IR 05000298/1997007

Insp Rept 50-298/97-07 on 970629-0809.Violations Noted.Major Areas Inspected:Licensee Operations,Engineering,Maint & Plant Support

ML20216C686
Person / Time
Site:	Cooper
Issue date:	09/04/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20216C620	List: IR 05000298/1997007 ML20216C615 ML20216C632
References
50-298-97-07, 50-298-97-7, NUDOCS 9709090119
Download: ML20216C686 (37)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket No(s).: 50-298 License No(s).: DPR 46 Report No., 50 298/97-07 Licensee: Nebraska Public Power District Facility: Cooper Nuclear Station Location: P.O. Box 98 Brownville, Nebraska '

Dates: June 29 through August 9,1997 Inspectors: Mary Miller, Senior Resident inspector Chris Skinner, Resident inspector Linda Smith, Reactor Engineer Charles Marschall, Project Engineer Approved By: Elmo Collins, Chief, Branch C Division of Reactor Projects Attechment: Supplemental Information 9709090119 970904 PDR ADOCK 05000298 0 PDR

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EXECUTIVE SUM. MARY Cooper Nuclear Station NRC Inspection Report 50 298/97-07 9 05 51120%

The inspectors observed that operatort promptly and appropriately identified condittof.s r vienna crMary conteinment inoperable and appropriately initiated and

) complead a plaat shut down. Crew commano and control were effective and well I

coordinated. Srv.down activities were well;ontrolled. E >ntrol room logs of the shutdown nre excellent (Scotion 01.1).

The startup activities observed were conducted in an offective, well controlled mai aer. Inspectors observed sustained implementation of methodical communications, anticipation and evaluation of plant response, procedural usage, cre;f workload management, and oversight by quality assurance and plant maiiagement (Section 01.2).

The inspectort 8dentiised that the licensee failed to implement adequate corrective actions in respanse to Violation 50 298/96013-05 to prevent the recurrence of failing to follow Procedure 4.15 when placing radiation monitors in service (Sections 02.1,0 *

Based on inspectors' questions, the licensee identified three examples where operators failed to futiy implement procedural requirements. The licensee had implemented interim corrective actions and issued problem identification reports (PIRs) to determine the cause of and initiate long term corrective actions for each issue (Section 04.2).

The inspectors identified several examples of operability assessments which did not identify or address all of the relevant design functions of the degraded componen The incomplete operability assessments required additional evaluation by the plant staff to demonstrate component operability, in one case, the licensee changed the design and licensing basis of a system to provide a basis for an operability determination (Section 07.1).

Inspectors dentified four examples in which the licensee did not identify, correct, determine the caus6, or prevent recuirence of conditions adverse to quality, in three cases, plant staff addressed failed tests by repeating the test or changing the test conditions to obtain satisf actory results (Sections 07.2, E2.1.b.1).

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The inspectors concluded that licensee management had not provided plant staff with clear guidance for determining whether degraded conditions affected i

operability of plant equipment, in addition, management had not effectively monitored the quahty of operability assessments performed by plant staff ISection 07.3).

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Maintenance

Inspectors identified inadequacies in vacuum breaker lif t force surveillance testing and problem identification. Tolerances for force gage inaccuracies and measurement location inaccuracies were not included in the as found acceptance criteria, a PIR was not initiated to identify and correct this nonconservatism, as-fourd testing of the lif t force was not conducted, and a differential pressure from the torus to the drywell was not considered when lift force measurements were j taken (Section M1.1).

The inspectors determined that the licenseo performed adequate troubleshooting to

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eliminate other possible causes and were able to identify the problem that resulted I in diesel generator oscillations and properly identify a condition adverse to quality regarding past problems (Section M2.1).

The timeliness of scheduling and risk assessment to support an unscheduled outage was excellent. Four hours after initiation of a 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> shutdown action statement, a schedule was issued, including a risk assessment (Section M4.1).

Ennineerina

The inspectors found that the operability evaluation conclusions for seven PIRs were acceptable (Section E2.1).

The licensee inadequately documented the basis for operability of several reactor vessellevelinstruments when the safety-related solenoid valve used to fill the instrument's reference leg was determineci to be inoperable (Section E2.1 *

The inspectors found minor documentation weaknesses, which did not affect the final conclusion, within several safety evaluations (Section E2.1).

The licensee identified incorrect acceptance criteria in a surveillance procedure, failed to promptly correct the procedure prior to the next performance, and used the incorrect surveillance procedure for a valve stroke time test (Section E2.1).

The inspector found several potential weaknesses ;n the bcensee's program for ,

determining when a 10 CFR 50.59 safety evaluation was required. The licensee did j not perform a 10 CFR 50.59 screenino review for non-intent procedure changes, calculation changes, or changes which met an Appendix B screen (Section 3.1).

• A system engineer provided an excellent briefing of a recent plant modification to !

the control room crew, in preparation for reactor startup operations (Section E4.1). l I

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DETAILS I

I LQperations Summarv of Plant Statug The plant operated at full power until July 29,1997, when the plant was shut down due to a failure to close a drywell to suppression pool vacuum breaker. During the shutdown, l

the licensee replaced all 12 vacuum breaker gaskets and satisfactory tested the vacuum i breakers. The licensee returned the plant to full power on August 5,1997, and the plant remained at 100 percent power to the end of the inspection perio Conduct of Operations 01.1 Plant Shutdown in Response to Indications of Eauioment Failure inspection Scope (92701)

The inspectors followed the licensee's activities and actions taken to shut down the

- plant in response to a drywell to suppression pool vacuum breaker failure to close, Observations and Findinns On July 29,1997, during a monthly surveillance, operators found that a drywell-to-suppression pool vacuum breaker (AO NRV 28) failed to close as required. They declared the valve inoperable, tested the drywell to torus differential pressure as required, and found that test results were outside the acceptance criteria. The primary containment was declared inoperable and a controlled plant shutdown'was performed as required by Technical Specifications. The inspectors noted that the operators made prompt and appropriate decisions regarding Valve AO NRV 28 and primary containment operability, commensurate with the safety significance of a degraded condition with the potential to adversely affect containment integrit Inspectors observed the plant shutdown, performed in accordance with Procedure 2.1.4, " Normal Shutdown From Power," Revision 47c1. Crew activities were performed in accordance with procedures, and actions to control plant conditions were clearly communicated to the crew. Operators verified that expected plant responses were obtained after control actions. Briefings were performed at regular intervals, describing expected activities. Plant management

- was present in the control room during the shutdown and provided appropriate standards for operating during control room briefings. Review of control room logs after the shutdown indicated excellent documentation of activities and plant conditions during the shutdow After the manual plant scram in accordance with procedure, Groups 2,3, and 6 isolations occurred in response to a low reactor vessellevel. A report to the NRC was made in accordance with 10 CFR 50.72. This issue will be reviewed during closure activities of the licensee _ event report (LER 97 009, open).

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2-l Conclusiong I

The inspectors observed that operatus promptly and appropriately identified conditions rendering primary containment inoperable and appropriately initiated and completed a plant shut down. Crew command and control were effective and well coordinated. Shutdown activities were well controlled. Control room logs of the shutdown were excellen .2 Plant Startuo from U2 scheduled ShutdowD

Scope (92701)

Inspectors observed control room activities on August 1,1997, during a startup from an unscheduled shutdown, inspectors observed control room operators'

implementation of Procedures 2.1.1.1, " Plant Startup Review and Authorization,"

Revision 10; Procedure 2.1.1.2, " Technical Specifications Pre-Startup Checks,"

Revision 13; Procedure 2.1.1, "Startup Procedure," Revision 73c3; and Procedure 2.0.1.1, " Conduct of infrequently Performed Tests or Evolutions,"

Revision 3c Observstions and Findinog During plant startup activities, operators followed these applicable procedures in a methodical fashion, communicating changes in plant raditions and documenting completion of activities. Crew briefings were conducted at regular intervals, including emphasis on procedural adherence, recognition and verification of expected plant response to operator actions, and clear communications. Expected startup activities were summarized and management expectations were provided regarding safety focus, observance and understanding of plant conditions, procedural adherence, conservative decision-making, and effective crew command and contro During significant portions of two shifts, inspectors observed systematic implementation of proceduras, effective control of crew workload, attentive operator demeanor, prompt recognition and response to plant parameters, proper responses to alarms, clear and pertinent communication of plant conditions to the control room crew, organized briefings performed at intervals appropriate to plant conditions, and significant oversight by plant management and quality assuranc Review of control room logs indicated good lookeeping describing startup activitie During the startup, while not under inspector observation, torus level exceeded the maximum limit allowed by Technical Specifications during high pressure coolant injection system testing. This issue will be addressed during closure of LER 97-012 in a later report (LER 97-012, open).

