IR 05000440/1992024
| ML20127M804 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 01/20/1993 |
| From: | Lanksbury R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20127M789 | List: |
| References | |
| 50-440-92-24, NUDOCS 9301290018 | |
| Download: ML20127M804 (21) | |
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U. S. NUCLEAR REGULATORY _ COMMISSIO . REGION Ill Report No. 50-440/92024(PRP) ,
Docket No. 50-440 License No. NPF-58~
Licensee: _ Cleveland Electric illuminating Company Post Office Box 5000 Cleveland, OH 44101
. Facility Name: Perry Nuclear Power Plant Inspection At: Perry Site, Perry, Ohio Inspection Conducted: November 21 through December 28, 1992 Inspectors: A'. Vegel
, E. Duncan J. Hopkins i N Approved By: N]O - 72 - s ll2d93 f, Ch'tef 'Date D. Lank Reactor Proje %bu' cts Section 3B Inspection Summary Inspection on November 21 throuah December 28. 1992 (Report N '50-440L92024 (DRP))
Areas Inspected: Routine unannounced safety inspection-by resident and region based inspectors of licensee action on previous. inspection findings, licensee event report _ followup, surveillance observations, maintenance observations, ,
operational: safety verificatica, event followup, engineered-safety features-system walkdown, and evaluation of licensee self_-assessment capabilitie Results: Of the eight areas inspected, three violations were identified concerning a missed average power-range monitor (APRM)-surveillance (paragraph 3..d),-an inoperable accident monitoring containment. pressure-instrumentation channel (paragraph 3.e), and failure to maintain adequate cleanliness (paragraph 7.b). In addition, two non-cited violations (NCVs)
were idens;fied-in-_the area of licensee event report followup (paragraphs 3.-a
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and 3.b).
The'following is a summary of the licensee's performance during this inspection period:
Plant Operations The reactor plant was operated at or near full' power during the-inspection period with the exception of a downpower to.80_-percent reactor power on December 6, to perform a flux tilt _and fix steam leaks ~.
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Operator response to-two safety relief valves unexpectedly opening on 9301290018 930121-i PDR .ADOCK 05000440 1 0 PDR ,,
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November 21 was_ considered good. A violation was cited concerning a <
personnel error' that resulted in an inoperable accident monitoring-
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containment pressure instrument'(N0V-50-440/92024-02).
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daintenance/ Surveillance The quality of observed maintenance and surveillance activities was generally good. Repair efforts of the Division 1 diesel generator and resi6 sal heat removal pump "B" room cooler were _ considered good. - Poor post-maintenance cleanup practices contributed to poor housekeeping conditions in the plant (NOV 50-440/92024-03). .A violation was cited concerning a personnel. error that resulted in a missed average power-range monitor surveillance test (NOV 50-440/92024-01).
Emeraency Preparedness The licensee conducted on emergency preparedness exercise on December 9, 1992. The results:of the NRC evaluation were documented in ,
Inspection Report 50-440/9202 Enaineerina and Technical Support The engineering evaluation and troubleshooting of the residual heat removal loop "B" test return valve was goo Safety Assessment and Quality Verification ,
The quality of reviewed event reports was acceptable. The onsite and--
offsite review committees were evaluated as effectiv '
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DETAILS
. 1, Persons Contacted Cleveland Electric Illuminatina Comoany- -
- M. Edelman, Executive Vice President - Power Generation, Centerior Energy
- R. Stratman, Vice President - Nuclear
- D. Igyarto, General Manager, Perry Nuclear Power Plant (PNPP) -o K. Donovan, Manager, licensing and Compliance
- M. Gmyrek, Operations Manager, PNPP-S. Kensicki, Director, Perry Nuclear Engineering ,
Department (PNED)
F. Stead, Director, Perry Nuclear SJpport Department -
(PNSD)
- H. Hegrat, Compliance Engineer, PNSD E. Riley, Director, Perry Nuclear Assurance Department (PNAD)
- Coleman, Manager, Quality Assurance Section, PNAD V. Concel, Manager, Technical Section, PNED D. Conran, Compliance Engineer, PNSD M. Cohen, Manager, Maintenance Section, PNPP P. Volza, Manager, Rcdiation Protection Section
- R, Tadych, Manager, Quality Control Section, PNAD D. Cobb, Superintendent, Plant Operations, PNPP U. S. Nuclear Reaulatory Commission
- A. Davis, Regional Administrator, RIII
- E. Greenman, Director, Division of Reactor Projects, RIII
- H. Miller, Director, Division of Reactor Safety, RIII -
- J. Hannon, Director, Project Directorate III-3, Office of Nuclear =
Reactor Regulation (NRR)
L. Greger, Chief, Branch 3, Division of Reactor Projects, RIII
- A. Vegel, Resident Inspector, RIII E. Duncan, Reactor Engineer, RIII J. Hopkins, Project Engineer, RIII w
- Denotes those attending the exit meeting held on December 28, 199 # Denotes those attending the management meeting h_ eld i_n tha Region III office on December 17, 1992 Licensee Action on Previous Inspection Findinas (92701. 92702) (Closed) Violation (440/91G12-01(DRP)): Multiple examples of -
failure to follow procedures-during control rod manipulations on -
May 17, 1991. During this inspection period the inspectors reviewed licensee corrective actions and assessed effectiveness of
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l those actions to prevent recurrenc Based on review of licensee documentation, all corrective action commitments were complete . In addition, based on routine observation of control room activities since this event, specifically observation of control rod manipulations during plant startups and shutdowns, operators demonstrated procedurel compliance and received proper oversight during control rod movements. The inspectors concluded that licensee corrective actions appeared adequate to prevent recurrence. This item is close Closed) Violation (149/31022-Ola and q1hlQRf)l: Non-licensed plant operators failed to use existing plant procedures while shifting instrument air afterfilters and while performing an equalizing charge on the Division I station battery. The errors made while trying to perform the evolutions from memory had the _
