IR 05000244/1994012

Insp Rept 50-244/94-12 on 940503-0606.Noncited Violation Noted.Major Areas Inspected:Plant Operations,Maint, Engineering,Plant Support & Safety Assessment

ML17263A721
Person / Time
Site:	Ginna
Issue date:	07/07/1994
From:	Lazarus W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML17263A720	List: IR 05000244/1994012 ML17263A719
References
50-244-94-12, NUDOCS 9407180009
Download: ML17263A721 (17)

Text

U. S. NUCLEAR REGULATORY COMMISSION

REGION I

Inspection Report 50-244/94-12 License: DPR-18 Facility:

Inspection:

Inspectors:

R. E. Ginna Nuclear Power Plant Rochester Gas and Electric Corporation (RG&E)

May 3 through June 6, 1994 T. A. Moslak, Senior Resident Inspector, Ginna E. C. Knutson, Resident Inspector, Ginna Approved by:

W.

s, Ch', Reactor Projects Section 3B INSPECTION SCOPE Plant operations, maintenance, engineering, plant support, and safety assessment/quality verification, 9'407i80009 940707 PDR ADOCK 05000244 Q

PDR

Executive Summary INSPECTION EXECUIIVE SUMMARY Operations At the beginning of the inspection period the plant was operating at full power (approximately 97 percent).

On May 3, 1994, a controlled reactor shutdown was performed to replace a leaking component cooling water system flexible hose to the "A" reactor coolant pump motor lower bearing oil cooler. A reactor startup was conducted on May 5, 1994, and the plant was returned to fullpower operation on May 6, 1994.

The plant operated at fullpower for the remainder of the inspection period.

Maintenance The "C" service water pump was replaced after routine surveillance testing revealed low pump differential pressure and high vibrations. The threads between the pump upper shaft bushing and the pump casing were found to be stripped; the cause of the reduced pump differential pressure and root cause of the pump failure were under investigation at the close of the inspection period.

The pump had been in service for approximately one year, prior to which it had been refurbished by the licensee.

Strong management involvement was observed in directing repair activities.

The "B," standby auxiliary feedwater pump room cooler was inoperable for 16 days due'o incomplete restoration from maintenance.

The inspector considered the event to be an apparent violation of 10 CFR 50 Appendix B, criterion XVI, "Corrective Action," in that the condition was not promptly identified. The licensee was not cited because the criteria specified in Section VII.Bof the Enforcement Policy were satisfied.

The inspector identified a seismic mount in the containment air sampling system held together by a C-clamp.

Subsequent engineering evaluation demonstrated that containment reliability had not been affected by= this deficiency.

A comprehensive corrective action plan was initiated; actions were ongoing at the close of the inspection period.

Although, in itself, not safety significant, the inspector considered that this event demonstrated weaknesses in the identification of seismic supports and in closure control of maintenance work packages.

Engineering Vibration levels on the "A" reactor coolant pump approximately doubled (-10 mils) following motor replacement and seal package inspection.

Evaluation by licensee engineering and vendor technical representatives determined that continued operation would not degrade pump reliability.

Plant modifications and improvements to eliminate feedwater flow oscillations have effectively eliminated this long-standing problem; however, small oscillations in steam pressure have persisted.

The licensee has contracted an independent consultant to attempt to identify the cause of the oscillation Executive Summary Temporary leak repair activities and the temporary modification program were reviewed.

Recommendations for program improvements were discussed with the licensee; no significant deficiencies were noted.

Plant Support Routine observations in the areas ofradiological controls, security, and fire protection indicated that these programs were effectively implemented.

Safety Assessment/Quality Verification At a meeting of the Nuclear Safety Audit and Review Board, the inspector observed candid, knowledgeable discussion of agenda item TABLEOF CONTENTS EXECUTIVE SUMMARY e

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TABLE OF CONTENTS o

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1V 1.0 OPERATIONS (71707)

1.1 Operational Experiences 1.2 Control ofOperations...............................

1.3 Component Cooling Water System Leak From "A"Reactor Coolant Pump

, Motor Bearing Cooler 1.4 Control Room Operations During Reactor Startup and Power Increase

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2 2.0 MAINTENANCE(62703, 61726).....................