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3-l l Conclusions The startup activities observed were conducted in an effective, well controlled manner, inspectors observed sustained implementation of methodical communications, anticipation and evaluation of plant response, procedural usage, crew workload management, and oversight by quality assurance and plant managemen Operat!onal Status of Facilities and Equipment l

O2.1 Evoluqtpd Release Point Radiation Monitor Returned to Service in Inoperable Confinuration

inspection Scop _g (717071 The inspectors interviewed operations, radiation protection, and plant management personnel. The inspectors reviewed the Technical Specifications, procedures, control room logs, and PIR 2 1610 Observations and Findinas On July 14,1997, chemistry personnel identified that the elevated release point radiation monitor was out of its proper configuration, which resulted in the monitor tcking a continuous iodine and particulate sample frorn the room instead of the elevated release point flow path. Technical Specification 3.21.A.2.c requires that, with less then the minimum number of radioactive gaseous effluent monitoring instrumentation channels operable, the action shown in Table 3.21.A.2 be take Table 3.21.A.2, Action 29, states that effluent releases via this pathway could continue provided samples are continuously collected with auxiliary sampling equipment. From 4

47 p.m. on July 13,1997, to 10:50 a.m. on July 14, there was no sampling of the elevated release point discharge flow while effluents continued to be released, which did not comply with Technical Specification 3.21.A. The licensee's preliminary investigation into this incident identified that the operator did not perform all the steps of Procedure 4.15, " Elevated Release Point and Building Radiation Monitoring Systeme " Revision 18c1, when placing the radiation monitor in service. While the normal range monitor was out of service, a flexible hose was removed from its suction path to troubleshoot and calibrate the radiation monitor. Later, the operator was inaccurately informed that both the auxiliary sampler and the normal radiation monitor were in service. Based on the inaccurate information that the normal renge sampler was in service, the operator removed the auxiliary sampler from service. The licensee's investigation later identified that the flexible hose was not replaced, which resulted in the room being monitored instead of the elevated release point flo .

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1 in NRC Inspection Report 50 298/96 013, one example of a violation was the i

failure of a technician to perform Section 8.2 prior to Section 8.1 of i

- Procedure 4.15, while placing the radwaste building high range radiation monitor in '

service. This resulted in a radiation monitor which did not sample the ventilation flow stream. The licensee's corrective actions, completed on December 11, '

1996, were ineffective in preventing recurrence of the failure to follow Procedure 4.15 Criterion XVI of Appendix B to 10 CFR Part 50, requires for significant conditions adverse to quality, the causes of the conditions be identified and action taken to prevent recurrence. This example of weak or ineffective corrective action is unresolved (50 298/97007-01), Conclusics The licensee found that the elevated release point radiation monitor had not been properly placed into service. Procedure instructions were not properly implemente Operator Knowledge and Performance 04.1 Operations Failure to Follow Procedures lnspection S.g_goe (71701)

The inspectors reviewed the control room logs and selected PIRs and interviewed shift supervisors, shif t technical engireers, and plant management personnel, Observations and Findinas On August 4,1997, while performing the high pressure coolant injection system surveillance test, the control room received a high suppression pool level alarm. The operators f ailed to perform Steps 2.2 and 2.3 of Procedure 2.3.2.22, " Panel 9 3 - Annunciator 9 3-2," Revision 18.1c Step 2.2 required that the operators terminate activities adding water to the suppression pool, and Step 2.3 required that the operators reduce suppression pool level by discharging suppression pool water to radwaste. A PIR was issued. The licensee implemented interim corrective actions which consisted of revising the alarm response procedure and issuing a night orde The procedure was revised to stop addition of water to the torus under high level conditions, and a night order, issued on August 4, described this event and provided management's expectations to preclude exceeding Technical Specification values and design basis values. Additional evaluation and corrective actions will be followed during review of LER 97-11. An NRC open item (50-298/96026-06) also addresses a contributor to this concern regarding instrument scaling and tolerance . A night order had been issued on May 24,1997, which required, once per shift, switches on Control Room Panels 9 5,9 4, and 9 3 to be verified in

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5-l the proper system configuration and the completion of the verification to be logged in the control room log. On August 6,1997, the inspectors identified that the day shift on August 5 failed to log that the control room panel walkdown was performed. Based on the inspectors' questions, the licensee identified that there were no similar required log entries for July 31 (night shift) and July 29 (day shift). The operators responded by contacting both licensed operators and the control room supervisor, who were on shift that day, to determine if the switch checks were performed. From the interviews !

conducted, the licensee determined that the switch checks were performe The licensee initiated PIR 2-16358. For interim corrective actions the licensee revised, prior to the night shift, the operator's shif t logs, ;

Procedure 6. LOG.601, to include documentation of these walkdowns on the checklis . On August 5,1997, an unexpected half Group 1 isolation on Channel B

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occurred while preparing to perform a surveillance on Channel A. The technician had opened a panel door and informed the control room of the presence of water in the panel. Based on steam pressure indication that no conditions requiring an isolation existed, the operators cleared the half isolation signal and then investigated the cause of the half isolation. The licensee discovered that Pressure Switch MS PS 134B was leaking. The inspectors questicned whether recetting the half isolation signal before ,

determining the cause was appropriate. Based on the inspectors' questions, the licensee identified that Alarm Response Procedure 2.3.2.27, " Panel 9 5 Annunciator 9 5-1," Revision 22c2, Step 2.4, was not followed. Step stated that, if only Channel B trips, determine cause, correct, and reset half Group 1 isolation. The operatots made an assumption concerning the cause of the isolation, reset the half Group 1 isolation, determined the correct cause, implemented the half isolation, and corrected the cause. A night order, issued August 8, reported two instances wherein all of the actions contained within annunciator response procedures were not followed and re-emphasized management's expectation of procedural compliance. Also, PIR 2 22241 was initiated to enter this item into the licensee's corrective action progra The licensee and inspectors identified three examples where operators failed to fully implement procedural requirements. The licensee had implemented interim corrective actions and issued PIRs to determine the cause and initiate long term corrective actions for each issue. The failure to follow the procedures documented in items 2 and 3 above are examples of a violation of Technical Specification 6.3.2.B, which requires that procedures and instructions be established, implemented, and maintained for actions to be taken to correct specific and foreseen potential or actual malfunctions of safety-related systems or components. Failures 2 and 3 constitute a violation of minor significance which is being treated as a noncited violation consistent with Section IV of NRC Enforcement

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Policy 1298/97007 021. The concern associated with exceeding torus water level will be evaluated during review of LER 97 012 and NRC Open item 298/96026 0 Conclusion Based on inspectors' questions, the licensee identified three examples wherein operators f ailed to fully implement procedural requirements. The licensee implemented interim corrective actions and issued PIRs to determine the cause of and initiate long term corrective actions for each issu Quality Assurance in Operations 07.1 Operability Assessments Insoection Scone (71707)

Inspectors reviewed approximately 75 operability assessments to assess time liness and to determine if the licensee had made reasonable determinations of operability based on design and licensing basis information. In addition, the inspectors reviewed licensee guidance for performing operability assessments, Observations and Findinas inspectors reviewed PIRs that identified degraded plant conditions. The inspectors found that plant staff fraquently did not effectively consider design and licensing basis information in assessing operability. For example: As discussed in NRC Inspection Report 50 278/97 05, in April 1997, inspectors found that plant staff used acceptance criteria in Procedure 6.CRD 401, " Control Rod Drive Housing Support inspection,"

Revision Oc1, to verify correct assembly of the control rod drive housing support that differed from the design information contained in the Updated Safety Analysis Report and the Bases for Technical Specification 3.3.8.2. In response to the inspectors' questions documented in PIR 2-14794, plant staff did not identify all the relevant design and licensing basis functions of the control rod drive housing support. Engineering staff identified the design requirement to maintain sufficient clearance between the control rod drive housing and the housing support to prevent vertical contact stresses caused by thermal expansion. The engineers did not, however, identify the design requirement to limit control rod withdrawal on control rod drive housing f ailure to less than a notch or the Technical Specification licensing basis requirement to limit control rod drive withdrawal to less than 3 inches for a control rod drive housing failure. As a result, the engineering staff developed an operability assessment which was nonconservat!ve with respect to the procedure acceptance criteria and the requirements specified in the design and licensing basis. Inspectors concluded that, based on the design and

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licensing basis information in existence at the time of the inspectors'

questions, the licensee had not demonstrated operability of tha control rod drive housing support structure after reassembly. The licensee later i successfully demonstrated that changing the design and licensing basis did not constitute an unreviewed safety question.

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J drive housing support in Procedure 6.CRD.401 is a violation of Criterion lli of

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Appendix B to 10 CFR Part 50, which requires that applicable regulatory requirements and the design basis be correctly translated into specifications, drawings, and procedures. This failure constitutes a violation of minor significance and is being treated as a noncited violation, consistent with Section IV of the NRC Enforcement Policy (50 298/97007 03). The inspectors reviewed PIR 2 22815, which discussed that the time delays for the south scram discharge instrument volume vent and drain valves were found outside administrative limits during surveillance testing. The time delay limits were established to ensure that the inboard valves would close

before the outboard valves to prevent trapping fluid between the valves.

1 The operability assessment discussed only the Technical Specification requirement that the valves close within 30 seconds and that water trapped between the valves would not affect the closing function of the valves.

