IR 05000346/1999008

Insp Rept 50-346/99-08 on 990513-0628.Three Violations Noted & Being Treated as non-cited Violations.Major Areas Inspected:Aspects of Licensee Operations,Maint,Engineering & Plant Support

ML20210C446
Person / Time
Site:	Davis Besse
Issue date:	07/20/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20210C443	List: IR 05000346/1999008 ML20210C438
References
50-346-99-08, NUDOCS 9907260065
Download: ML20210C446 (14)

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U. S. NUCLEAR REGULATORY COMMISSION I

REGION 111 Docket No: 50-346 License No: NPF-3 Report No: 50-346/99008(DRP)

Licensee: Toledo Edison Company Facility: Davis-Besse Nuclear Power Station Location: 5501 N. State Route 2 Oak Harbor, OH 43449-9760 Dates: May 13 - June 22,1999 Inspectors: K. Zellers, Senior Resident inspector S. Dupont, Project Engineer Approved by: Thomas J. Kozak, Chief Reactor Projects Branch 4 Division of Reactor Projects f

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9907260065 990720 l PDR ADOCK 05000346 i 9 PDR j

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EXECUTIVE SUMMARY Davis-Besse Nuclear Power Sta?!on NRC Inspection Report 50-346/99008(DRP)

This report included observations and conclusions pertaining to aspects of licensee operations, maintenance, engineering, and plant support and covers a 6-week period of resident inspectio Operations

• The inspectors concluded that the plant was operated in a conservative, risk-informed manner (Section 01.1).

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The inspectors determined that the station review board effectively performed its duties for the items discussed during the meetir,0 (Section 07.1).

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The inspectors determined that the issues discussed in the Quality Trend Summary Report served as an awareness tool to help communicate collectively important adverse trends to the line organization (Section 07.2).

Maintenance

- The inspectors noted opportunities existed to improve station personnel calibrEtion of the reactor protection system. The governing procedure did not obtain recorded data for the nuclear instrumentation upper and lower detector gain potentiometers, but rather relied on as-found setpoints. Also, no verification that the potentiometers provide the desired output was made at the conclusion of the procedure (Section M3.1).

• The inspectors concluded that the licensee did not perform Technical Specification surveillance testing on the Steam and Feedwater Rupture Control System within the required periodicity prior to heating up the plant to mode 3 during the 11* refueling outage that was conducted April-May 1998. This was a Non-Cited Violation (Section M8.1).

Enaineerina

- The unidentified reactor coolant system leak rate approached the Technical Specification limit of 1 gallon per minute prior to the recently completed maintenance outage. The leak rate was effectively reduced during the outage (Section 02.1).

- Subsequent to the outage, low flow rates have been routinely occurring in the !

containment atmosphere particulate and gaseous radiation monitoring system. The ]

plant staff %s been aggressive in attempting to identify the reasons for this ;

phenomer.n, but the frequent filter changes required to address the low flow conditions have been a distraction to plant personnel (Section E2.2).

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• The inspectors concluded that the licensee unknowingly exceeded the acceptance criterion for decay heat valve pit enclosure leakage in 1983, because the surveillance

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procedure that provided instructions for the conduct of the test had nonconservative acceptance criteria. This was repcrted in a licensee event report and resulted in a Non-Cited Violation (Section E8.1).

• Safe shutdown earthquake seismic load calculations failed to consider the nozzle loads on the decay heat coolers. This was a Non-Cited Violation (Section E8.2).

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Report Details Summarv of Plant Status The plant was operated at about 100 percent power throughout the inspection perio l. Operations 01 Conduct of Operations 011 General Comments (71707)

Operators managed the plant risk for proposed maintenance activities by observing the requirements of the on-line risk administrative programs. Station management was made aware of and responded to materialissues consistent with their potential safety consequence. Operators were aware of plant material status and conducted pre-evolution briefs as necessary in order to convey important operational information to the operating crews. Shift tumovers were done in an orderly fashion and changes in plant status were effectively communicated to oncoming crews. Administrative activities were conducted in accordance with goveming procedures. The inspectors concluded that the plant was operated in a conservative, risk informed manne O2 Operational Status of Facilities and Equipment O2.1 System Walkdowns (71707)

The inspectors walked down the accessible portions of the following engineered safety features and important-to-safety systems during the inspection period:

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Motor Driven Feed Pump

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Emergency Ventilation System

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Containment Spray System

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Low Pressure injection System

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High Pressure Injection System

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Emergency Diesel Generators and Station Blackout Diesel Generator

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Safety-Related invertors and Rectifiers No substantive concems were identified during the walkdowns. Equipment material condition was excellent in all cases. However, some minor deficiencies were noted such as dirty floor drain strainers in the emergency diesel generator (EDG) rooms', a rag stuffed into a station public address system speaker, and a seismic support pin snap ring that was missing on an auxiliary feedwater steam '.ne. These items were communicated to licensee personnel and handled in a manner consistent with their safety significance.

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! 07 Quality Assurance in Operations 07.1 Station Review Board (71707)

l The inspectors observed station management conduct a station raview board meeting on June 16,1999. The Technical Specification (TS) administra%ve requirements pertaining to board composition were met. Members were well-prepared for the

meeting. Discrepancies with the quality of documentation presented for review y ere l retumed to the originator for correction or upgrading. Members were knowledgeable of l station requirements and the corrective action process. The inspectors determined that l the station review board effectively performed its duties for the items discussed during the meetin .2 Quality Trend Summarv Report (71707)

The inspectors reviewed a quality trend summary report that was generated on June 10,1999, by station nuclear assurance personnel and attended a subsequent station management meeting during which the results of the report were discusse The purpose of the report was to condense station quality assessment findings (e.g., from condition reports, audits, observations, assessments) into common themes to enable the line organization to address themes in a timely manner. The inspectors

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l determined that the problems described in the report were handled relative with their significance and that the characterization of performance was generally consistent with the inspectors' assessment. The issues discussed in the Quality Trend Summary Report served as an awareness tool to help communicate collectively important adverse l trends to the line organizatio Miscellaneous Operations issues (92700)

08.1 (Closed) Licensee Event Report (LER) 50-346/96-008-02: Switchyard Circuit inoperable Due to Switchyard Breaker Alignment. This update clarified information provided in the original LER regarding revisions to switchyard operating procedures. No additional review of this information is required. The original LER was closed out in Inspection Report 50-346/97009 as a Non-Cited Violation. This LER is close .2 (Closed) LER 50-346/98-006-00: On June 24,1998, a tornado damaged the electrical j distribution switchyard which resulted in a loss of offsite power, reactor trip, and turbine i trip. An NRC team inspection was conducted to evaluate the damage to the plant, the f l

licensee's response, and corrective actions taken. The results of that inspection were J documented in inspection Reports 50-346/98012 and 50-346/98013. The inspectors i determined that the description of the event in the LER was similar to the NRC

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inspection team's findings. No further action is required and this LER is closed.

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, 11. Maintenance

'M1 Conduct of Maintenance M1.1 Maintenance and Surveillance Activities (61726. 62707)

The following maintenance and surveillance testing activities were observed / reviewed during the inspection period:

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_ DB-MI-03058 RPS Channel Testing

=: DB-SS-03091 Motor Driven Feed Pump Quarterly a

DB-MI-03014 CRD Breaker C Testing

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MWO-99-000102-000 Engrave Dipstick of EDG 2

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DB-SC-03071 EDG 2 Monthly Test

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DB-MI-03205 Channel Functional Test / Calibration and Response Time of RCP Monitor (RC 3601) to Steam and Feedwater Rupture Control System (SFRCS) Logic Channel 1 and RPS CH 1 The inspectors concluded that engineering and maintenance personnel effectively conducted the test and maintenance activities listed above. Specific observations are discussed belo M3 Maintenance Procedures and Documentation M3.1 Nuclear Instrumentation Imbalance Inocerable l Insoection Scope (61726) )

- The inspectors reviewed the results of a reactor protection system calibration activit Qbservations and Findinos On May 27, Instrumentation and Control personnel performed Procedure DB-Mi-03057,

"RPS Channel 1 Calibration of Overpower, Power / Imbalance Flow, and Power / Pumps Trip Functions." Portions of this procedure require the manipulation of the NI-6 Ic ver detector gain potentiometer. At the end of the procedure, the vemler on the potentiometer was adjusted to the as-found setpoint in accordance with the procedur Subsequently, operations personnel noted during a panel walkdown for turnover, that NI-6 imbalance indicated approximatelyzero on the plant computer and on the control room meters. Station personnel then performed surveillance tests for imbalance and