potential for a reactor scra The inspectors reviewed the applicable licensee documentation and concluded that corrective actions for the violations appeared reasonable and adequate to prevent recurrence of the specific events. However, as noted in the letter transmitting Inspection Report (IR) 50-440/92002(DRP), parsonnel errors had been a continuing concern at the Perry plant. Since IR 50-440/92002(DRP)
was issued, personnel errors have continued to occur. Exampics included a reactor scram due to an improperly installed oil gasket on a reactor feed pump turbine and both trains of the standby liquid control system being inadvertently isolated. Both events were documented in IR 50-440/92020(DRP). Although the overall number of events caused by personnel error have decreased somewhat since 1991, continued effort in this area is still warranted, The adequsty of the licensee's efforts to reduce personnel errors will continue to be evaluated in future inspection report This item is close _
No violations or deviations were identifie . Ucensee Lvent Report (LER) rollowun (9911L_12700)
Through review of records, the following event reports were reviewed to determine if reportability requirements were fulfilled, immediate corrective actions were accomplished in accordance with technical specifications (TS), and corrective action to prevent recurrence had been established: (Closed) LER 50-440131038-03: On April 15, 1992, control room operators discovered that the "A" residual heat removal outboard containment isolation valve, IE12-F0027A, had been opened and deenergized for approximately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> without the actions of TS 3.6.4 being taken. At the time of the event, the plant was in operational condition 5 (REf0EL) with core alterations in progres )
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Licensee Inysstication of Root Cause and Corrective Actions Root Cause As discussed in the subject LER, valve IE12-f0027A was locally opened as part of a tagout restoration. However, the motor control center (MCC) for the valve was not re-energized immediately. With the MCC deenergized the control room did not- .4 have remote valve position indication. A control room shift-turnover caused a suspension of the tagout restoration activitie When the valve's MCC was re-energized approximately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> later, the control room operators identified that the valve was open and
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closed the valve. The root cause of the event was an inadequate ,
procedure. The guidance in plant administrative. procedure ,
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(pap-1401), " Safety Tagging," was not adequate to ensure the proper " returned condition" for valve IE12-f0027 Corrective Action As immediate corrective action, valve IE12-f0027A was closed from the control room. Long term corrective actions were to issue a standing instruction to all licensed and non-licensed operators that required repositioning of MOVs from the control room during i tagout restoration. Procedure PAP-1401 was also revised to require repositioning of MOVs from the control room during tagout restoratio Inspectors Review The inspectors reviewed the applicable licensee documentation -an ;
concluded that corrective actions for the subject LER appeared -
reasonable and adequate to prevent recurrence. Technical specification 3.6.4 required, in part, that with one or more i containment isolation valves inoperable, maintain at least one isolation valve in each affected per.etration operable and within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> isolate the affected penetration by the use of-at least one deactivated automatic valve secured in the isolated positio Otherwise, suspend all operations involving core alteration During the approximate 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> that valve lE12-f0027A was-open, core alterations were in progress. The licensee's failure to '
isolate the affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> while core-alterations were in orogress was a violation of TS 3.6.4. This-violation was not cited because- the licensee's efforts in- - '
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identifying and correcting the. violation met the criteria .
specified in Section Vll.B of the " General Statement of Policy and Procedure for NRC Enforcement Actions," (Enforcement Policy, 10 CFR Part 2, Appendix C (1992)). This item is closed, (Closed) LER 50-440/92010-00; On April 30, 1992, while in operational condition 5, surveillance instruction (SVI)
SVI-821-T1402, " Reactor Water Cleanup Isolation Logic System l Functional Test," was commenced as part of an Instrumentation and 5 l
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Controls (!&C) work order to replace control relay Specifically, SVI-B21-T1402 was used as part of work order 91-3120
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to place the plant in the appropriate condition for relay replacement by inserting a trip signal on the "A" isolation channel. In accordance with the work order, technicians then removed relay 1821-K0148D for replacement, which unexpectedly tripped the "D" isolation channel. This satisfied the nuclear steam supply shutoff system (NSSSS) logic for a balance of plant-(BOP), outboard, containment isolation. Valves which subsequently closed included those in the fuel pool cooling and cleanup system which caused a loss of shutdown cooling. Operators took appropriate actions and restored decay heat removal within 15 minutes. No increase in reactor water temperature was nuted.