3 The inspectors reviewed the Updated Safety Analysis Report, Section 5.5.2.', " Scram Discharge Volumes," the licensee's Design Criteria

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Document, " Control Rod Drive System DCD 17," and the licensee's In-l service Testing Basis Document, Revision 1.1. From this review, the inspectors identified that the Design Criteria Documents also addressed the i opening function of the valves. Appendix B of Design Criteria Document 17

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includes performance requirements that state the valves are required to open

or close on demand and pass or isolate flow accordingly. Appendix L of the same document has acceptance criteria that states the alves are required to open or close on demand. The inspector also noted that Emergency Operating Procedure 5.8.3, " Alternate Rod Insertion Methods," Revision 5, provided instruction to manually open the valves in case normal scramming F and the alternate rod insertion system failed to function. These design

requirements were not discussed in the operability assessmen . The inspectors reviewed PIR 2 22314, which discussed configuration control problems with the diesel generator mechanical butterfly valve overspeed shutdown cable / tube clamp mounting. Three tube clamps were used, although the drawing showed four tube clamps. The operability assessment stated that the three tube clamp configuration was acceptable as

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documented in Engineering Judgement 95106. The inspectors reviewed

- Drawing KSv.72 24, Revision 1, and Engineering Judgement 95106, dated November 1,1995. The engineering judgement documented only that three

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8-tube clamps were acceptable for seismic and vibratory loads, while the drawing stated that the tube clamps allowed smooth travel (nonkinking)

movement of the cable. The failure to address the smooth travel function of the tube clamps is another example in which not all design requirements were included in the oprability assessmen . The inspectors reviewed PIR 2-22860, which discussed an unexpected automatic depressurization system cooling interlock alarm being received and reset several times during a surveillance test on Residual Heat Removal Pump A. The operability assessment discussed the reason for the unexpected alarm and the basis for operability of the automatic

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depressurization system. The inspectors identified that the operability ( assessment did not document the effect of the unexpected cycling on the l components' service life. Plant staff revised the operability assessment to l document that the short term degradation had not posed a concern, since there were only two documented times when this condition occurred. The licensee noted, however, that continued occurrences could pose a concer . The inspectors reviewed PIR 2-11319, which discussed Procedures 6.1SWBP.101 and 6.2SWBP.101, "RHR Scrvice Water Booster Pump Flow Test" and " Valve Operability Test," Revisions 2c2 and 2c1. The procedures tested interlocks between the service water booster pumps and their corresponding Residual Heat Removal Heat Exchange Service Water Outlet Valves SW-MO-8.9A(B) and required that operators ensure that the outlet valve closed. The licensee's requirements associated with the term

" ensure" allowed the operator to manually position equipment (close the valve) without recording a discrepancy for acceptance criteria steps. The operability assessment indicated that operators verified the automatic closure of the outlet valve by the absence of alarm Through discussions with operators, the inspector identified that alarms are received during the surveillance and clear when the valves are fully closed (either automatically or with operator action). The operability assessment did not consider that operator action to manually close the valves, allowed by procedure, would not verify automatic valve operation but would satisfy the procedure requirements. As a result, the operability assessment did not identify the potential inability to identify an inoperable automatic valve closure function. The inspector determined, through interviews with the operators, that their practice was to observe automatic valve closure, in response to inspector questions, the plant staff revised the operability assessment to documant that the operators that performed the last surveillance stated that the valves closed automatically. The inspectors concluded that the plant staff performing the operability assessment did not verify that the procedure demonstrated proper function of the valve control The licensee changed the procedure to state " verify" rather than " ensure."

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-9-P 6. The inspectors reviewed PlR 2-05592, which identified that the service water supply valve to Diesel Generator 2 was leaking by the valve seat. The operability assessment indicated that the safety function of the valve was to cpen on the start of a diesel generator; therefore, this leakage is not a safety concern. The operability assessment did not discuss the potential that cold water going though the diesel generator cooling loop when the diesel i

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generator was not running could adversely impact the ability of the emergency diesel generator to start and accept load within the time allowed by Technical Specifications. Addit onally, t

the operability assessment did not address a potential cause of leakage to be the degradation of the valve liner, potentially releasing materials and plugging downstream essential heat

exchangers. The diesel generator jacket water and lube oil subsystems are i required to be maintained above 90 F. Maintenance personnel replaced the valve to address the leakage and inspected the heat exchangers in the outage. The inspectors noted that the operability assessment did not address the requirement of the valve to isolate service water flow when the diesel generator was not in operation. The safety consequences were 4 minimal based on the temperature for the jacket water, lube oil systems were maintained above 90 F, and no foreign material was found in the heat exchangers.

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7. On August 6,1997, the inspectors questioned the operability assessment performed when Reactor Core injection Cooling Valve RCIC-AO 34 failed to close during a surveillance test. The shift supervisor documented in the shift supervisor's log that the valve was inoperable, but the. .ystem remained operable due to a nonessential valve which would perform the isolation function. Based on the inspectors' questions, the licensee identified that two procedures required that the reactor core injection cooling system be

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declared inoperable, Procedure 6.RCIC.102, "RCIC IST and Quarterly Test,"

Revision 4 (if Valve RCIC-AO-34 did not close when Valve RCIC-MO 131,

the steam supply to turbine, opened) and Procedure 6.RCIC.201, "RCIC Power Operated Valve Operability Test," Revision 3c4 (if Valve RCIC-AO-34 did not close in less than or equal to 6 seconds). The inspectors reviewed the licensee's Design Criteria Document, " Reactor Core Isolation Cooling System DCD 18," and the in-service Testing Basis Document. Both documents stated that the safety function of the valve was to close to prevent the release of radioactive materials during accident conditions and seismic event The licensee issued PIR 2-16096 to document this issue and to determine if the Design Criteria Document and the in-service Testing Basis Document were correct with regard w che safety function of the valve. When the valve was repaired and tested 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> later, the system was placed back in its normal configuration. The inspectors concluded that the operability assessment did not address why a nonessential valve could be relied upon to perform the function of an essential valve under all accident condition . - - - .

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- 10- Conclusio n j Operability assessments frequently did not identify or address all of the relevant design functions of the degraded component. The incomplete operability assessments required additional evaluation by the plant staff to determine component operability. In one case, the licensee changed the design and licensing basis of a system to provide a basis for an operability determinatio .2 Corrective Action Associated with Operability Assessments

, Inspection Scoce (717071

While reviewing operability assessments, the inspectors also evaluated whether the licensee took appropriate actions to identify and correct the root cause and extent of degraded conditions.

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' Observations and Findinos The inspectors found several examples of operability assessments where corrective actions did not effectively address the cause and extent of the identified concer For example: Inspectors reviewed PIR 2-15824, initiated in May 1997: " Unable to establish differential pressure between drywell and suppression chamber during SP 6.PC.503 Need to develop rapid pressurization or slight differential pressure between drywell & torus to seat NRVs." The inspectors found that the Updated Safety Analysis Report states: "In order to maintain the valve in the closed position unless the differential pressure of at least 0.1 psi exists, each valve is equipped with a magnetic latch." The -

inspectors concluded that the valve design should ensure closure with no differential pressure across the valve (normal containment conditions during-full power operation). The inspectors reviewed the results of the test and concluded that the licensee missed an opportunity to identify degraded NRV seating surfaces. Rather than determine the root cause of the inability to pass the drywell to suppression pool leak test, plant staff changed the test conditions (raised the torus to-drywell differential pressure to a value closer to the design requirement) to obtain acceptable test results. After changing the procedure, the licensee demonstrated valve operability as specified by Technical Specification 3.7 A. On July 29,1997, Valve AO-NRV-28 failed to close during a monthly surveillance, and the licensee was unable to satisf actorily complete SP 6.PC.503. Plant staff subsequently identified that four vacuum breakers exhibited degraded conditions, including degraded seating surfaces caused by over torquing the seal ring during installation in 1982. Although the licensee concluded that the vacuum breakers would have seated under

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accident conditions, the inspectors concluded that the licensee failed to identify and take corrective action for a degraded condition in May 199 The failure to identify and correct a condition adverse to quality is another example of ineffective corrective actions and is an unresolved item (URI 298/97007-01).

2. The inspectors reviewed PIR 2-10378/CR97-0190, repetitive testing of time delay relays. The PIR identified that, in testing the timing of the emergency diesel generator breaker trip relay, the relays changed state as demanded, but consistently failed to actuate the test equipment (digital timer) used for their timing. This unacceptable condition had been ongoing and had existed since installation acceptance testing was performed. The PlR further stated that instrument and controls technicians were " instructed that the conditions described exist and we are to test the relays until the test equipment is triggered correctly and a discrepancy will be recorded on the surveillance procedure." The operability assessment concluded that the test equipment installation used an inappropriate configuration. Corrective action included revising the procedure to correct installation difficulties and resulted in

" successful" surveillance results. The operability assessment did not address the possibility that previous surveillance results had been invalid, nor did the operability assessment or the PIR address the long-standing instructions to technicians to repeat the test until acceptable results were j achieved. The inspectors concluded that the operability assessment and the associated PIR did not identify the lack of corrective action for long-standing, recurring test failures as a significant condition adverse to quality or take action to correct the cause. This is another example of ineffective problem recognition and corrective actions and is an unresolved item (URI 298/97007-01).