. determined that the axial offset error was in excess of the acceptance criteri Operators then entered the appropriate TS actions. Subsequently NI-6 was recalibrated, retumed to service, and declared operabl After the calibration was completed, the vemler for the potentiometer for the lower detector gain adjust was at the same spot as before, which potentially indicated that the l potentiometer output was momentarily effected by a piece of dust or dirt that was removed when the potentiometer was subsequently manipulated. The inspectors questioned why the output of the potentiometer was not verified prior to test completion to ensure that it performed according to expectationc. Licensee personnel stated that

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the potentiometer is normally very reliable and that there was not an easy method to verify the amplifier output. The inspectors determined that the potentiometer setting was not recorded, but rather the setpoint was maintained by recording in the as-found position before potentiometer manipulation for testing so that it could be retumed to the as-found condition upon completion of the testin c.- Conclusions -

. The inspectors noted opportunities to improve station personnel calibration of the

. reactor protection system. The goveming procedure did not require technicians to obtain recorded data for the nuclear instrumentation upper and lower tetector gain potentiometers, but rather relied on as-found setpoints. Also, no verification that the potentiometers provide the desired output was made at the conclusion of the procedur M8 ' Miscellaneous Maintenance issues (92700).

M8.1 ' (Closed) LER 50-346/98-003-00: On May 19,1998, after the plant was placed in operation mode 3 during start up from the refueling outage, the licensee discovered that TS surveillance requirements were not met. Instrumentation and control technicians incorrectly determined that portions of three of the four surveillance tests for reactor coolant pump rnonitor circuit inputs to the SFRCS, did not need to be performed. Since the surveillance requirements were not met before the plant entered mooe 3 on May 18, 1998, the plant was in a condition prohibited by the TS. When the condition was discovered on May 19, the licensee declared the protection systems inoperable and subsequently successfully completed the surveillance tests. The licensee determined that the root cause of the event was that the surveillance test procedures were inadequate because they did not specify which portions were required to be completed

- after a refueling outage. The licensee then revised the surveillance test procedures (DB-Ml-03205 through DB-Ml-03208, reactor coolant pump monitor channel functional test) so that the procedures specified the portion of the test that was required to be performed subsequent to a maintenance activity or a refueling outag Technical Specification 6.8.1.c, Procedures and Programs, states, in part, that written procedures shall be established, implemented and maintWned covering surveillance and test activities of safety-related equipment. Technical Specification surveillance

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requirement 4.3.2.2.1 states, in part, that each SFRCS instrumentation channel shall be demonstrated operable by the performance of the channel functional test at least every 31 days during modes 1,2, and 3. Technical Specification surveillance requirement 4.0.4 states, in part, that entry into an operational mode or other specified applicable condition shall not be made unless the surveillance requirements associated with the limiting condition for operation have been performed within the stated surveillance interval or as otherwise specified. The proper performance of Procedures DB-MI-03205, DB-MI-3206, and DB-MI-03208 would have satisfied the ,

requirement to conduct TS 4.3.2.2.1 surveillance testing for SFRCS channels 1,3,  !

. and 4, respectively.- Contrary to this, on May 18,1998, the licensee did not properly complete Procedures DB-MI-03205, DB-MI-3206, and DB-MI-03208, in order to satisfy  :

TS surveillance requirements 4.3.2.2.1 for SFRCS channels 1,3, and 4 within the '

required surveillance interval prior to entering mode 3. ' This Severity Level IV violation is ;

being treated as a Non-Cited Violation, consistent with Appendix C of the NRC .

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Enforcement Policy. This violation is in the licensee's corrective action program as LER 98-003-00 (NCV 50-346199008-01). j

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M8.2 (Closed) LER 50-346/97-008-14f Inadequate Testing of Safety-Related Circuits. This update clarified root causes and provided updates on corrective action efforts. This LER was previously closed as a Non-Cited Violation in inspection Report 50-346/9800 M8.3 ' (Closed) LER 50-346/97-011-00 - On August 25,'1997, maintenance personnel rendered the emergency ventilation system (EVS) inoperable. Maintenance personnel

performed scheduled preventive maintenance on an emergency core cooling system room ventilation isolation valve. The maintenance activity required maintenance personnel to block open the isolation valve and open an adjacent ventilation access door to inspect the valve. Maintenance personnel completed the inspection and returned the valve to service and closed the access door. During the maintenance, the

valve and the ventilation access door were simultaneously open for about 15 minute On September 2,1997, station system engineers and licensing personnel reviewed the completed maintenance activity and concluded that with the isolation valve and access