i licensee Investication of Root Cause and Corrective Actions Root Cause:
The licensee determined the root cause for this event was en inadequate work order. Both the 1&C planner who drafted the-work order and the I&C supervisor who reviewed the work order failed to recognize that the IB21-K01480 relay was in a channel that would cause a BOP isolation if remove Corrective Action To prevent recurrence, 150 personnel were trained on this event with emphasis placed on the importance of attentian to detail in all aspects of work order preparation and review._ As part of the established requalification training program, all plantilicensed-operators were instructed on the lessons learned from this even Inspectors Review The inspectors reviewed applicable licensee documentation and noted that all corrective action commitments were completed. The inspectors concluded that the corrective actions appeared adequate and reasonable to prevent recurrence.- Appendix B of 10 CFR Part 50, Criterion V, " Instructions, Procedures,-and Drawings," required, in part, that activities affecting quality be prescribed by dccumented instructions of a type appropriate to the circumstances and be accomplished in accordance with these instructions. Contrary-to the-above, the licensee failed to
. ensure that work order 91-3120 was properly written to replace-relay IB21-K01480. This was a violation of 10 CFR Part 50, Appendix B, Criterion V. This violation was not cited because the licensee's efforts in identifying and correcting the violation met the criteria specified in'Section Vll B of the " General Statement of Policy and Procedure for NRC Enforcement Actions," (Enforcement Policy,.10 CFR Part 2, Appendix C (1992)). This item is close i
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' (00en) LER 50-440/9202(L-Mi On October 23, 1992, an inadequate retest of the lower containment inner air lock door seal pneumatic '
. system pressure switch resulted in a violation of TS requirement Licensee Investication of Root Cause and Corrective Actions Root Cause The licensee determined the.cause of this event was a drawing discrepancy which resulted in the wrong inner air lock door seal pneumatic system being tested on October 24, 1992. The )iping system diagram erroneously identified the pressure switcles associated with the upper and lower containment air lock door- seal pneumatic systems. A contributing cause-of the event was personnel error, inattention to detail. ' An alternate responsible e system engineer prescribed the retest without realizing the impact that the maintenance on the pressure switch had on the integrity of the lower containment inner air lock door seal pneumatic system-pressure boundar Inspectors Review Initial investigation of this event was documented in Inspection Report 50-440/92022, dated December 10, 1992. During this inspection period the inspectors reviewed licensee documentation, discussed the event with licensee management and reviewed corrective action plans. On December 18, 1992, the licensee identified an additional drawing discrepancy related to the containment air locks. While troubleshooting the upper air lock inner door to determine the cause for a seal on the door not depressurizing, the system engineer identified that the test
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connections to the small and large seals were _ incorrectly identified on the piping system diagram. As a result of this discrepancy, the potential existed to te:t the incorrect sea The licensee initiated action to determine if the condition of the r seals on all' of the doors were indeterminate at any time following initial plant startu In addition, a walkdown of all the air .
lock doors was to be conducted to identify any additional drawing _ :
discrepancies. Pending the inspectors review of the' licensee's
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investigation results and corrective actions, this LER will remain ope (Closed) LER 50-440/92021-00: On November 1,_199_2, during a plant startup, the licensee discovered that the average power range monitor (APRM) gain and channel calibration was not completed as required by TSs within 12-hours after exceeding 25 percent reactor power. On October 31, with the plant in operational condition-1 (POWER OPERATION) at 24.4 percent power, the licensee completed SVI-C51-T0024, "APRM Gain and Channel Calibration," at 8:12 p.m.,
using feedwater pump inlet feedwater flow data. Subsequently, power was raised to greater than 25 percent at-approximatel :00 a.m. on November 1. ~At 2:48 a.m. preparations were commenced
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to reperform the surveillance but difficulties were encountered with the indicated flow received from the feedwater venturi . Plant precedures required that feedwater flow for the heat balance be obtained from the venturis when greater than 25 percent reactor powe Consequently, SVI-C51-10024 could not be performed as written. At 3:45 p.m. the oncoming unit supervisor noted that the surveillance had not been performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of reaching 25 percent power. As a result, the provisions of 15 4.0.3 were applied, allowing the surveillance to be completed satisfactorily within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, following a power increase to 52 percent, SVI-C51-10024 was completed at 10:30 Lknaste Invesligation of Rontlanstandlemstire Actions RORllM10 -
The licensee determined the root cause of this event was multiple personnel errors. Inadequate communication, failure to follow procedure, and inattention to detail all contributed to missing this surveillance. feedwater flow indication problems, which interfered with obtaining a satisfactory power calculation, diverted the operators' attention from the requirement to complete this surveillance within the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> time limit, proper turnover and proper use of the potential limiting condition for operation (LCO) tracing system could have prevented this even Contributing to this event were procedural inadequacie (orrectiye Ardign To prevent recurrence the licensee counseled all personnel involved in this event on the importance of proper communication and proper use of the LCO tracking system. In addition, all licensed operators were to review this event as part of requalification trainin The operating and surveillance -
instructions involved in this event were to be revised to include specific time limitations for completion of the ApRM calibratio IDspMARD_EEifW As documented in IR 50-440/92022(DRP), the inspectors previously evaluated the event and the licensee's immediate corrective actions. During this inspection period, the inspectors reviewed licensee documentation of the event, including investigation results and long term corrective actions. As noted in paragraph 2.b., the issue of personnel errors is still of concer The licensee has implemented a program to trend and reduce personnel errors; however, the corrective actions for previous personnel errors have apparently not been fully effective. Technical specification Table 4.3.1.1-1, footnote (d), required that APRM channels be calibrated to conform to the power values calculated through heat balance during operational condition 1 when thermal power was greater than or equal to 25 percent powe l l
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bNa ..y, the provisions of TS 4.0.4 were not applicable, ;
,.rovideo the surveillance was performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of
. reaching 25 percent power. Technical specification 4. .