3. Inspectors reviewed PIR 2-13047 that described a 4160 volt circuit breaker antipump test. By design, the breaker has an antipump device with the capability to prevent reclosure after the breaker trips open. During a test of the antipump device on February 28,1997, the breaker reclosed. The technician called the control room and received permission to retest the breaker twice more. During the two retests, the antipump device prevented breaker reclosure as designed. The technician noted the discrepancy on the appropriate attachment and initiated a PIR. The operability assessment initially identified breaker misalignment as a possible cause of the initial failure of the antipump. feature. The operability assessment also noted that the procedure did not identify testing of the antipump feature as part of the acceptance criteria. The initial operability assessment concluded that the identification of a possible cause and the lack of an associated acceptance criteria was a basis for reasonable assurance of operabilit The inspectors considered the identification of a possible cause without supporting evidence and lack of an associated acceptance criteria inadequate

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bases for initial reasonable assurance of operability. The inspectors also considered the further identification of two possible causes, refuted by the craftsman performing the test, as an inadequate basis for concluding that no degraded condition existed. The inspectors conclud7d that the licensee, in effect, accepted repeateo tests as appropriate corrective action for a degraded condition, in response to the inspectors' observation, the licensee stated that the l l breaker supplied electrical protection associated with Service Water Booster Pump D, which did not require the antipump feature, and the lack of operability of the feature did not affect operability of the service water booster pump. The inspectors noted that the operability assessment did not reflect that the antipump feature was not required and that the licensee did not address that no controls had been put in place to prevent use of this specific breaker to supply power for equipment that did require Jerability of the antipump feature. This is another example of failure to take affective action to correct a condition adverse to quality and is an unreso!ved item (URI 298/97007-01).

!' The inspectors reviewed PIR 2-22086, which identified that a licensee calculation contained incorrect assumptions on the source and amount of water going to Sump Z and PIR 2-16251, which identified that the Sump Z-2 pump failed to start after 30 minutes of filling Sump Z. During a loss of offsite power, Sump Z pumps would lose power and the water from the sump would backfill the standby gas treatment drain lines and eventually would degrade the standby gas treatment flow path, The operability assessment for both of these PIRs discussed that standby gas treatment system was operable since augmented offgas was out of service. The operability assessments did not document whether any other sources that added water to Sump Z could contain enough water which, if completely discharged to Sump Z, would make the standby gas treatment system inoperable. Even though not documented, the licensee had initiated efforts to determine if smaller quantities of water could make the standby gas treatment system inoperable. The inspectors considered the lack of documentation a weakness, c. Conclusions inspectors identified four examples where the licensee did not identify, correct, determine the cause, or prevent recurrence of conditions adverse to quality, in three cases, plant staff addressed failed tests by repeating the test or changing the test conditions to obtain satisfactory result .

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-13-07.3 Administrative Control of Operability Assessments a.- Inspection Scone (71707)

The inspectors reviewed procedures and guidance that provided expectations and controls to plant staff for performing operability assessments, Observations and Findinos Procedure 0.5, " Problem identification and Resolution," Attachment 3, contained guidance on operability assessment'.. Inspectors found that the procedure contained substantial information related to operability assessments. For example, it provided different types of examples of operable and inoperable systems, subsystems, and components. The procedure did not, however, provide guidance on how to perform an operability assessment. Operations department expectations, Instruction 7, Attachment K, " Assessment of Operability when Degraded Conditionc are identified," provided guidance on the process for operability assessments and provided management expectations for assessing operability. However, as indicated in Section 07.1, plant staff frequently did not meet the stated expectations. For example, Attachment K stated that plant staff is expected to identify the safety function performed by the degraded system or component and identify the impact of the degraded condition on the identified function. Operability assessments frequently did not identify or address the safety function of systems or component Plant staff had performed an evaluation of operability assessments immediately prior to this inspection. The licensee evaluation noted that Quality Assurance does not evaluate performance with respect to operability assessments. The NRC inspectors further noted that the licensee's staff did perform any recurring, systematic evaluation of operability assessments. In addition, the NRC inspectors noted that-no performance indicators exist for operability assessments. As a result, managers could not effectively monitor the large number (approximately 68), age, integrated effect, or quality of open operability assessment Conclusions The inspectors concluded that plant management had not provided plant staff with clear guidance for determining whether degraded conditions affected the operability of plant equipment, in addition, management had not effectively monitored the quality of operability assessments performed by plant staf Miscellaneous Operations issues 08.1 (Closed) Insoector Followuo item 50 298/97005-02: Inadequate operability assessment of the interlock between the service water booster pumps and the corresponding residual heat removal heat exchanger service water outlet valve. The I

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- 14-inspectors reviewed a number of operability assessments and included this item in that review (Section 07.1 of this report).

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08.2 (Closed) Unresolved item 50-298/97005-01: Control rod drive housing supports procedure acceptance criteria and operability assessments were inadequate. The inspectors reviewed a number of operability assessments and safety evaluations and included this item in that review (Section 07.1 of this report).

08.3 (Closed LER 50-298/95-017 00 and -01 and LER 94-33-00: Safety Relief Valves Found Outside Technical Specification. The root cause of this concern is discussed in LER 298/97-002. These past issues will be reviewed during that evaluation (Open LER 298/97-002).

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08.4 (Closed) Violation 50-298/96013-01: Opening and closing of the secondary containment hatch without procedures. The violation involved a June 19 and 20, 1996i opening and closing the secondary containment hatch without using the procedure for this purpose. Although the licensee addressed the failure of the hatch to be opened by procedure, the corrective action to close the ietch did not use the l- required procedure to close the hatch, but instead performed the surveillance procedure to correct the condition. The steps of the procedure to close the hatch were not performed from the time of the violation until af ter the hatch was opened during the subsequent scheduled outage several months later. The licensee issued a PlR on August 23,1997, to address this issue. The corrective actions for the violation did not implement the procedure to secure the hatch as required. This is another example of failure to take comprehensive corrective actions to correct a condition adverse to quality and is an unresolved item (URI 298/97007-01).

08.5 (Closed) Violation 50-298/96013-05: The inspector reviewed the licensee's response letter, NLS960187, dated October 9,1996, and related PIRs regarding multiple violations of 10 CFR Part 50, Appendix B, Criterion V. These were multiple procedural violations that occurred from April 10 through July 27,1996, regarding:

(1) not performing a step in Procedure 4.15, " Elevated Release Point and Building Radiation Monitoring Systems," (2) not performing required Procedure 6.4.6.4.1,

" Turbine Building Kaman Monitor Calibration," (3) performed required steps of Procedure 7.2.55.1, " Replacement of HCU Accumulator," out of their intended order, and (4) mislogged stop times while performing Procedure 6.1EE.302, "4160v Bus 1F Undervoltage Relay and Relay Timer Functional Test." The inspector reviewed the corrective action taken by the licensee to prevent a recurrence of this violation. The corrective actions included a revision of implementing procedures and counselling of associated personnel. The licensee also initiated a station stand-down on July 25,1996, to discuss these issues and other items important to safety. Section O2.1 of this report documented an example where the corrective actions implemented for Examples 1 and-2 of this violation were ineffective. The remainder of the corrective actions for this violation appeared adequat .

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-15-11. Maintenin_g;e M1 Conduct of Maintenance M 1.1 Surveillance on Torus to-Drvwell Vacuum Breakers Inspection Scope (61726)

Inspectors observed surveillance testing of torus-to-drywell vacuum breaker lift settings. Inspectors reviewed drawings, procedures, control of plant conditions, and licensee response to the inspectors finding Observations and Findinas On July 31,1997, inspectors observed use of Procedure 6.PC.308, "Drywell Pressure Suppression Chamber Vacuum Breaker Calibration and Functional Test,"

i Revision 1c1. A force gage placed against the lower portion of the disk was used to push the valve open. The reading was then compared with the surveillance test acceptance criteria for the required disk forc The inspector noted that maintenance personnel performed the test on each valve in

! accordance with the procedure and in a step-by-step manner. The inspectors noted l that the procedure listed the acceptance criteria as equivalent to the Technical Specifications and USAR lift pressure of 0.1 to 0.5 pounds per square inch differential (psid) pressure. The inspector questioned whether the accuracy of the force gage and the positioning of the force gage had been factored into the acceptance criteria limits. The licensee stated that the accuracy and positioning were not included. The licensee then evaluated the as-found and as left values of the lift force and determined that the safety significance was minimal since the !

instrument accuracy and force gage location tolerance effects were bounded by the as-left values. No PIR was written for this concern. Had as-found values been outside of measurement instrument tolerance, the valve would have met the as-found acceptance criteria and would have been considered operable. The inspector noted that the lack of consideration of instrument and measurement accuracy in plant procedures is an example of the concern documented in Open Item 298/96026-06, which raised the issue that instrument tolerances and scaling did not always appear to be considered in many operational and testing procedure The failure to initiate a PIR to identify the norsconservative acceptance criteria is an

{ example of a failure to identify a condition adverse to quality and is an unresolved item (URI 298/97007-01).