' door simultaneously open, both trains of the EVS were inoperable because the hole in I

the pressure boundary of the EVS was sufficiently large enough to prevent the EVS from drawing down the pressure to those values assumed in the Updated Safety Analysis Report. The licensee also determined that in the event of a design basis accident, the condition could have resulted in an unfiltered fission product release path to the environment. Because of the short duration of the event (15 minutes), no TS

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Limiting Condition of Operation was exceeded. The licensee determined that the 1 preventive maintenance activity procedure was inadequate because it did not specify that the activity should only be performed when the TSs did not require that the EVS be q operable (modes 5 and 6). As a corrective action, the licensee changed the )

- maintenance procedure to restrict its use to when the EVS was not required by TS . This TS procedure violation had no actual and little or no potential impact on safety, was

an isolated problem, and would not have presented a more significant safety and

. regulatory concem if the procedure would have been left uncorrected. This failure constitutes a violation of minor significance and is not subject to. formal enforcement action.-

111. Engineering E2- Engineering Support of Facilities and Equipment E Reactor Coolant System (RCS) Uniderdiried Leakaoe Inspection Scope (37551)

The inspectors reviewed the licensee's efforts to resolve higher than normal unidentified :

i RCS leakag ; Observations and Findinos l During past operating cycles, the RCS unidentified leak rate was normally less than ;

.1 gallon per minute (gpm); however, after the most recent refueling outage, the unidentified leakage rate trended up to about .8 gpm. The RCS leakage caused containment atmosphere boric acid particulate concentrations to increase as evidenced ,

by the accumulation of boric acid on the containment air coolers which required their i periodic cleanin ;

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The licensee initiated a maintenance outage prior to its scheduled start date to address the RCS leakage before the TS limit of 1 gpm unidentified leakage was reached. The major contributor to the leakage was determined to be two leaking pressurizer code safety valves. Additionally, an inspection plan was executed to determine other sources of RCS leakage. Troubleshooting efforts resulted in the RCS unidentified leakage being reduced to approximately .3 gpm. The licensee continued efforts through a detailed action plan to identify leakage sources and reduce the leak rate to previously observed values. The action plan included the following steps:

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Station personnel conducted several containment walkdowns to identify leakag .

~ Engineering personnel used portable acoustic instruments in an attempt to identify any high pressure leak acoustic signature =

Engineering personnel attempted to correlate CAC boric acid fouling to an RCS leak rate in containmen .

Chemistry personnel calculated estimates of the RCS leak in containment from

' the activity level of the CAC/ containment sump wate * Engineering, Chemistry, and Radiation Protection personnel monitored miscellaneous tank and sump levels to determine if any inter-system leakage existed.-

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Engineering personnel planned to integrate the results of the containment atmosphere sample skid action plan and containment air cooler action pla *

Performance engineering personnel'used portable instrumentation to quantify known interfacing system valve leakage more accuratel Radiation Protection personnel performed a noble gas air sample program

. outside of containment to locate leak *

Operators verified that interfacing system boundary valves to the RCS were close l The above listed efforts somewhat reduced unidentified leakage and demonstrated that the majority of the remaining RCS leakage was in containmen Conclusions The unidentified reactor coolant system leak rate approached the TS limit of 1 gallon per minute prior to the recently completed maintenance outage. The leak rate was effectively reduced during the outag E2.2 RCS Leakaae Detection System Problems a. - Inspection Scope (37551)

The inspectors evaluated licensee efforts to resolve performance issues with portions of 1 the RCS leakage detection syste j

! Observations and Findinas  !