2.1 Preventive/Corrective Maintenance................

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"C" Service Water Pump Low Differential Pressure

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"B" Standby AuxiliaryFeedwater Pump Room Cooler Due To Incomplete Restoration From Maintenance 2.1.3 Valve AOV-1599 Seismic Mount 2.2 Surveillance Observations 2.2.1 Routine Observations....................

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Inoperable

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3.0 ENGINEERING (71707)

3.1 Temporary Modifications Program Review................

3.2 Site Engineering Activities..........................

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4.0 PLANT SUPPORT (71707)

4.1 Radiological Controls....

4.1.1 Routine Observations 4.2 Security 4.2.1 Routine Observations 4.3 Fire Protection......,..

4.3.1 Routine Observations

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5.0 SAFETY ASSESSMENT/QUALITY VERIFICATION(71 5.1 Periodic Reports.....................

5.2 Licensee Event Reports.................

5.3 Nuclear Safety Audit and Review Board Meeting

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6.0 ADMINISTRATIVEo

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6.1 Backshift and Deep Backshift Inspection 6i2 Exit Meetlilgs

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1.0 OPERATIONS (71707)

DETAILS

.1'.1 Operational Experiences At the beginning of the inspection period the plant was operating at full power (approximately 97 percent).

On May 3, 1994, a controlled reactor shutdown was performed to support repair of a component cooling water system leak in containment.

The source of the leak was flexible hose to the "A" reactor coolant pump motor lower bearing oil cooler.

A reactor startup.was conducted on May 5, 1994, and the plant was returned to fullpower operation on May 6, 1994.

The plant operated at full power for the remainder of the inspection period.

There were no other significant operational events or challenges during the inspection period.

1.2 Control of Operations Control room staffing was as required.

Operators exercised control over access to the control room.

Shift supervisors maintained authority over activities and provided detailed turnover briefings to relief crews.

Operators adhered to approved procedures and were knowledgeable of off-normal plant conditions.

The inspectors reviewed control room log books for activities and trends, observed recorder traces for abnormalities, assessed compliance with technical specifications, and verified equipment availability was consistent with the requirements for existing plant conditions.

During normal work hours and on backshifts, accessible areas of the plant were toured.

No operational inadequacies or concerns were identified.

1.3 Component Cooling Water System Leak From "A" Reactor Coolant Pump Motor Bearing Cooler On the afternoon ofMay 3, 1994, control room operators observed, using the containment video monitoring system, that a small water leak had developed in the vicinity of the "A" reactor coolant pump (RCP).

The source of the leak was outside of the fixed camera's field of view.

A containment entry was conducted, and the source was determined to be a component cooling water (CCW) system flexible hose to the "A" RCP motor lower bearing oil cooler.

Repair would require the "A" RCP to be secured, which, in turn, required that the reactor be shut down.

Operators commenced a 20 percent per minute load reduction at 6:15 p.m.

Reactor shutdown was completed at 10:51 p.m., and the."A" RCP was secured at 11:12 p.m., May 3, 1994.

The "A"RCP motor had been removed for refurbishment during the 1994 refueling outage and had been replaced with an onsite spare.

The physical arrangement of the supply and return

'piping for the lower motor bearing oil cooler on the new niotor was slightly different than the original motor. The difference had been accommodated by flexible hoses that connect the cooler to the CCW system hard piping; however, in doing so, these hoses had been installed in a configuration with tight bends, and the supply hose had been forced against piping for the RCP oil collection system.

The resultant combination of forces had lead to failure ofthe flexible hose at the hose-to-flange transitio The installed CCW supply and return piping was modified to reduce the amount ofbend that the flexible hoses would be required to assume.

In addition to replacing the failed supply hose, the return hose was replaced as a preventive measure.

Another significant repair activity during this forced outage was to correct seat leakage from relief valve 203 goop "B" letdown to nonregenerative heat exchanger relief to pressurizer relief tank). The valve had been replaced during the 1994 refueling outage.

Seat leakage had developed after the valve was challenged gifted and reseated)

during the transient produced by the April 27 reactor trip (see inspection report 50-244/94-07).

On April 28, the valve was replaced with an onsite spare, which also developed seat leakage.

The original valve (the valve that had been removed on April28) was rebuilt and reinstalled on May 4.

The valve performed satisfactorily for the remainder of the inspection period.