The inspector observed that, when a valve was tested by pushing it open, a significant amount of air blew from the drywell to the torus. The inspector noted that plant conditions had not been set to control the pressure differential between the torus and the drywell, nor did the surveillance test compensate for the force imposed by the torus-to-drywell pressure differential. The licensee responded by

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- 16-I writing a PIR, changing the test procedure to secure drywell ventilation, wEch eliminated the differential pressure, and reperforming the test for the eight valves which had been set while affected by differential pressure. Technical Specification 3.7 A.4.a requires, in part, that all drywell-suppression chamber

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vacuum breakers shall be operable at the 0.5 psid setpoint. The failure to properly implement a Technical Specification test procedure is a violation of Technical

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l Specification 6.3.3.A, which requires that testing of Technical Specification equipment be provided to satisfy operating license requirements (298/97007-05).

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The inspector also noted that the licensee had not conducted as-found testing on the valves to determine as found force values. A PIR was written to address the concern. This issue and the root cause evaluation of the valve failures will be followed in resolution of the associated LER (LER 298/97-011, open), Conclusions inspectors identified inadequacies in vacuum breaker lift force surveillance testing and probiam identification. Tolerances for force gage inaccuracies and measurement location inaccuracies were not included in the as found acceptance criteria, a PIR was not initiated to identify and correct this nonconservatism, as-found testing of the lift force was not conducted, and a differential pressure from the torus to the drywell was not considered when litt force measurements were take M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Failure of Diesel Generator Soeed Controller Insoection Scone (627071 The inspectors followed the licensee's activities concerning the failure of the diesel generator speed controller by observing troubleshooting and testing and interviewing plant personnel, Observations and Findinas On July 15,1997, during Diesel Generator 2 performance of Procedure 6.2DG.101,

" Diesel Generator Monthly Operability Test (Div 2)," Revision 10c1, the diesel generator oscillated between 200 to 400 Kw. The operators attempted to lower the load when the diesel generator output breaker tripped on reverse current. The licensee verified that the systems in Division I were operable and performed Procedure 6.1DG.101, " Diesel Generator Monthly Operability Test (Div 1),"

Revision 9, on Diesel Generator 1. No problems were identified on Diesel Generator _- _ _ _ _ _ _ _ _

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-17-The licensee identified that the speed sensing amphenol connection in the control loop circuit was found to contain magnetic particles and, through troubleshooting, the licensee eliminated other possibilities that could have caused the diesel Generator electrical output to oscillate. The licensee cleaned the connection and reperformed the surveillance test satisfactorily. On July 21, the licensee examined the Diesel Generator 1 speed sensing amphenol connection and found a smaller amount of debris in that connectio An earlier problem with this connection occurred in the 1995 outage, where the connection became loose during postmaintenance testing on Diesel Generator The licensee had used a sealant between the connectors on both diesel generators to prevent the connection from coming loose. The licensee initiated PIR 2-17527 to determine the appropriateness of the earlier corrective actions and to identify the root cause of the source of the magnetic particles. The PIR was still open at the end of the inspection period.

! Conclusion l

The inspectors determined that the licensee performed adequate troubleshooting associated with a diesel generator speed control problem. The licensee eliminated other possible causes, identified the problem that resulted in diesel generator oscillations, and properly identified a condition adverse to quality regarding past problem M4 Maintenance Staff Knowledge and Performance M4.1 Prenaration for Unscheduled _Outaoe Scone (62707)

Inspectors observed preparations for an unscheduled outage on July 29,1997, Observations and Findino2 Inspectors noted that, in response to the July 29,1997,6-hour shutdown action statement initiated at 9:50 a.m., scheduling had prepared and issued a 120-line item schedule for forced outage control of plant conditions, repair, and testing activities by approximately 12 noon. Entry into the hot shutdown condition was not required until 2:35 p.m. Maintenance activities preparing for the shutdown commenced promptly after the shutdown was initiated. A risk assessment of outage activities was prepared before the plant was shut down and was compared with scheduled activities. As a result of the comparison, the licensee rescheduled a surveillance activity. Comprehensive scheduling and risk assessment support was provided throughout the outag _- - - - _ - - - - - - -

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-18- Conclusion The timeliness of scheduling and risk assessment to support an unscheduled outage

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was excellent. Four hours after initiation of a 6-hour shutdown action statement, a schedule was issued, including a risk assessmen M8 Miscellaneous Maintenance issues (92902)

M8.1 (Closed) Violation 50 298/96004-02: Diesel Generator 2 muffler bypass valve inoperable due an unauthorized modification. During the installation of Design Change 93-024, craft personnel installed a three-way solenoid pilot operated valve that had an upward directed exhaust port. The craft personnel installed a J tube on the exhaust port that directed the air flow downward and prevented foreign material from entering the exhaust port without using plant procedures to gain the necessary engineering revisions. The J tube also restricted the exhaust flow, thus preventing the muffler bypass valve from openin The root cause of this violation was the failure of management to effectively communicate the expectations for installation of modifications. Contributing causes were a weakness in field supervision and inadequate training for craft labor and field engineering. The licensee's violation response letter dated May 17,1996, stated that all corrective actions would be completed by November 11,199 On July 17,1997, the licensee identified five problems where the as-built configuration on a diesel generator did not match the design drawing. Both the J tubes addressed in the violation and the five additional problems had been installed during the same design change (DC-93-024). The five additional problems identified were determined by the licensee not to be an operability concern. The corrective actions implemented to identify and correct the violation did not identify these additional problems. The PIR initiated to document these items did not identify that these findings should have been identified and corrected in response to the initial problem. The failure to identify and correct unauthorized modifications associated with Design Change 93-024, a condition adverse to quality, is another example of weak corrective actions and is an unresolved item (URI 50-298/97007-01).

M8.2 (Closed) Violation 50-298/96019-01: Undervoltage Jumper. The inspector reviewed the licensee's response letter NLS960206, dated November 6,1996, and related PIRs regarding two violations of 10 CFR Part 50, Appendix B, Criterion On September 6,1996, the licensee installed jumpers and replaced a relay, per Maintenance Work Order 961398,'. hat affected the operability of the essential second level undervoltage protection circuit for Vital Bus F relay de power. An operability assessment was not performed as required by Procedure 0.5, " Problem identification and Resolution," and Procedure 7.3.16." Low Voltage Test Relay Removal and Installation." The inspector reviewed t.ie corrective action taken by-the licensee to prevent a recurrence of this violation. The corrective actions

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-19-included interim management action to ensure that a review of a part's full contribution to the system or component is documented prior to issuance from the warehouse and revision of Procedure 7.3.16, " Low Voltage Test Relay Removal and Installation," to require a validation that, pending maintenance activity, will not

. influence operable equipment, The inspector determined that the licensee's actions were adequate and closed this violatio M8.3 (Closed) LER 298/97-004: Automatic actuations as a result of improper maintenance work during an outage. During a refueling outage, while level indication was being restored, maintenance technicians implemented an incorrect section of a procedure for reactor vessellevelindication operation, introduced air into the level indication, and caused erroneous level indications. As a result, a reactor scram and group isolation signals occurred, including a loss of shutdown coolin The inspector reviewed the time line associated with resetting the scram, resetting isolations, and restoring shutdown cooling. Although the time taken was longer than typically expected (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to reset the scram,21/2 hours to reset the isolations, and 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> to restore shutdown cooling), the flushing and restoration of the level indication took a significant amount of time since the system was relatively complex, system configuration was called into question and required full verification, and actions to verify and restore the system were required in various parts of the drywell, Reset of the scram took longer, in part, because the scram discharge volume was being modified and drainage of the volume relied on a smaller, temporary line. Also, operators had not performed this procedure very often and took some time implementing the temporary instructions. Based on interviews, operations and maintenance had recognized during the event that, after they performed their immediate actions, they had time to recover from this event (since core heatup was very slow) with a more systematic and thorough approach rather than urgent actions. The safety significance was minimal since core heatup rate was about 6 F per hour, resulting in a rise from 104 to 123 F over a 3-hour perio The failure to properly implement the procedure for restoration of the vessellevel indication system is a violation of Technical Specifications 6.3.2.A, which requires, in part, that written procedures and instructions shall be implemented for normal operation and shutdown of all systems and components involving nuclear safet This licensee-identified and corrected violation is being treated as a noncited violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy (50-298/97007-04).

M8.4 (Ocen) Open item 50-298_L960026-06: Failure to include instrument tolerance and scaling for Technical Specification settings without automatic actuation. The inspector identified that the torus-to-drywell nonreturn valves lift setting, listed in Technical Specifications as less than 0.5 pounds differential pressure, did not I

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-20-include tolerance for instrument accuracy or force gage positioning accuracy. This issue is described in Section M Ill. Ennineerina E2 Engineering Support of Facilities and Equipment E2.1 PIRs insoection Scoce (37551)

The inspector selected eight PIRs which appeared to require engineering support to resolve. The inspector reviewed the operability determinations and related 10 CFR 50.59 safety evaluations to determine whether the licensee's immediate corrective actions addressed appropriate safety function Observations and Findinas Motor 02erator Valve Response Time Failures j t

The inspector reviewed PIR 2-14630, which was initiated on May 4,1997, during the last refueling outage. The licensee had initiated the PIR when, following a modification, Valve RHR-MOV-MO34B failed its stroke time limi The licensee determined that this failure was caused by use of an incorrect test procedure. As a part of the modification process, the licensee had specified a postmodification test to establish a new baseline stroke time and had provided instructions to update the routine surveillance procedure, if required. However, plant personnel incorrectly used the routine surveillance : ;

procedure, which had not been updateil, to perform the test. The failure to properly perform surveillance testing is a violation of Tecnnical Specification 6.3.3.A, which requires, in part, that routine test procedures be provided to satisfy operating license requirs.nents such as routine testing of equipment required by t'ie Technical Specifications (298/97007-05).