prevented entry into an operational condition unless the '
surveillance requirements associated with the LC0 had been performed within the applicable surveillance interva Contrary to the above, on November 1, 1992, after reaching 25 percent reuctor power at 1:00 a.m., the APRM gain and channel calibration was not completed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> which resulted in a violation of TS 4.0.4 and TS Table 4.3.1.1-1, footnote (d). This is a vioiation (50-440/92024-01(DRP)). (Closed) LER 50-440/92023-00: On November 8, 1992, while in operational condition 1, control room operators identified a discrepancy between the two wide range containment pressure channels displayed on the Emergency Response Information System (ERIS). (This event was previously discussed in IR 50-440/92022(DRP)). On November 10 the licensee determined that containment pressure transmitter D23-N270A was inoperable due to an improper valve lineup. The pressure transmitter and instrument '
loop for D23-N270A were satisfactorily calibrated and D23-N270A was returned to service on November 11. Containment pressure transmitter D23-N270A was inoperable from March 26, 1992, until *
November 11 without the actions of TS 3.3.7.5, Table 1, being
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take Licensee Investiaation of Root Cause and Corrective Actions Root Cause:
The licensee determined that the root causes of this event were an inadequate SVI and inattention to detail on the part of.the I&C technicians who returned D23-N270A to service. The steps in SVI-D23-T2002, " Containment Atmosphere Monitoring Isolation Valves Seat Leakage and Position Indication Test," that isolated and later restored pressure transmitter D23-N270A to service were inadequate because they did not specify the pressure transmitter sensing line isolation valve. Those steps were not in conformance with PAP-0517, " Preparation of Technical Specification Surveillance Instructions," which required that SVis "should be written to stand alone." The I&C technicians that restored D23-N270A to service had additional drawings in the field that identified the correct isolation valve and the drawings and valves-
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were properly labeled. Nevertheless, the I&C technician opened the instrument manifold test valve instead of the instrument sensing line manifold isolation valve, which resulted in the instrument remaining isolate Corrective Actions .
To prevent recurrence, the licensee initiated action to revise the surveillance instruction to explicitly identify applicable
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instrument valves and develop procedural guidance to ensure
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applicable instrumentation is in service prior to operation condition changes. In addition, the technician who restored the transmitter to service on March 26, 1992, was counseled and lessons learned from this event were to be reviewed by all I&C technicians, supervisors, and surveillance writer l!LPfC1911 RevicW in IR 50-440/92022(DRP), the inspectors documented initial event occurrence. During this inspection period, the inspectors reviewed licensee documentation of the event, specifically the investigation results and corrective actions, in addition, the inspectors reviewed drawings and procedures and conducted a walkdown of the affected instrumentation. The inspectors concluded that the licensee's investigation appeared thorough in reviewing the event and adequate in attributing the cause to a combination of inadequate procedures and inattention to detai As noted previously, the corrective actions for previous personnel errors have not been fully effectiv Technical specification Table 3.3.7.5-1, Action 80-a, required, in part, that with the number of operable accident monitoring containment pressure instrumentation channels less than two, restore the inoperable channel to operable status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHU100WN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Contrary to the above, from June 9, 1992 until November 10, 1992, with the exception of some short periods in which the plant was shut down for repairs, the plant was in operational conditions 1, 2, or 3 with less than the two required channels of accident monitoring containment pressure instrumentation, without the required actions of TS Table 3.3.7.5-1, Action 80-a, being taken. This is a Violation (50-440/92024-02(DRP)).