At the close of the inspection period, the licensee completed a root cause analysis of the RV-203 repeated failures.

Maintenance and repair activities were completed on May 5, 1994, and a plant startup was performed.

Criticality was achieved at 10:01 a.m. and the main generator was closed on the grid at 2:37 p.m.

Following a hold at 30 percent power while feedwater chemistry was stabilized, the plant was returned to full power operation on May 6, 1994.

The inspector observed operations in the control room from a time shortly after the leak was initiallydiscovered until the "A" RCP was secured and the leak was isolated.

The inspector noted that prompt action was taken to identify the source of the leak; until this was done, operator attention was appropriately focused on parameters and trends that would identify degrading conditions in any of the potentially affected systems.

Once the source was known to be CCW, a course ofaction was developed and promptly carried out. Management involvement was appropriate, with a representative present in the control room throughout the event.

Operators discussed contingent actions against the condition worsening or the flexible hose failing completely.

The power reduction and reactor shutdown were completed expeditiously and professionally.

The inspector reviewed the licensee's root cause analysis report for this event.

The report, M-94-008, "'A'CP Lower Bearing CCW Flexible Hose Failure," identified the cause of the flexible hose failure as a lack of knowledge of the correct bending limitation parameters associated with installation of flexible hoses.

The inspector considered that the root cause analysis report was thorough and presented an accurate assessment of the primary and contributing causes of the failure.

1.4 Control Room Operations During Reactor Startup and Power Increase On May 5, 1994, the inspector observed the reactor startup from the control room.

The inspector considered that attention in the control room did not appear to be as highly focused on the reactor-startup as had been observed during past startups.

In particular, the inspector considered that some routine operational and administrative activities, conducted in parallel with the startup, could have been deferred until after criticality was achieved, thereby avoiding unnecessary distractions.

Additionally, the final stepped rod withdrawal to criticality was

intentionally initiated before the final point on the 1/M plot was plotted. This was done because reactor power was approaching the point at which the source range nuclear instruments (SRNIs)

were required to be deenergized to prevent an automatic reactor trip.

The 1/M plot is an operator aid for verifying reactivity addition and anticipating criticality, and is directed to be maintained prior to criticality by operating procedure 1.2, Plant Startup From Hot Shutdown to Full Load."

Although the operators were within procedural allowances in omitting the last 1/M point, the inspector considered that this condition could have been anticipated such that hold points for plotting 1/M would not have conflicted with achieving criticalityand deenergizing the SRNIs.

Overall, operator performance was adequate.

The inspector observed control room operations during the power escalation to fullpower. The inspector noted good procedural adherence and excellent communication between the operators.

Major operations, such as starting the main feedwater pumps, were methodically approached and well coordinated with other plant operations.

2.0 MAINTENANCE(62703, 61726)

2.1 Preventive/Corrective Maintenance 2.1.1

"C" Service Water Pump Low Differential Pressure Performance test (PT)-2.7.1, "Service Water Pumps - Quarterly," conducted on May 18, 1994, indicated that a problem had developed with the "C" service water (SW) pump.

Specifically, pump differential pressure (d/p), at 52.1 psid, was less than the minimum allowable d/p of53.7 psid; additionally, two measured values of pump motor vibration were in the alert range.

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"C" SW pump was subsequently declared inoperable; no restrictions are imposed by technical specifications for inoperability of a single SW pump.

Corrective action was initiated through administrative procedure (A)-25.1, "Ginna Station Event Reports."

Immediate corrective action included:

Initiation of a work request/trouble report; satisfactory calibration verification of the "C" SW'pump discharge pressure gauge and flushing of the associated sensing line; data collection for pump performance testing; and review of the condition by electrical engineering and results and test (R&T) personnel to assess the available information and develop a tentative program for subsequent testing. The followingday, licensee management held a meeting to establish the plan for additional troubleshooting.

It was concluded that the PT should be conducted again to verify repeatability of the data, as well as to obtain additional information on pump motor performance and measurements for enhanced vibration analysis.

Results of this testing confirmed the previous day's results and established the pump (as opposed to the motor) as the likely source of the problem.

Upon pump removal, the inspectors noted that the threads between the pump upper shaft bushing and the pump casing were stripped.

The licensee considered that this could have been the. cause of the increased motor vibration.