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The licensee reviewed the as-found test data and determined that the valve performed within the calculated limits. The inspector reviewed the basis for this determination, Engineering Judgement 96-121 " Change in Stroke Time for RHR-MOV-MO34A and RHR-MOV-MO34B," and the 10 CFR 50.59 safety evaluation for the applicable Modification Package 96 054. The inspector found that the licensee had provided enough information in the two documents and the PIR to conclude that the as-left stroke times were acceptable.

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-21-In the 10 CFR 50.59 safety evaluation, the licensee concluded that the increase in stroke time was acceptable because the new stroke time was within the inservice test maximum stroke time for operability. However, the inspector noted that the licensee had quoted the inservice Test Reference / Acceptance Limit Data File limit for closure of Valves A and B to be 46 and 47 seconds, respectively. Based on the information in the engineering judgement, the inspector determinad that the use of these values as acceptance limits could result in exceeding the analytical time limit for low pressure coolant injec; ion. Additional inspection is planned to determine the purpose of the Inservice Test Reference / Acceptance Limit Data File and to confirm that the licensee has an adequate program for assessing valve operability. This inspection will be tracked as inspection Followup Item 50-298/97007-06.

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In addition, PIR 2-14630 indicated thet Valve RHR MOV-MO348 was a

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primary containment isolation valve. The licensee stated the valve was l

l required to close in less than 90 seconds to meet primary containment isolation requirements. While the test documentation indicated that the valve was capable of closing within 90 seconds, the safety evaluation did not explicitly address whether the new valve stroke time met the containment isolation safety function stroke time requiremen The inspector also reviewed PIR 2-14800, initiated on May 5,1997, during the last refueling outage. The licensee had identified that the current close-stroke operability limits for Valve RCIC-MOV-MOIS, 23 seconds, exceeded the analytical assumption in the High Energy Line Break Analysis, i.e., a 15-second close. The licensee evaluated this design discrepancy in Engineering Evaluation 97-108, " Evaluation of RCIC-MOV-M015 and RCIC-MOV-MOl6 Closing time," and determined that the installed valves were acceptable. They determined that the inservice test operating limits and the snalysis assumptions could acceptably be changed to 20 seconds. Past valve close stroke times were less than 20 second The inspector reviewed Engineering Evaluation 97-108 and the 10 CFR 50.59 safety evaluation associated with the use-as-is dispositio The inspector found a minor technical conflict between the two document The 10 CFR 50.59 safety evaluation stated that the environmental conditions due to a longer valve closure time do not exceed the values of temperature established by the environmental qualification program. However, the engineering evaluation stated that extending the isolation time by 8 seconds would alter the maximum, localized room temperature curve in EQDP-4 The engineering evaluation went on to find this increase in localized room temperature to be acceptable, since no environmentally qualified equipment was in the immediate area of the steam pipe. The inspector determined that the conclusions of the 10 CFR 50.59 safety evaluation were not affected by this discrepanc __

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-2 2-2. Unknown Confiauration of Reactor Vessel Level Reference Lea Iniection Valve Wirina

The inspector reviewed PIR 2 23448, which was initiated on June 27,1997, when the unit was at full power. The licensee had initiated the PIR when ,

they noted that Reference Leg injection Valve NBI-SOV 739 was apparently wired incorrectly. The solenoid valve unexpectedly lost power when a fuse in a separate circuit was pulle While resolving the configuration control discrepancy, the licensee declared the valve inoperable, but determined that the system was still operable. No basis was given for the determination that the system was still operabl The valve was inoperable from June 27,1997, at 7:16 p.m. when the fuse was pulled to July 2,1997, at 2:31 p.m., when the wiring error was correcte Generic Letter 92-04, " Resolution of the issues Related to Reactor Vessel Water Level Instrumentation in BWRs Pursuant to.10 CFR 50.54(f)," dated August 19,1992, and Bulletin 93-03, " Resolution of issues Related to Reactor Vessel Water Level Water Level Instrumentation in BWRs," dated May 28,1993, discussed level instrumentation inaccuracies caused by the accumulation of noncondensibles in instrument reference legs. The NRC staff was concerned that noncondensible gases may become dissolved in the refererce leg which could lead to a false high levelindication during depressurization or cooldow The inspector noted that Valve NBI SOV-739 was the safety-related valve which was used to flush noncondensibles from the reference legs. The inspector noted that not flushing noncondensibles from the cold reference leg -

could result in inaccurate reactor vessel level indication in the control room, premature trip of the reactor core isolation cooling turbine, premature trip of the high pressure coolant injection turbine, incorrect levelinput into the anticipated transient without scram circuitry, incorrect level input into the feedwater control circuitry, incorrect level input into a residual heat removal system interlock, and incorrect input into the reactor protection and primary containment isolation logic. While not all of these circuits may be needed when the accumulation of noncondensible gases is expected to occur, the inspector was concerned that the operability of these instruments had not been carefully evaluate I Subsequent to the completion of the inspection, the licensee provided !

additional information to the inspector related to the design basis safety j function for Valve NBI-SOV 739 and to their plant-specific commitments for -l Generic Letter 92-04 and Bulletin 93-03. The licensee noted that Valve NBI- l SOV-739 was originally credited in their Three Mile Island action plan for i NUREG-0737, item II.F.2, " inadequate Core Cooling instrumentation." The l l

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23-licensee's design bases accidents included scenarios which would result in -

reference leg boiling, which adversely aff! ts level indication. The licensee was still evaluating the safety function to Valve NBI-SOV 739 but, preliminarily, had determined the valve had o safety support function to be available for the operator to manually refill the reference leg if boiling occur The licensee also provided background correspondence for their commitments to Generic Letter 92-04 and Bulletin 93-03. The inspector reviewed this documentation and found that the licensee had originally submitted that the existing configuration was sufficient to address the control of noncondensibles. However, the NRC staff had not found this response to be acceptable. The licensee subsequently installed an additional nonsafety grade continuous reference leg backfill system using the control rod drive system to address the issue of noncondensibles in the reference leg The licensee planned to re-evaluate the impact of Valve NBI-SOV 739 on the operability of the associated level transmitters. This item is unresolved c pending further review of the licensee's revised operability determination (50-298/97007-07) Separation Criteria The inspector reviewed PIR 2 16435, which was initiated on June 26,1997, when the unit was at full power. The licensee had initiated the PIR when they noted that both divisions of electrical power were present in one terminal box. Primary Containment Isolation Valves RHR SOV-SSV60, -61,

-95, and -96 were affected by this conditio While resolving the separation criteria issue, the licensee implemented Clearance Order 97-1293 to maintain the affected valves in a de-energized and isolated condition. This resolved the immediate ccr.tainment integrity concerns. However, the valves were to be operated in the emergency operating procedures. No resolution of that potential requirement was addresse The licensee subsequently initiated Condition Report 97-1224, " Reinstallation of Fuses for PCIS isolation Valves RHR-SOV SSV60-61,95 and 96," and performed a 10 CFR 50.59 safety evaluation to justify a use-as-is disposition for the design weakness. The licensee evaluated the possible failures-introduced by the lack of separation and demonstrated that the following safety functions would still be met: reactor protection, emergency core cooling, safe shutdown during a fire, and limiting the release of radioactive materials to acceptable levels. The inspector reviewed these documents and found them to be acceptabl _ _ . . . _ . _ _ _ _ _ _ _ . _ _ . . e

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24-4. Batterv Room Temneratures The inspector reviewed PIRs 214369 and 2-14518, which were initiated on April 7 and 9,1997, respectively, during the last refueling outage. The licensee had initiated these PIRs when they discovered the battery room temperatures out of specification low. The licensee implemented System Operating Procedure 2.2.38, " Portable heating System," and installed portable heaters in Battery Room 1 A to bring the room temperature back into specification. On PIR 2-14369 they noted that Design Criteria Document 35,