No deviations were identified; however, two violations and two non-cited violations (NCVs) were identifie . Monthly Surveillance _0jnervations (0123)
For the surveillance activities listed below, the inspectors verified one or more of the following: testing was performed in accordance with procedures; test instrumentation was calibrated; limiting conditions for operation were met; removal and restoration of the affected components were properly accomplished; test results conformed with technical specifications, procedure requirements, and were reviewed by personnel other than the individual directing the test; and any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personne __m_____ .
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lurveillanct Activity Title SVI-Cll-T0223A Setpoint Channel A Calibration for ICll-f4054 SVI-E22-T1319 Diesel Generator Start and Load Division 3 fio violations or deviations were identifie . tignthly Maintrnance Observation (62703)
Station maintenance activities of safety-related systems and components listed below were observed and/or reviewed to ascertain that activities were conducted in accordance with approved procedures, regulatory guides _
and industry codes or standards, and in conformance with technical specification The following items were considered during this review: the limiting conditions for operations were met while components or systems were removed from service, approvals were obtained prior to initiating the work, activities were accomplished using approved procedures and were inspected as applicable, functional testing and/or calibrations were performed prior to returning components or systems to service, quality control records were maintained, activities were accomplished by qualified personnel, parts and materials used were properly certified, radiological controls were implemented, and fire prevention controls were implemente Work requests were reviewed to determine status of outstanding jobs and to assure that priority was assigned to safety-related equipment maintenance which may affect system performanc During the inspection period, the ir.spectors noted two examples of -
maintenance activities that were performed well. On December 1, 1992, the licensee replaced the Division 1 diesel generator left bank number 1 cylinder gasket following identification of a jacket water leak. The maintenance activity was planned for approximately 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> but was completed in approximately 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />, in planning for this repair, the licensee contacted other plants for information on repairing the leak, and used a Unit 2 diesel generator as a mockup for training the maintenance crews. As a result of incorporating lessons learned from other plants and the use of the mockup, the time that the diesel generator was inoperable was minimized. The second example involved the repair of the residual heat removal "B (RHR-B) room cooler on December 5-16, 1992. On December 5, the RHR-B room cooler f ailed due to a fan belt coming off and failure of the fan bearing and housing. The room cooler was subsequently repaired on December 6, by replacing the damaged components with parts obtained from Unit 2. Due to the prompt corrective maintenance effort, the licensee minimized the time that the RHR-B loop was not operabl Il
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Specific Maintenance Activities Observed: ,
. Work Order (WO)/ Repetitive Task N Title WO-920005064 Division 11 Diesel Generator Fuel Injector Sticks WO-R85-9795 Control Room HVAC Return Fan "B" WO-910004014 Replace Unit-1, Division-2 Battery WO-R85-00286 IP450130A Limitorque Operator Maintenance per PMI-0030 WO-92-03641 Install Lube Oil Reservoir Drain Valve, Condensate Booster Pump C No violations or deviations were. identifie ,
6. Enaineered Safety features System Walkdowns (71710)
In addition to routine observations made during regular plant tours, the inspectors conducted walkdowns of the accessible portions of selected safety-related systems. During this inspection period the inspectors conducted a walkdown of the high pressure core spray (HPCS) syste The inspectors verified system operability through reviews of valve lineups, system prints, equipment conditions, and control room indication As a result of the walkdowns, the inspectors noted that the general ,
condition of the HPCS system was good. The system was aligned in accordance with the appropriate valve lineup sheet. the HPCS pump and valve rooms were well lit, and the components were properly labele In the HPCS pump room, several general housekeeping deficiencies were identified:
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Debris under the HPCS waterleg pump ski Loose parts on the HPCS waterleg pump ski There was a small packing leak on HPCS condensate storage tank suction valve, IE22-F001. A catch basin was under the valve to direct an leakage to a-drain and a WO had been prev _iously written to repair the leak. Additionally, there was a slight amount of oil dripping ~ from the manual actuator of the HPCS pump discharge check valve bypass valve, 1E22-F026. A maintenance work request was written to investigate and l
repair the lea l
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In the HpCS valve room, several general housekeeping deficiencies were identified:
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Several tools laying on piping insulation and on pipe support Debris on floor including rags, plastic tie-wraps, and tap Drain hoses laying on the floo The inspectors noted that the deficiencies were of relatively little safety significance. However, they were indications of inadequate post-maintenance cleanup, inattention to detail, and overall poor housekeeping 3ractices. Upon being notified of the deficiencies by the inspectors, t1e licensee took action to correct the' deficiencies. The general material condition of the plant is further discussed in paragraph 7.b. of this report. The inspectors-concluded that the observed condition of the HPCS system appeared adequate to support performance of the systems intended safety functio No violations or deviations were identifie . Operational Safety Verification (71707)