The pump was replaced with a rebuilt onsite spare.

Additionally, the shaft that connects the pump to the motor was found to be slightly bowed, and

was replaced.

Following run-in and acceptance testing, the "C" SW pump was declared operable on June 1,

1994.

As of the end of the inspection period, the cause of the low-differential pressure had not been determined.

A root cause analysis investigation (M-94-010)

of the pump failure was in progress.

The inspector assessed that the licensee's response to the "C" SW pump problem resulted in prompt action to restore operability.

Management effectively organized and directed troubleshooting efforts, and support from the cognizant organizations was readily available.

Computer assisted trending and spectral vibration analysis were particularly valuable in establishing the pump as the source of the problem.

The licensee was particularly concerned with this pump failure, in that, following rebuild, the pump had been in service for a relatively short period (approximately one year).

Extensive testing was conducted prior to pump replacement to establish the measurable characteristics of the particular mode of. failure.

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"B"Standby AuxiliaryFeedwater Pump Room Cooler Inoperable Due To Incomplete Restoration From Maintenance While attempting to perform PT-60.2, "'B'tandby AuxiliaryFeedwater (SAFW) Pump Room Cooler Performance Test," on May 6, 1994, the service water (SW) discharge isolation valve for the "B" SAFW pump room cooler, the test personnel found V-9635 shut.

This valve misalignment caused SW flow to the heat exchanger to be secured, and therefore rendered the

"B" SAFW pump room cooler inoperable; in that the SAFW pump operability analysis was based on the associated cooler being operable, the "B" SAFW pump therefore also had to be considered inoperable.

Technical specification 3.4.2.3.a allows one SAFW pump to be inoperable for up to 14 days before requiring the plant to be shutdown.

Operability of the "B" SAFW pump was promptly restored by opening V-9635. Additionally, the remainder of the SAFW system was verified to be properly aligned.

Subsequent investigation revealed that V-9635 had been closed on April20, 1994, to provide maintenance isolation for work on the "A" SAFW pump room cooler.

This work was completed on April 22, 1994, however, V-9635 was mistakenly omitted from the restoration valve lineup.

Therefore, the "B" SAFW pump room cooler had been inoperable from April 20 to May 6, 1994, for a total of 16 days.

The licensee conducted an engineering evaluation to determine ifSAFW pump operability was dependent on operability of its associated room cooler.

This evaluation demonstrated that the limiting parameter in the initial operability analysis, room temperature, would remain within acceptable limits, ifone SAFW pump room cooler was inoperable.

Therefore, the licensee concluded that a violation of technical specification 3.4.2.3.a had not occurre The inspector attended the PORC meeting at which this engineering evaluation was presented.

The inspector considered that a more in-depth presentation (such as, for example, discussion of the assumptions made in preparing the analysis) might have been useful in ensuring that PORC was fullyapprised of the relevant safety considerations.

The inspector reviewed the evaluation and considered that the analysis was thorough and conservative.

At the close of the inspection period, the licensee was conducting a human performance review (94-10) of this event.

Based on the licensee's engineering evaluation, the inspector considered that inoperability of the

"B" SAFW pump room cooler did not constitute a violation of technical specifications.

However, 10 CFR 50 Appendix B, criterion XVI, "Corrective Action," states that, "Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected."

The inspector considered this event to be an apparent violation of this requirement, in that the mispositioned valve went undetected for 16 days. However, the licensee was not cited, in that the requirements of 10 CFR 2 Appendix C, section VII, "Exercise of Discretion," were satisfied.

Specifically, the violation was:

Not a willfulviolation or a violati prevented by the licensee's correcti 2.1.3 Valve AOV-1599 Seismic Mount Identified by the licensee; Of minor safety significance; Corrected, including measures to prevent recurrence, within a reasonable time; and

'on that could reasonably be expected to have been

've action for a previous violation.

On May 12, 1994, the inspector identified that the mounting bracket for valve AOV-1599 (containment air sample isolation valve) was apparently not properly made up.

As found, the mounting bracket was held together by a C-clamp; however, existing holes in the mount suggested that it was supposed to be held together by two bolts.

The same day, the licensee generated a work request for the condition of the mount to be evaluated.

In the normal course of work order processing, the subject work request was forwarded to site technical engineering on May 19, 1994.

