP-ST-FP-001 requires that it be performed.in accordance w...i Procedure 01-FP-6, " Fire protection Syitem Inspection and Test."
The operator did not-consider the fire door inoperable since a note on the fire door checklist in Procedure 01-FP-6 st ".ed that all doors shall be closed unless in use or a fire watch is posted.
Since the door would close, it was not declared inoperable.
On June 13, the shift supervisor reviewed the completed surveillance test but did not consider a ffre watch to be needed based on the same note.
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However, on June 17, a general maintenance craftsperson, familiar with fire door requirements, noted the broken door latch and contacted system engineering to generate a fire barrier impairment and initiate a fire watch.
The applicable system engineer noted that the Technical-Specification requirement for instituting a fire watch was not met. The licensee's immediate corrective was to institute a fire watch and repair the latch.
The licensee determined the root cause to be' ambiguous instructions contained in the note at the beginning of the checklist for Procedure 01-FP-6. A contributing cause was determined to be the failure of. the operator to report his findings to the shift si.pervisor as required by procedure. The licensee's corrective actions to prevent recurrence included revising Procedure 01-FP-6
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and Standing Order S0-G-58, " Control of Fire Protection System Impairments,"
to clearly define that a.n opersble fire door.must have a functioning latch-mechanism.
In addition, the shift supervisors discussed with their crews the importance of innediately notifying the shift supervisor of any anorc.alies or deficiencies v: hen performing a surveillance test. These actions are-sufficient to close this licensee event report.
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7.13.(Closedl Licensee Event Report 92-023:
Reactor Trip Due to inverter Malfunction and Subseguent Pressurizer Safety Valve Leak This event occurred on July 3, 1992, when the plant tripped on high pressurizer pressure while at_100 percent power. Maintenance on'a nonsafety-related inverter resulted in the monentary loss of power to the turbine electrohydraulic control system and the subsequent closure of the
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turbine control valves.
The high pressurizer pressure resulted in Pressurizer Safety Valve RC-142 lifting and failing to tully reseat. This-failure of Valve RC-142 resulted in the loss of approximately 20,000 gallons of ~ reactor coolant to the containment sump.
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This event resulted in an Augmented Inspection Team being dispatched to-the-Fort Calhoun Station on-July 4.
The cause of this event and the corrective actions taken are documented in NRC Inspection Report 50-285/92-18.
In addition, a special inspection was conducted from August 24 th.ough September 3, due to the premature opening af Valve RC-142 on August 22. The results of this inspection and the licensea's corrective actions are documented in NRC Inspection Report 50-285/92-21. This licensee event report-is closed based on the inspections listed above.
7.14 (Closed) Licensee Event Report 92-028:
Portial loss of load Resultina in Pressurizer Safety Valve lift and Subseauent Reactor Trio On August 22, 1992, the plant tripped on thermal margin / low pressure while at 100 percent power. A failed power converter in the electrohydraulic control system resulted in the partial closure of the turbine control valves. With a-partial Inss of load, the reactor coolant system pressure increased, but a pressurtzer code safety valve (RC-142) lifted prematurely before the plant
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tripped on high pressure.
~,ne thermal margin / low pressure trip. occurred as pressure decreased due to the open safety valve.
This event was the subject
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of a special inspection conducted from August 24 through September 3.
The
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results of that inspection and the licensee's corrective actions are documented in NRC Inspection Report 50-285/92-21. This licensee event report is closed.
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ATTACMMENT 1.-
PERSONS CONTACTED-
- R. Andrews, Division Manager, Nuclear Services J. Bobba, Supervisor, Maintenance-J. Chase, Assistant Manager, Fort'Calhoun Statfon
- M. Frans, Supervisor, Systems Engineering
- W. Gates, Vice President, Nuclear J Geschwender, Station Licensing Engineer R. Jaworski, Manager, Station Engineering
- R. Johansen, Supervisor, Maintenance Support
- W. Jones, Senicr Vice President
- D. Lippy, Station Licensing Engineer
- W. _Orr, Manager, Quality Assurar,ce and Quality Control
- T. Patterson, Manager, Fort Calhoun Station
- R. Phelps, Manager, Design Engineering A. Richard, Assistant Manager, Fort Calhoun S+ation
- J. Sefick, Manager, Security Services
- C. Simmons, Station Licensing Engineer F. Smith, Supervisor, Chemistry
- R. Short, Manager, Nuclear Licensing and Industry Affairs
- J. Tesarek, Supervisor, Simulator Services J. Til's, Operations Superviscr
- Denotes licensee personnel that attended the exit meeting.
In addition to the personnel listed above, the inspectors contacted other personnel during this inspection period.
2 EXIT MEETING
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An exit meeting was conducted on November 25, 1992. During this meeting, the inspectors reviewed the scope and findings of the report.
The licensee did not identify as proprietary any information prov e d to, or reviewed by, the-inspectors.