" Station Blackout," required that the battery room temperature be at or above 72.5*F during normal operation. The portable heating procedure instructs operators to place the portable heaters in service when the battery room temperature f alls below 76 The inspector noted that the supervisor review section of the PIR 2 14518 discussed low battery room temperatures as a continuing problem. The inspector was concerned that use of the portable heating equipment was an operator work around and that the licensee was not addressing design deficiencies in the ventilation system. The licensee stated that these specific low battery room temperatures were related to scheouled outage woik, which affected the room temperature. They stated that, under normal conditions, the existing ventilation system adequately maintains the temperatures in the battery rooms. The inspectors noted that, under severe winter conditions (-30 F) during 1996 plant operations, battery room temperatures were low and required monitorin The inspector reviewed the 10 CFR 50.59 safety evaluation for Revision 0 of Procedure 2.2.38 to evaluate the acceptability of using portable heaters to maintain the battery rooms within their temperature specifications. The inspector also reviewed the procedure change notices for other revisions of Procedure 2.2.28, i.e., Revisions 1, 2, 3, 3,1, 4, 5, and Sc The inspecter found that use of the procedure degraded the fire protection features, because doors to the battery rooms were propped open. However, the licensee had adequate plans for establishing a fire watc The inspector found that the licensee had adequately addressed load addition to the safety-related power supply and electricalisolation between the nonsafety portable heaters and the safety-related power supply in Revision However, in Revision 1, the licensee added an alternate safety related power supply but did not discuss electricalisolation requirements. The licensee added a second alternate power supply in Revision 2 and deleted e fuse

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without discussing the impact on electricalisolation requirements. The licensee stated that double fusing was maintained in all configurations to i separate the safety-related circuits from the nonsafety-related portable

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25-heaters. The inspector concluded that, while past safety evaluation i documentation was not thorough, use of this procedure was acceptable for implementation during degraded condition _l L

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. Diesel Fire Pumo Batterv Technical Specification The inspector reviewed PlR 2 14982, identified on June 5,1997. The licensee had initiated this PIR when they noted that a Procedure Change Notice for Surveillance Procedure 6.EE.602, "125v/250v Station and Diesel Fire Pump Battery Quarterly Check," Revision 1, incorrectly revised the acceptance criteria. The acceptance criteria for the specific gravity of the diesel fire pump batteries had not been appropriately corrected for temperatur The inspector reviewed the licensee's operability determination. The

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inspector found that the licensee had appropriately reviewed the test data I from the prior surveillance (dated March 19,1997) and determined that the specific gravity would have met the corrected acceptance criteri However, the inspector found that'the licensee re-performed the test on June 11,1997, without revising Procedure 6.EE.602 to correct the acceptance criteria. The inspector determined that the licensee had not promptly taken effective corrective action for PIR 2-14982 in that they allowed a procedure to be used again without correcting the acceptance criteri The licensee acknowledged the inspector's findings and initiated PIR 218574 to assess current operability of the diesel fire pump batteries and to perform a root cause analysis for the ineffective corrective actio The failure to correct the acceptance criteria in Surveillance Procedure 6.EE.602 prior to the performance on June 11,1997, was another example of ineffective corrective actions and is an unresolved item (URI 298/97007-01).

c. Conclusions The operability evaluation conclusions for seven of the eight PIRs were acceptabl In addition, the licensee inadequately documented the basis for operability of several reactor vessellevel instruments when the safety-related solenoid valve used to fill the instrument's reference leg was determined to be inoperable. This item is unresolved, while the licensee researches their license basis and reperforms this operability determinatio Minor documentation weaknesses, which did not affect the final conclusion, existed in several safety evaluations. However, in one case, the data in the Inservice Test

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26-L Reference / Acceptance Limit Data File for valve stroke times was not conservative-with respect to the valve stroke time performance assumed in the accident analysi _

Further inspection is planried regarding assessment of valve operability with respect to stroke time .

- The licensee identified incorrect acceptance criteria in surveillance procedures and failed to use proper procedures for the next surveillance performanc E3- Engineering Procedures and Documentation E3,1 ' Administrative Controls a, insoection Scoce (37551)

The inspector reviewed the submittals for the proposed, July 29,1997, routine Station Operations Review Committee agenda to evaluate the licensee's program for determining which facility and procedure changes received a 10 CFR 50.59 safety evaluation prior to implementatiore The inspector also reviewed related -

administrative procedure ' Observations and Findinas Nonintent Procedure Chanoes t

The inspector found that the licensee's program for implementing the  ;

requirements of 10 CFR 50.59 included a screening review of procedure-changes. The licensee used the screening review to identify procedure .

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changes which required a full safety evaluation pursuant to 10 CFR 50.59, The inspector ncted that the licensee did not perform a screening review of

- nonintent procedures changes.- The inspector was concerned that, without

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performing a rigorous screening review,-the licensee could approve procedure changes which affected the plant licensing basi After discussions with the inspector, the licensee evaluated some of the proposed nonintent changes listed on the proposed July 29,1997, agend The licensee identified that one of the instructions, which had been submitted as a nonintent change, included the addition of acceptance criteria

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- to meet Technical Specification surveillance requirements. While the procedure change had not been approved by_the Station Operations Review-Committee, the licensee determined that any procedure which added new or recently recognized requirements should have received a safety evaluatio The' licensee noted that they had indirectly defined nonintent changes to be changes which did not affect the plant licensing basis. However, they also agreed that their definition of intent changes needed re evaluation. They

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- initiated PIR 2-17874 to furtner evaluate this issue. Inspectors noted that Violation 50-298/97006 01 addressed a nonintent change which allowed a Technical Specification violatio )

- Lack of 10 CFR 50.59 Fvaluations for Calculation Chanagg The inspector noted that the licensee did not perform 10 CFR 50.59 saf6ty l evaluations for calculation changes. The inspector reviewed the licensee's j l administrative procedure for changing calculations, Engineering l l- Procedure 3.4.7, " Design Calculations," Revision 11c1, The inspector found l that the procedure included a requirement to confirm compliance with the Updated Final Safety Analysis Report and to identify all documents which had been affected by the calculation change. The licensee stated that, following a calculation change, the proposed changes to the affected i documents and the Updated Final Safety Analysis Report would be evaluated pursuant to 10 CFR 50.59 prior to implementatio The inspector concluded that this program met regulatory requirement However, the inspector noted that this approach may not always accomplish the intent of the regulation. The inspector noted that calculation changes offer a clear view of possible reductions in the margin of safety (which reflects the assumed capability of the system). The absence of 10 CFR 50.59 safety evaluations for calculation changes is a missed opportunity to rigorously evaluate possible reductions in the margin of safety. The inspector did not complete performance-based inspection to judge the effectiveness of program implementatio . 10 CFR 50.59 Evaluations for Drawina Chanae Notices The inspector noted that the licensee did not perform 10 CFR 50.59 safety evaluations for all drawing change notices. The inspector interviewed licensee personnel and reviewed Procedure 3.7, " Drawing Change Notices,"

Revision 15, and Procedure 3.8, " Drawing Control," Revision 12. The procedures stated that drawing change notices alert drawing users of approved and/or installed changes, such as modifications and as-builts not yet reflected on the drawin The licensee stated that for the most part the safety evaluation for drawing change notices was performed with the associated modification package or engineering evaluation. The licensee further stated that stand-alone drawing change notices receive a 10 CFR 50.59 safety evaluation. The inspector reviewed three drawing change notices which were associated with a

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modification package and one stand-alone drawing change notice. The inspectcr determined that the licensee was appropriately performing 10 CFR 50.59 evaluations for drawing change notice .  :---;,

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-28- CFR 50.59 Safety Evaluations Not Performed for Enaineerina Evaluations Which Meet Annendix B Screen j i

The inspector reviewed Engineering Procedure 3.4.5, " Engineering )

Evaluations," Revision 1c2, The inspector found that in part the licensee defined engineering evaluations as an engineering document used to authorize station modifications via the repair, use-as-is, or rework of plant systems, structures, or components not appropriately supported by any engineering, design basis, or design output document. The inspector found that the procedure required a 10 CFR 50.59 screening review for all engineering evaluations, except those dispositioned as rework (return to

! original configuration), those ured to collect information, and those which l met an Appendix B screen per Procedure 3.4.

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' The inspector noted that the requirements of 10 CFR 50,59 extend beyond the scope of 10 CFR Part 50, Appendix B. The licensee stated that the 10 CFR Part 50, Appendix B, screen effectively functions the same as a 10 CFR 50,59 screen and as a result can be used interchangeably to determine if a full 10 CFR 50,59 safety evaluation is required Further inspection is planned to review the content of Procedure 3,4 to determine if the licensee is performing required 10 CFR 50.59 evaluations. This item will be tracked as inspection Followup Item 50-298/97007-0 Conclusions The inspector found several potential weaknesses in the licensee's program for determining when a 10 CFR 50.59 safety evaluation was required. The licensee did not perform a 10 CFR 50.59 screening review for nonintent procedure changes, calculation changes, or changes which met an Appendix B screen, More 'nspection ,

is planned to evaluate effectiveness of the Appendix B screen. The inspector noted, however, that when weaknesses were discussed with licensee personnel, they promptly evaluated the concerns to determine if corrective actions were necessar E4 Engineering Staff Knowledge and Performance E Control Room Briefina Bv System Enaineer Inspection Scoce (37551)

During observation of startup activities, inspectors observed a briefing by a system engineer to the control room crew of a newly installed modification to a safety-related syste _ _ _ - _ - _ _ - _ - - -

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e 29-i Observations and Findinas  !