Quality Assurance (QA) investigated and determined that the support in question was a seismic support and was supposed to be held together by two bolts.

A nonconformance report (NCR 94-095) was generated to document the deficiency.

During subsequent review within the technical engineering group, the system engineer realized that the condition of the'AOV-1599 mount constituted a possible violation of containment integrity, as required by technical specification 3.6.1. On May 23, 1994, an A-25.1 event report was generated and site management was informed of the AOV-1599 mount deficiency.

Multiplecorrective actions were initiated as a result of the AOV-1599 event report. Within the first day, the mount was made up usirig two bolts of slightly smaller diameter (the mount had apparently not been restored following maintenance because misalignment of the existing bolt

holes prevented reinstallation of the original bolts); adequacy of this new arrangement was established by mechanical engineering design analysis DA-ME-94-066, completed the same day, and the C-clamp was removed.

An engineering evaluation was initiated to determine whether containment integrity had been violated while the temporary fastener had been in place.

Quality assurance reviewed outage work orders for items that involved seismic supports; from this screening, a sample group was selected for in-plant verification of support integrity. A human performance evaluation system (HPES 94-11) review of the event was initiated. Additionally, licensee management directed that a corrective action report (CAR-2091) be initiated to progress these efforts and to ensure that all aspects of the event were investigated and resolved.

On May 25, 1994, the inspector attended a licensee meeting to discuss CAR-2091. The licensee'ad determined the sequence of events that led to the undocumented mounting deficiency on AOV-1599.

Maintenance had been performed on AOV-1599 during the 1994 refueling outage.

The work had been authorized by addition to a previously existing work order (9200766) to replace a nearby valve, AOV-1596 (inlet block valve. to containment air sample inlet valve). The AOV-1596 replacement required disassembly ofAOV-1599 due to physical interference; the AOV-1599 maintenance, periodic replacement of the diaphragm in the air operator, was specified in the work order instructions to be accomplished in conjunction with disassembly.

Inclusion ofthis additional maintenance to the AOV-1596 work package was within the allowances ofadministrative procedure (A)-1603.2, "Work Order Initiation."

Work on AOV-1599 was to be performed in accordance with maintenance procedure (M)-37.10, "Grinnell Diaphragm Valves, Air Operated, Maintenance."

This generic procedure contains a step to invoke additional guidance ifseismic supports are involved.

It was apparently the maintenance planner's understanding that all seismic supports were identified by brass identification tags.

In performing the pre-job walkdown, the maintenance planner saw no such identification tag on the AOV-1599 mount, and therefore marked the procedure step concerning seismic supports as "not applicable."

In that the AOV-1599 maintenance was being accomplished within the scope of an interference removal in conjunction with a previously approved work package, no reviews beyond the maintenance planner were required.

The valve maintenance was performed and, following reassembly, the maintenance technician noted misalignment between the bolt holes in the two halves of the AOV-1599 mount.

The technician temporarily secured the mount with a C-clamp.

The technician apparently forgot about this deficiency and, because the work package contained nothing that addressed reassembly of the AOV-1599 mount, the work package was closed ou Mechanical engineering design analysis DA-ME-94-067 was completed on May 24, 1994, and demonstrated that integrity of the containment air sample line from the containment penetration to AOV-1599 would have been maintained during a design basis seismic event even with the AOV-1599 mount disconnected.

Therefore, the licensee concluded that a violation oftechnical specification 3.6.1 had not occurred.

Based on the licensee s engineering evaluation, the inspector concluded that this event, in itself, was not safety significant, in that containment reliability was not compromised.

However, the inspector considered that this event demonstrated two significant deficiencies in the licensee's maintenance program; specifically, 1) identification of seismic supports was not well understood by maintenance personnel; and 2) it was possible for work package closure to occur prior to the actual completion of work.

As noted, the licensee is addressing these deficiencies through a human performance review; 2.2 Surveillance Observations 2.2.1 Routine Observations Inspectors observed portions ofsurveillances to verifyproper calibiation oftest instrumentation, use of approved procedures, performance of work by qualified personnel, conformance to limitingconditions for operation (LCOs), and correct system restoration following testing.