The inspectors observed control room operations, reviewed applicable logs, and conducted discussions with control room operators during this inspection period. The inspectors verified the operability of selected emergency systems, reviewed tagout records, and verified tracking of limiting conditions for operation associated with affected. component Tours of the pump houses, control complex, the intermediate, auxiliary, reactor, radwaste, and turbine buildings were conducted to observe plant equipment conditions including potential fire hazards, fluid leaks, and excessive vibrations, and to verify that maintenance requests had been initiated for certain pieces of equipment in need of maintenance. The inspectors by observation and direct interview verified that the physical security plan was being implemented in accordance with the station security pla The inspectors observed plant housekeeping, general plant cleanliness conditions, and verified implementation of radiation protection:
controls. In addition, the inspectors observed construction of the low level radioactive waste building, Reactor Fuel Leak-E During the course of the current operating cycle, an increase in coolant activity had been noted due to leaking reactor fuel. In
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September 1992, the dose equivalent iodine (1-131) concentration was .005 microcuries per gram. At the end of the inspection period activity levels measured were approximately .012 microcuries per gram. Technical specification 3.4.5a requires specific activity of the primary coolant to be-limited to less than or' equal to 0.2 microcuries per gram dose equivalent 1-13 l l
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As the coolant activity increased during the operating cycle, the licensee increased sampling frequency and performed two flux tilts to locali:e the location of the leaking fue The flux tilt involved the sequential insertion and withdrawal of control rods in various locations of the reactor core while monitoring for changes in offgas system pretreat activity. As a result of the flux tilts, the fuel leak was localized to cne fuel bundl To suppress the release rate, control rods near the leaking bundle were inserted. Coolant activity levels initially decreased following the control rod insertions, but eventually continued to increase. Concurrent with increased coolant activity, plant dose rates also increased which impacted personnel accessibility to the plant. As a result of the effect on the plant by the fuel bundle leeks, the licensee announced on November 30, 1992, plans to perform a mid-cycle outage commencing January 8, 1993. During the mid-cycle outage, the leaking fuel bundle were to be remove The inspectors evaluated licensee actions in response to the fuel leak, including the licensee's monitoring of plant areas for changing radiological conditions due to increased coolant activity. The inspectors concluded that licensee efforts appear to be conservative in minimizing the effects of the fuel leak on offsite and onsite radiological activity level I Plant Material Conditions Based on inspectors observations of plant areas, a decline in the general housekeeping and condition of plant equipment was noted during this inspection perio Specific deficiencies noted include:
Plant Area Discrepancies Reactor Core Isolation -
Debris below floor grating:
Pump Room: tools, scaffolding material, plastic bags, face shield, insulation material laying around loos Heater Bay: -
Reactor feed pump turbines (RFPT) A and B have numerous oil leaks of approximately g al s . / da Insulation not installed on B RFPT contribute to high room temperatur RFPT B has tools laying around <
loos .I
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Low Pressure Core Spray -
Tools, hoses, plastic bags Pump Room, "C" Residual laying around loose below
, Heat Removal pump room, grating, and High Pressure Core Spray Pump and Valve Rooms:
The above deficiencies indicate continued poor post maintenance cleanup practices and inattention to detail. The inspectors discussed these deficiencies with licensee management and action was taken to correct the problems. The inspectors previously documented the status of the housekeeping and material condition of the plant in IR 50-440/92002(ORP) dated March '3, 1992, 50-440/92012(DRP) dated July 22, 1992, and 50-440/92022(DRP) dated December 10, 1992. Though the decline in housekeeping by itself s was not of safety significance, it was an indicator of a lack of-attention to detail and poor maintenance practices. In addition,
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the impact poor cleanliness practices had on plant operations was
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demonstrated during the plant restart from the third refueling outage. As previously documented in IR 50-440/92012(ORP),several plant transients resulted from post maintenance debris contributing to the clogging of the hotwell pump suction strainers. Appendix B of 10 CFR Part 50, Criterion II, Quality Assurance Program required, in part, that activities affecting quality shall be accomplished under suitable controlled conditions. Controlled conditions include the use of appropriate equipment, suitable environmental conditions for accomplishing the activity, such as adeauate cleanliness, and assurance that all prerequisites for the given activities have-been satisfie Contrary to the above, based on the inspectors observation of debris below the grating in the reactor core isolation cooling pump room and observation of oil leaks and debris in the reactor feed pump turbine rooms, the licensee failed to maintain adequate cleanliness. This is a violation (50-440/92024-03(DRP)). Trainina Observations During the report period, the inspectors attended the licensee's general employee training and radiological controls training (RCT). For the training observed, the inspectors noted that pertinent course material was available to each trainee and classroom lectures were provided by knowledgeable licensee -
, personnel. Of note was the practical exercise required for i successful completion of RCT training which included proper
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donning and removal of protective clothing. Based on the observations noted above,-the inspectors concluded that the training provided was well planned and useful for the attendees.