l During August 3,1997, observations of plant startup activities, inspectors observed l a senior system engineer brief the control room on a modification of a sump l affecting the standby gas treatment system. The engineer observed all control room conduct of operations requirements and provided a succinct, clear, accurate, and well organized description of the modification and the impact of the modification on the operations crew and procedures. The safety significance of the changes was also described. The engineer referred to drawings and procedures, describing the changes in a logical fashion and tailored to the operations crew requirements. Reference information, consisting of excerpts of marked-up drawings, procedures, and relevant descriptions of expected visual changes to plant equipment and changes to plant operations, was provided to all crew member Conclusion A system engineer provided an excellent briefing of a recent plant modification to the control room crew, in preparation for reactor startup operation E8 Miscellaneous Engineering issues E8.1 (Open) LER 50-298/96-014-01: Fuel preparation machine upper stop set in violation of Technical Specifications. Technical Specification 3.10.C stated, in part, whenever irradiated fuel is stored in the spent fuel pool, the pool water level shall be maintained at or above 8.5 feet above the top of fuel. The original fuel design was based on an active fuellength of 144 inches. In 1972 and 1978 the licensee installed fuel with active fuellengths of 146 and 150 inches, respectively. The increased active fuellength allowed the potential for fuel held in the fuel preparation machine at the existing upper stop to have less than the Technical Specification limit ot 8.5 feet of water above the top of active fue In the LER, the licensee concluded that there was no safety sign.7icance regarding improper setting of the upper stop based on Standard Technical Specifications for General Electric Plants, BWR/4 (NUREG-1433). The Standard Technical Specifications for General Electric Plants state that the bases for the spent fuel pool water level Technical Specification limit is an explicit assumption of the fuel handling accident regarding level over the top of irradiated fuel assemblies seated in the spent fuel storage pool racks. The inspectors reviewed the licensee's current Technical Specification and the associated bases. The documents described that the minimum water level of 8.5 feet must be maintained above irradiated fuel assemblies while in the spent fuel pool for adequate water to shield and cool the fuel. Also the inspectors reviewed NUREG-1433, Updated Safety Analysis Report Chapter XIV, Station. Safety Analysis, and held discussions with the license Through review and discussions, the inspectors verified that the refueling accident, dropping a fuel bundle onto the top of the core, would still be the most limiting accident for offsite dose associated with fuel storage and handling (fuel preparation

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-30 I machine use). The licensee determined during a nonaccident condition that the area dose rates would increase 3 mrem / hour if an irradiated fuel bundle was at the highest position of the fuel preparation machine. Based on the information provided by the licensee in the LER, the inspectors were unable to conclude that this issue had no safety significance. The inspectors identified that the licensee failed to describe relevant design basis requirements that pertained to the Cooper Nuclear Station license and to discuss the estimated higher dose rate The licensee tagged the fuel preparation machine out of service until the upper stop was reset, which was performed under Modification Package 96134 Also, the licensee issued drawing changes to update the applicable drawings. The licensee determined that no further corrective actions were required since all future fuel design changes were performed using Procedure 3.4, " Station Modification," which provided for verification of design change impact on site documents, including Technical Specifications. This was not a modification requirement in 1978, the latest fuellength change. The licensing manager has stated that this LER will be used as an example of an inadequate safety significance basis to communicato his expectations to his staff for outgoing correspondenc The licensee did not provide enough information for the inspectors to conclude that there was an absence of safety significance. However, based on reviews of design documents and discussions with the licensee's staff, the inspectors concluded that there were minimal safety consequences for having the fuel preparation machine upper stop at the nonconservative setpoint. The failure to maintain greater then feet of water above active fuel is a violation of Technical Specification 3.10.C, which requires, whenever irradiated fuel is stored in the spent fuel pool, the pool water level shall be maintained at or above 8.5 feet above the top of the fuel. This licensee-identified and corrected violation is being treated as a noncited violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy (NCV 298/97007-09), This LER will remain open pending submission of a supplement to correct the description of the safety significanc VI. Manaaement Meetinas X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management at an exit meeting on July 31,1997, covering operability assessments and 10 CFR 50.59 safety evaluations, and an exit meeting on August 14,1997, covering the remaining sections of the report. The licensee acknowledged the findings presente The inspectors asked the licenseo whether any materials examined during the inspection should be considered proprietary. No proprietary information was identifie , . . . . . . . . .

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e ATTACHMENT PARTIAL LIST OF PERSONS CONTACTED Licensee

1. Tom Black, Senior Engineering

1. Dave Bremer, Operations Support Group Supervisor

1. * Dan Buman, Engineering Support Manager

• Paul Caudill, Senior Manager Safety Assessment / Site Support

1. Linda Dewhirst, Nuclear Licensing and Safety Engineer

1. • Fadi Diya, Design Engineering Manager

1. Jerry Dorn, Containment Engineering Supervisor

1. Larry Dugger, Institute of Nuclear Power Operations -

1. Jim Florence, Actin Nuclear Training Manager

• Lisa Freeman, Oper dions Clerk l # Chuck Gaines, Maintenance Manager l # Philip Graham, Vice President - Nuclear

1. • Brad Houston, Manager Nuclear Licensing

1. Ralph Krause, Acting l&C Design Engineering Supervisor

1. Jim Long, Performance Analysis Manager

1. Chris Moeller, Licensing Engineer

• Brian Newell, Assistant Maintenance Manager

1. • Larry Newman, Assistant Operations Manager

1. ' Ole Olson, Piant Engineering Department Manager

1. Dhiren Pandya, Civil DED Engineering Supervisor

• Mike Peckham, Plant Manager

' Jim Pelletier, Senior Manager of Engineering

1. D. Robinson, Quality Assessment Manager

' Richard Sessoms, Senior Manager Quality Assurance

1. Michael Spencer, Engineering Programs Supervisor

1. Rick Wachowiak, Reliability Engineering Supervisor

1. Steve Wheeler, Station Technical Engineer NRC

1. Charles Marschall, Project inspector

1. Linda Smith, Reactor inspector

1. Attendees at July 31,1997, meeting

• Attendees at August 14,1997, meeting

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s-2-INSPECTION PROCEDURES USED ,

IP 37550: Engineering I IP 375f 1: Onsite Engineering j IP 37828: Engineering - Install and Test Modifications >

IP 61726: Surveillance Observations IP 62707: Maintenance Observation IP 71707: Plant Operations

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l IP 71750: Plant Support Activities l l

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lP 92901: Followup - Plant Operations -

IP 92902: Followup - Maintenance IP 92700: Onsite Followup of Written Reports of Non-Routine Eunts at Power Reactor l Facilities IP 93702: Prompt Onsite Response to Reactor Events I

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ITEMS OPENED, OPENED AND CLOSED, REVIEWED, AND CLOSED Opened 298/97007-01 URI Multiple examples wherein inadequate corrective actions were taken (Sections 02.1, 07.2, 08.2, 08.4, M 1.1, M8.1, and E2.1.b.5).

298/97007-05 VIO Failure to include proper test criteria (Sections E2.1 and M1.1)

298/97007-06 IFl Verify adequate program for assessing valve operebility exists (Section E2.1,b.1).

298/97007-07 URI Review revised operability assessment on the reactor vessel water level leg (Section E2.1.b.2).

298/97007-08 IFl Determine whether 10 CFR 50.59 safety evaluations are performed when required (Section E3.1.b.4).

Opened and Closed 298/97007-02 NCV Two examples where operators f ailed to follow procedures (Section 04.1).

298/97007-03 NCV Failure to maintain design requirements in procedure (Section 07.1.b.1).

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-3-298/97007-04 NCV Failure to implement procedure for level restoration (Section M8.3)

298/97007-09 NCV Failure to maintain greater then 8.5 feet of water above active fuel (Section E8.1).

Closed 298/97-004-00 LER Automatic actuations as a result of improper maintenance work on vessel level indication system (Section M8.3).

298/97005-02 IFl Inadequate operability assessment of the interlock between the service water booster pumps and the curresponding residual heat removal heat exchanger service water outlet val <e (Section 08.1).

l 298/97005-01 URI Control rod drive housing supports procedure acceptance

! criteria and operability assessments were inadequate (Section 08.2).

298/95-017-00 298/95-017-01 298/94-033-00 LER Safety Relief Valves Found Outside Technical Specification (Section 08.3).

298/96013-01 VIO Opening and closing of containment hatch without procedure (Section 08.4)

298/97013-05 VIO Multiple examples of 10 CFR Part 50, Appendix B, Criterion V (Section 08.5).

298/96004-02 VIO Diesel Generator 2 muffler bypass valve inoperable due an unauthorized modification (Section M8.1).

298/96019-01 VIO Undervoltage Jumper (Section M8.2).

Reviewed 298/97-002-00 LER Safety relief valves outside Technical Specification (Section 08.3)

298/97-009-00 LER Groups 2,3, and 6 isolations during controlled shutdown (Section 01.1).

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298/97-010-00 LER Standby gas system inoperability due to collection of water in sump (Section 07.2).

298/97 011-00 LER Vacuum breaker failure resulting in plant shutdown (Sections 01.1, M1.1).

298/97-012-00 LER High torus water level (Sections 01.2 and 04.1).

l 298/96026-06 IFl Failure to include instrument tolerances in plant procedures for Technical Specification equipment (Sections 04.1, M1.1, M8.4).

298/96-014-01 LER Fuel preparation machine upper stop set in violation of Technical Specifications (Section E8.1).