The following surveillances were observed:

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Performance test (PT)-1, "Rod Control System," revision 35, effective date October 14, 1993, observed May 18, 1994 Annunciator C-23, "Insertion LimitBank C Lo," did not clear upon restoration ofcontrol bank C to its original configuration. Following discussions with results and test personnel and site engineering, testing was resumed with the annunciator still in alarm. A work request was generated to correct the annunciator problem.

Troubleshooting under work order 19400022, "Bistable TC-405UL," determined the cause to be a failed SCR in the low insertion limitalarm circuit, which had caused the alarm setpoint to drift.

PT-2.7.1, "Service Water Pumps," revision 11, effective date April 13, 1994, observed May 19, 1994 Performed to verify unsatisfactory results obtained on May 18, 1994, for "C" service water pump differential pressure and vibration.

This testing confirmed the previous day's results and led to pump replacement.

The inspector determined through observing this testing that operations and test personnel adhered to procedures, test results and equipment operating parameters met acceptance criteria, and redundant equipment was available for emergency operatio.0 ENGINEERING (71707)

3.1 Temporary Modifications Program Review The inspector conducted a review of the licensee's temporary modifications program.

Administrative procedure (A)-1406, "Control ofTemporary Modifications," establishes program applicability, requirements, and controls.

The inspector noted several instances where quantitative guidance would be more appropriate than the current qualitative guidance; these items were discussed with the licensee for information.

Overall, the inspector considered that the program established appropriate controls for the initiation, installation, administrative tracking, periodic inspection and review, and removal of temporary modifications.

Requirements for safety evaluations and Plant. Operations Review Committee (PORC) review were clearly specified and appropriate.

The inspector had no additional concerns with this item.

3.2 Site Engineering Activities

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Increase in "A"Reactor Coolant Pump Vibration During the 1994 refueling outage, the "A"reactor coolant pump (RCP) motor was removed for refurbishment and replaced with an onsite spare.

The RCP sea1 package was also disassembled and inspected.

During plant startup and escalation to fullpower operation, vibration of the "A" RCP, as indicated by the installed monitoring equipment, was noted to increase.

Prior to motor replacement, vibration. was on the order of five mils; with the new motor, vibration had stabilized at approximately 10 mils.

Although higher than the previous value, this level of vibration did not present an operability concern; the RCP vendor's recommended vibration limit is 20 mils. Nonetheless,, the licensee was concerned that the increased RCP vibration could be due to a material deficiency, and attempted to establish the cause.

Westinghouse technicians were brought in to evaluate the "A"RCP vibration. The technicians conducted extensive data gathering and onsite data reduction and evaluation.

Although the cause of the increased vibration was not definitely established, the technicians were confident that continued operation of the RCP at the existing vibration level would not degrade pump operability or reliability.

The inspector considered that licensee action to identify the cause of increased vibration of the

"A" RCP was prudent.

Site and corporate engineering support was readily available, and data gathering was well coordinated.

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Steam Pressure Oscillations Following installation ofthe advanced digital feedwater control system (ADFCS) during the 1991 refueling outage, the plant experienced feedwater flow oscillations.

These oscillations, with a period on the order of several minutes, translated into oscillations in various plant parameters,

including steam flow, steam pressure, and average reactor coolant temperature.

On several occasions, the magnitude ofthe oscillations has increased to the point ofadversely affecting plant operations and requiring operator intervention to stabilize conditions.

Numerous corrective actions have been attempted to eliminate the feedwater flow oscillations.

Modification of the feedwater regulating valve control air system during the 1993 refueling outage, and, most recently, modification of the feedwater regulating valve internals to provide improved flow control, has greatly reduced the magnitude of feedwater flow oscillations.

Despite having effectively eliminated feedwater flow oscillations, the plant is continuing to experience oscillations in steam pressure.

Although these oscillations do not present an apparent operational challenge or a safety concern, they could be indicative of a control system malfunction or a destabilizing control system interaction.

The licensee has contracted an

, independent consultant to analyze the steam pressure oscillations and to identify the source.

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Main Steam Check Valve 3518 Temporary Leak Repair On May 9, 1994, the inspe:tor identified a minor steam leak from the stuffing box flange of valve 3518 (steam generator "B" main steam check valve). The licensee subsequently conducted an on-line temporary leak repair by injecting a leak sealing compound into the gasket cavity.