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! Residual Heat Removal "B" Test Valve Failure I
! On December 22, 1992, at 6:29 a.m., while securing from the ,
suppression pool cooling mode on the RHR-B system, valve E12-l F0248, RHR-B Test Valve to suppression pool, lost power while
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being stroked closed. Subsequent investigation determined that ,
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two mainline fuses to the valve motor were blown. The valve is an
. 18 inch motor operated gate valve, with three 12 amp mainline fuses installed to protect the valve moto The licensee entered the TS 3.6.3.3 LC0 action requirements for the "B" suppression pool cooling loop bein9' inoperable. The LC0 requires restoration of the loop to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT SHUTDOWN within the next 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> .
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The licensee initiated troubleshooting of the valve to determine the cause for the failure. Various possible causes were reviewed including:
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valve wedge being cocked
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crud in seat of valve
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motor electrical fault
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inadequate stem lubrication
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bent stem
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limit switch degradation / settin guide rib wear /high spots / nicks "
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fuses degraded -
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valve mechanism degradation As a result of evaluating the above possible causes, the licensee concluded that the most likely contributors.to the event were valve degradation due to wear, coupled with some fuse degradation over time with valve usage. Based on successful' stroking of the valve during troubleshooting and assessment of degradation of the valve, the' licensee evaluated-the valve as operable on December 24, 1992,.at 9:52 a.m. To minimize wear, administrative controls were placed on the usage of the valve until the January 1993 maintenance outage. During the outage, testing will be performed on the valve to further evaluate the physical condition of the valve. In addition, the licensee initiated plans to evaluate the possibility of replacing the valve with a mlobe valve during the outag The inspectors, including NRC regional and headquarters staff
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personnel, reviewed the licensee's investigation of the valve failure and discusi,ed-the results with the plant engineering
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staff. Based on that review, the inspectors concluded the licensee's efforts appeared thorough and reasonable and in
. r cordance with T One violation concerning inadequate cleanliness was identified; no deviations were identifie . Onsite followup of Events at Donatina Power Reactors (93702) General The inspectors performed onsite followup activities for events which occurred during the inspection period. Followup inspection included one or more of the following: reviews of operating logs, procedures, and condition reports; direct observation of licensee actions; and interviews of licensee personnel. For each event, the inspectors reviewed one or more of the following: the sequence of actions, the functioning of safety systems required by plant conditions, licensee actions to verify consistency with plant procedures and license conditions, and verification of the nature of the event. Additionally, in some cases, the inspectors verified that the licensee's investigation identified root causes of equipment malfunctions and/or personnel errors and the licensee was taking or had taken appropriate corrective actions. Details of the events and licensee corrective actions noted during the inspector's followup are provided below, Detailt (1) Unexpec_ted Safety Relief Valve Actuation On November 21, 1992, at 3:57 a.m., with the reactor operating at 99 percent power, two safety relief valves (SRVs) opened for approximately 1 minute resulting in a plant transien On November 21, at approximately 12:00 a.m., during a panel walkdown, a plant operator identified that SRV Reactor Pressure Low Low Set trip units IB21-N0617B and 1821-N06180 were in the tripped condition. The tripped units were declared inoperable and an investigation was initiated to determine the cause. While performing a step in surveillance procedure SVI-B21-T0369-B, "SRV Pressure Actuation Channel B Functional for IB21-N66BB," as part of the troubleshooting effort, the two SRVs opene Specifically, as the technician pulled the Calibration Select /r.ommand switch on the calibration unit to verify that it was pulled out in accordance with the surveillance procedurs. SRVs IB21-F0051C and IB21-F0051D opene Plant operators responded in accordance with Off Normal Inst,"' tion ONI-B21-1, "SRV Inadvertent Opening / Stuck Open,"
and a tempts were made to close the valves. Upon resetting
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the B" Low Low Set logic, the SRVs close The SRVs were opened for approximately 1 minute and 12 seconds, with no
. change in containment parameters being noted. Upon SRV closing, reactor pressure increased slightly resulting in a power spike to 102.6 percent reactor power. At 3:58 power was reduced due to the power spike exceeding rated thermal power. Following engineering evaluation of the effect on the reactor vessel by the two SRV actuations and determination that no damage occurred, reactor power was restored to 99 percent at 4:59 At approximately 7:02 a.ni, the licensee informed the NRC operations center of this event via the Emergency Notification System (ENS).