The RCS leakage detection ' system consists of three monitoring systems: (1) the containment atmospheric particulate radioactivity monitoring system, (2) the containment atmosphere gaseous radioactivity monitoring system, and (3) the g

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containment sump level and flow monitoring system. Containment atmospheric particulate and gaseous radioactivity monitoring is achieved by continuously pumping containment atmosphere samples through two independent skid mounted monitoring systems.' The skids have process flow particulate and charcoal filters that have to be

- replaced on a periodic basis in order to maintain an acceptable air flow through the ski The TSs require that with less than two of the above leakage detection systems operable, the licensee has 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to restore two of the systems to an operable status or to be in hot standby in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and cold shutdown within the following

- 30 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> Prior to the mid-cycle outage, the RCS leakage caused containment atmosphere boric acid particulate concentrations to increase as evidenced by the accumulation of boric acid on the containment air coolers which required their periodic cleaning. The increase in containment atmosphere particulate concentration also caused the RCS leakage detection system filters to clog at a faster rate than normal which resulted in low skid air flow conditions prior to the periodic (monthly) filter replacement. When a low flow condition occurs on a skid, both the particulate and gaseous radioactivity monitors on

- that skid become inoperable until the filters are replaced to restore air flow to norma Filter replacen:ents ut, be performed in less than an hou Subsequent to the outage, particulates in the containment air have caused the filters on the skids to clog on nearly a ' daily basis. The licensee initiated an action plan which included initiatives such as purging the system of water moisture with nitrogen, calibrating the flow controller, verifying the low flow alarm setpoint, verifying the correct filter paper, checking the pump capacity, checking the calibration work done during the mid-cycle outage, inspecting the flow element for water / foreign material, replacing the flow control valve filter, checking sensor resistance readings, and analyzing the particulate filter residue As of the end of the inspection period, portions of the action plan were implemented with no noticeable benefit towards the performance of the skid On occasion, when routine maintenance caused one of the skids to become inoperable and the remaining operable skid received a low flow alarm, two of the three leakage detection systems were inoperable which caused the station to enter the 6-hour action statement. In all cases, the licensee restored one inoperable skid to an operable status f

before unit shutdown was require Conclusions Subsequent to the outage, low flow rates have been routinely occurring in the containment atmosphere particulate and gaseous radiation monitoring system. The plant staff has been aggressive in attempting to identify the reasons for this r phenomenon, but the frequent filter changes required to address the low flow conditions

' have been a distraction to plant personne E8 Miscellaneous Engineering issues (92903,92700) ]

E8.1. (Closed) LER 50-346/97-007-00 and 01: The licensee identified that the acceptance criterion of an 18-month periodicity TS surveillance test to measure the leakage into the decay heat valve pit was nonconservative. The leskage calculation that determined the acceptance criterion assumed that leakage into the pit was limited by a sealed cover over the pit. However, an additional potentialleakage path through the side wall

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gaskets at the bottom of the pit was identified. The licensee subsequently corrected the ,

acceptance criterion of the surveillance test and reviewed previous test data against the i revised criterion and determined that the surveillance test performed on September 10,

1983, did not meet the revised criterio Technical Specification Section 6.8.1.c, Procedures and Programs states, in part, that

~ written procedures shall be established, implemented and maintained covering surveillance and test activities of safety-related equipment. Technical Specification surveillance requirement 4.5.2.f required a vacuum leak rate test be performed on the decay heat valve pit watertight er, closure every 18 months. The performance of Procedure DB-SP-03125, " Decay Heat Valve Pit Test," was used to satisfy TS surveillance requirement 4.5.2.f requirements. Contrary to this, prior to September 4, ;

1997, Procedure DB-SP-03125 was inadequate in that it specified a valve pit leakage acceptance criterion that could have allowed the decay heat pit leakage to be -

unknowingly above the requirements of the TSs. This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC ;

Enforcement Policy. This violation is in the licensee's corrective action program as LER 97-007 (NCV 50-346/99008-02).

E8.2 (Closed) LER 50-346/97-012-00: On September 4,1997, the licensee identified that it could not be assured that the decay heat coolers (DHCs) would function during a safe shutdown earthquake.. The DHC support system would exhibit localized yleiding during a safe shutdown earthquake because of an inadequate design. The inadequate design was due, in part, to a failure to consider DHC nozzle loads in the DHC seismic qualification report stress load calculations. As an interim corrective action, engineering personnel determined that bolts on the cooler stand needed to be tightened to a higher torque rating. The licensee declared the decay heat coolers inoperable and entered TS 3.0.3. The licensee tightened the bolts to the new values and exited TS 3. Subsequently, a long term corrective action to modify the DHC support systems to make them more sturdy was complete Criterion ill to Appendix B to 10 CFR Part 50, " Design Control," states, in part, that ;