The inspector reviewed the engineering evaluation of the temporary leak repair.

The inspector noted some technical errors in this evaluation; for example, the values of temperature and pressure had been transposed.

Additionally, the inspector considered that the evaluation could have been more thorough; for example, no evaluation had been made of the possible impact that excess sealant on the valve pivot shaft might have on valve operability.

These concerns were discussed with the licensee, and subsequently addressed in a revised engineering evaluation.

The inspector considered that this revised evaluation was thorough.

4.0 PLANT SUPPORT (71707I 4.1 Radiological Controls 4.1.1 Routine Observations The inspectors periodically confirmed that radiation work permits were effectively implemented, dosimetry was correctly worn in controlled areas and dosimeter readings were accurately recorded, access to high radiation areas was adequately controlled, survey information was kept current, and postings and labeling were in compliance with regulatory requirements.

Through observations ofongoing activities and discussions with plant personnel, the inspectors concluded that the licensee's radiological controls were effective...

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4.2 Security 4.2.1 Routine Observations During this inspection period, the inspectors verified that x-ray machines and metal and explosive detectors were operable, protected area and vital area barriers were well maintained, personnel were properly badged for unescorted or escorted access, and compensatory measures were implemented when necessary.

No unacceptable conditions were identified.

4.3 Fire Protection 4.3.1 Routine Observations The inspectors periodically verified the adequacy of combustible material controls and storage in safety-related areas of the plant, monitored transient fire loads, verified the operability offire detection and suppression systems, assessed the condition of fire barriers, and verified the adequacy of required compensatory measures.

No discrepancies were noted.

5.0 SAFETY ASSESSMENT/QUALITY VERIFICATION (71707)

5.1 Periodic Reports

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Periodic re rts submitt po ed by the licensee pursuant to Technical Specification 6.9.1 were reviewed, Inspectors verified that the reports contained information required by the NRC, that test results and/or supporting information were consistent with design predictions and performance specifications, and that reported information was accurate.

The following report was reviewed:

Monthly Operating Report for April 1994 No unacceptable conditions were identified.

5.2 Licensee Event Reports

A Licensee Event Report (LER) submitted to the NRC was reviewed to determine whether details were clearly reported, causes were properly identified, and corrective actions were appropriate.

The inspectors also assessed whether potential safety consequences were properly evaluated, generic implications were indicated, events warranted additional onsite follow-up, and applicable requirements of 10 CFR 50.73 were met.

The following LER was reviewed (Note:

date indicated is event date):

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94-007, Feedwater Transient, Due to Loss of Abilityto Control Feedwater Regulating Valve, Causes a Lo Steam Generator Level Reactor Trip (April27, 1994)

The inspector concluded.

that the LER was accurate, met regulatory requirements, and appropriately identified the root cause.

5.3 Nuclear Safety Audit and Review Board Meeting On May 11, 1994 the inspector attended a meeting of the Nuclear Safety Audit and Review Board (NSARB). Topics included proposed changes to technical specifications and a proposed revision to the quality assurance program.

The inspector observed that information presented

.was relevant and clearly presented.

The depth and quality of board discussions reflected a knowledgeable membership.

The inspector concluded that the NSARB provided valuable oversight of plant management and operations.

6.0 ADMINISTRATIVE 6.1 Backshift and Deep Backshift Inspection During this inspection period, a backshift inspection was conducted on May 3, 1994.

Deep backshift inspections were conducted on May 15, 20, and 30, 1994.

6.2 Exit Meetings At periodic intervals and at the conclusion of the inspection, meetings were held with senior station management to discuss the scope and findings of inspections.

The exit meeting for inspection report 50-244/94-09 (corrective action programs, conducted May 2-6, 1994) was held by Mr. Edward Knutson on May 6, 1994.

The exit meeting for inspection report 50-244/94-13 (physical security, conducted May 9-13, 1994) was held by Mr. Arthur Della Ratta on May 13, 1994. The exit meeting for inspection report 50-244/94-14 (fire protection program, conducted May 9-13, 1994) was held by Ms. Leanne Kay on May 13, 1994.

The exit meeting for the current resident inspection report 50-244/94-12 was held on June 7, 1994 and was attended by the Chief of Projects, Branch 3, during a routine management oversight site visit.