The licensee's initial troubleshooting efforts were directed at recreation of conditions present at the time the event occurred. During this troubleshooting the problem was not duplicate Subsecuent inspection of the power supply, individual sp mocules, circuit cards, and instrument racks also did not identify any potential problems that could have caused the event. The licensee replaced trip units IB21-N6178 and ID21-N6188 and the calibration unit, though no specific problems were 4dentified with these units. The licensee preliminarily concluded that the likely cause of the SRV actuation was an unidentified electrical perturbation coupled with the effected trip units being apparently more susceptible to noise due to possible circuit component degradation over time. The licensee's investigation of the causes for this event was continuin The trip unit vendor was involved with the troubleshooting effort The inspectors reviewed plant and operators' response to the SRV actuation and monitored licensee troubleshooting efforts. The inspectors concluded that the operators responded in accordance with plant procedures and plant response was as expecte Licensee initial investigation efforts appeared thoroug The licensee initiated condition report (CR)92-271 to document the results of their investigation into the cause of this event and corrective actions taken, in a ' ion, LER 50-440/92024 was submitted on December 21, 195 a accordance with 10 CFR 50.73. The inspectors will review that report in a future inspection perio (2) Oil Spill On December 8, 1992, at 2:50 p.m. the licensee discovered an oil leak in the "A" main lube oil heat exchanger to the service water system. Upon review of lube oil usage data, approximately 700 gallons of main turbine lubricating oil was unaccounted for and had possibly been discharged to Lake
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Erie via the service water system over an 11 week perio Subsequently, the "A" heat exchanger was sampled for oil
. intrusion with no leaks identifie The licensee initiated CR-92-279 to document the event and track corrective actions. The licensee notified local and state authorities of the potential oil release in accordance with PAP-0806,
"0il/ Chemical Release Contingency Plan." On December 8, 1992 at 3:22 p.m. the licensee informed the NRC Operations Center of the event via the EN No violations or deviations were identifie f Evaluation of licensee Self-Assessment Canability(40500) On-Site Review Committee During the. report period, the inspectors observed an on-site'
review committee meeting to evaluate that organizations effectiveness, for the meeting attended, the inspectors considered the following attributes: the degree of plant management involvement and/or domination of discussions; if constructive discussion occurred; if the majority of the committee consistently voted the same as the chairperson; if the committee was biased toward operation or safety;-and, if the committee used design basis, the Updated Safety Analysis Report, or vendor technical manuals for their determinations in addition to the T In preparation for the meeting,.the inspectors reviewed the draft submittals given to the on-site review committee for approva *
Items presented to the on-site review committee included safety evaluations, temporary changes to procedures, setpoint change requests, procedural revisions, and design change package During this report period, the following on-site review committee meeting was observed by the inspectors:
Metina N Da /03/92 for the meeting observed, the inspectors concluded that the function of the on-site review committee was effectively implemented, Offsite Review Committee
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During this inspection period the inspectors reviewed the
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licensee's offsite review committee activities which were
- performed by the nuclear safety review committee (NSRC). To
! det . sine if the functions of the congnittee were being performed in accordance with regulatory requirements, the inspectors reviewed licensee documentation governing the composition, duties--
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and responsibilities of the NSRC, including section 6.5.2 of the TS. The inspectors reviewed previous NFRC meeting minutes and
. attended an NSRC meeting to evaluate the effectiveness of~the committee to provide an independent review and audit of plant activitie On December 17, 1992, the inspectors attended the quarterly NSRC meeting. The members were well qualified and prepared to perform the committee reviews. The quorum, composition, and function of the NSRC was in compliance with TS requirements. The NSRC meeting included reviews of various subconmittee reports, including Audit and Quality Assurance, Operations and Maintenance, and Engineering. Discussions were also held concerning planned mid-cycle outage activities and the impact of the fuel leak cn plant operatio The inspectors concluded that the NSRC was objective and effective ,
in the review of plant activities and that the TS requirements for the committee were me No violations or deviations were identifie . Manaaement Chances The licensee announced the selection of the Training Manager, Mr. Dave Igyarto as plant manager effective December 7,1992. Mr. Igyarto succeeded Mr. Robert Stratman, who vacated the position of plant manager to assume the duties of Vice President - Nuclear, upon Mr. Mike Lyster's selection as Vice President a! the Dresden Nuclear Power Statio . Manaaement Egeting NRC management met with licensee management on December 17, 1992, at the NRC Region III office in Glen Ellyn, Illinois. Personnel attending the meeting are designated by (#) in paragraph 1 of this report. The purpose of the meeting was to discuss recent plant issues including-plant material condition and scooe of the January 1993 outage. At the '
conclusion of the meeting, the NRC management acknowledged the licensee's efforts and planned activitie . Items for Which a " Notice of Violation" Will Not Be-Issued l
i During this inspection, ceitain activities, as described above in paragraph 3.a and 3.b, appeared to be in violation of NRC requirement However, the-licensee identified these violations and they will not be cited because the criteria specified in Section VII.B of the " General Statement of Policy and Procedure for NRC Enforcement Actions,"
(Enforcement Policy, 10 CFR Part 2, Appendix C, (1992)), were satisried.
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1 fyit Interviews b The inspectors met with the licensee-representatives denoted in ;
paragraph I throughout the inspection period and on December 28, 199 l
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The inspectors summarized the scope and results of the inspection and discussed the likely content of the inspection report. The licensee did ,
not indicate that any of the information disclosed during the inspection could be considered proprietary in natur ,
During the report period, the inspectors attended the following exit i interview: i Insocctor Exit Date S. Orth 12/10/92
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