measures shall be established to assure that applicable regulatory requirements and the design basis for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. Updated Safety Analysis Report Sections 3.9.2.2.b and 3.2.1 stated, in part, that the DHCs were designed to withstand the maximum possible earthquake (safe ,

i shutdown earthquake) condition. Contrary to this, prior to September 4,1997, the licensee failed to assure that the DHCs would withstand the safe shutdown earthquake by failing to consider DHC nozzle loads in the DHC seismic qualification report. This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcemer't Policy. This violation is in the licensee's corrective action program as LER 97-012 (NCV 50-346/99008-03).

E8.3 '(Closed) Insoection Follow-uo tieni 50-346/97201-03: The NRC identified that some safety-related electrical components in the emergency core cooling rooms were not included in the station's environmental qualification master list. These components included sump pump motors, sump level switches, local control hand switches, and electrical boxes. The NRC determined that failure of these components could affect the operability of the emergency core cooling systems. The licensee immediately  ;

performed an operability evaluation and determined that the emergency core cooling ]

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I systems were operable. This conclusion was based on engineering judgment, because l identical components in other systems were qualified. The licensee subsequently tested and verified the qualification of each component. The licensee completed its verification on November 23,- 1998. The inspectors reviewed the verification documentation and concluded that this issue is close V. Manaaement Meetinas-

'X1 Exit Meeting Summary

- The inspectors presented the inspection results to members of licensee management at the .

conclusion of the inspection on June 22,1999. The licensee acknowledged the findings presented. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identifie )

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PARTIAL LtST OF PERSONS CONTACTED Licensee M. C. Beier, Manager, Quality' Assessment T. J. Chambers, Supervisor, Supervisor, Quality Assurance S. A. Coakley, Manager, Work Management L. M. Dohrmann, Manager, Quality Services D. L. Eshelman, Manager, Operations J. L. Freets, Manager, Regulatory Affairs P. R. Hess, Manager, Supply D. H. Lockwood, Supervisor, Compliance A. A. McAllister, Supervisor, Test / Performance J. L. Michaelis, Manager, Maintenance C. A. Price, Manager, Business Services G. A. Skeel, Manager, Securit A. R. Stallard, Acting Superintendent, Operations H. W. Stevens, Jr., Manager, Nuclear Safety & Inspections F. L. Swanger, Manager, Design Basis Engineering NRC K. S. Zellers, Resident inspector, Davis-Besse

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INSPECTION PROCEDURES USED

IP 37551

Onsite Engineering .

- IP 61726: Surveillance Observations

' IP 62707: Maintenance Observation IP 71707: Plant Operations

' - IP 92700: Onsite Follow-up of Written Reports of Nonroutine Events at Power Reactor i Facilities {

. IP 92903: - Followup - Engineering

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LITEMS OPENED, CLOSED, AND DISCUSSED Opened-

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50-346/99008-01 .NCV inadequate TS Surveillance procedure 50-346/99008-02 ~ NCV Inadequate TS Surveillance procedure 50-346/99008-03 NCV Violatien of 10 CFR Part 50 Appendix B, I

Design Control' {

Closed 50-346/96008-02- LER Switchyard circuit inoperable due to switchyard breaker alignment

.50-346/97007-00,01 LER Acceptance criteria of 18-month periodicity TS surveillance test to measure leakage into the decay Heat valve pit was

. nonconservative j 50-346/97008-14 .LER Inadequate testing of safety-related circuits  ;

50-346/97011-00 LER ' Maintenance personnel rendered the EVS inoperable 4 50-346/97012-00 LER error with decay heat coolers seismic design calculations 50-346/97201-03' IFl Safety-related electrical components in emergency core cooling rooms were not included in station's environmental

. qualification master list

- 50-346/98003-00' LER TS surveillance requirements not satisfied after plant entered Operation Mode 5 while starting up

- 50-346/98006-00 LER Tornado damaged electrical. distribution switchyard from refueling outage 50-346/99008-01 NCV Inadequate TS Surveillance procedure 3

. 50-346/99008-02 NCV Inadequate TS Surveillance procedure ;j 50-346/99008-0 NCV Violation of 10 CFR Part 50 Appendix B, 1 Design Control  !

Discussed  ;

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