IR 05000244/1998001

Insp Rept 50-244/98-01 on 980105-0221.No Violations Noted. Major Areas Inspected:Operations,Engineering,Maintenance & Plant Support.Also Includes Results of Licensee Fire Protection & Corrective Action Programs

ML17265A195
Person / Time
Site:	Ginna
Issue date:	03/13/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML17265A194	List: IR 05000244/1998001 ML17265A193
References
50-244-98-01, 50-244-98-1, NUDOCS 9803190419
Download: ML17265A195 (55)

Text

U.S. NUCLEAR REGULATORYCOMMISSION

REGION I

License No.

Report No.

Docket No.

Licensee:

Facility Name:

DPR-18 50-244/98-01 50-244 Rochester Gas and Electric Corporation tRGS.E)

R. E. Ginna Nuclear Power Plant Location:

1503 Lake Road Ontario, New York 14519 Inspection Period:

Inspectors:

Approved by:

January 5 through February 21, 1997 P. D, Drysdale, Senior Resident Inspector C. C. Osterholtz, Resident Inspector R. L. Fuhrmeister, Senior Reactor Engineer B. S. Norris, Senior Resident Inspector, Nine Mile Point R. C. Ragland, Radiation Specialist L. T. Doerflein, Chief Projects Branch

Division of Reactor Projects 9803i904i9 9803i3 PDR ADOCK 05000244 G

PDR

EXECUTIVE SUMMARY R.E. Ginna Nuclear Power Plant NRC Inspection Report 50-244/98-01 This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support.

The report covers a 7-week period of resident inspection and includes the results of announced inspections in the licensee's fire protection and corrective action programs.

~Oerations Plant operators responded quickly and effectively to equipment deficiencies during the inspection period and those deficiencies did not significantly effect safe plant operation.

Overall, the Ginna Station was operated well and has sustained safe operational performance for an extended period.

The licensee has made progress in identifying and resolving plant and equipment configuration control problems.

Although plant configuration control problems were ongoing, the licensee's generation of an encompassing ACTION Report, their performance of system line-up verifications, and their planned case study training were positive initiatives toward resolving a general configuration control problem.

The equipment out-of-service (EOOS) computer training effectively initiated operations personnel in the basic use and operation of the programs designed to calculate probabilistic risk factors and to record the operational status of plant equipment.

The licensee's corrective action program was generally effective in identifying and correcting conditions adverse to quality, and in preventing recurrences.

However, some program weaknesses were identified which could affect the overall usefulness of the system; most notably was the practice of closing ACTION Reports before the corrective actions were completed.

Root cause evaluations were thorough, technically well-based, and aided in the identification of the appropriate corrective actions.

In addition, housekeeping and material condition of the station was generally good, but several deficiencies were identified by the inspectors.

The peer-assisted self-assessments were good and included a strong independent perspective.

Numerous findings for improvement were identified and evaluated, and valuable corrective actions and program enhancements resulted from the assessments.

However, the overall quality and value of the assessments would have been greater had the operations and maintenance organizations more fullyparticipated in the process.

The licensee effectively identified applicable industry events, and appropriately incorporated those events in the corrective action system.

The operating experience inputs to the morning management meeting routinely provided worthwhile information to the plant staf Executive Summary (cont'd)

The plant operations review committee, and the nuclear safety audit and review board meetings were effective and conducted in accordance with Improved Technical Specification requirements.

The quality assurance/quality control subcommittee effectively analyzed audits to identify areas for'improvement in the corrective action program.

The quality assurance staff was effective in identifying deficiencies.

However, the inspectors were concerned that some quality assurance auditors had difficultyassessing.

human performance issues.

Maintenance Maintenance and surveillance activities were performed in accordance with procedural, requirements.

Plant equipment received adequate post-maintenance, testing prior to its return to service.

The licensee practiced good personnel and plant safety practices.

The as-found and as-left test data met the expected performance values and the acceptance criteria stated in the Updated Final Safety Analysis Report.

The replacement of the recirculation line air operated valve (AOV) controllers in the auxiliary feedwater (AFW) system should effectively resolve previously noted problems with AFW recirculation line reliability. Instrumentation and control technicians displayed a good working knowledge of component operation and installation.

However, the

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procedure used to replace the controller was somewhat vague in that some specific instructions needed to perform the replacement were not included.

Enraineering The licensee's discovery that Boraflex degradation had occurred in the spent fuel pool (SFP) was timely and should permit corrective actions prior to the 1998 SFP rerack.

The licensee's immediate corrective actions to ensure subcriticality and their intention to perform analysis to determine longer term corrective actions were appropriate.

The inspectors concluded that the Maintenance Rule training for system engineers was effective with good student participation on the subject matter.

The written training material contained good information that should effectively aid the systems engineers in their assessments of system performance.

Plant Su ort Trending records showed that the number of contamination boundary control deficiencies had significantly decreased.

The licensee's response to violation 50-244/97-01-02was good, and this item w'as closed (VlO 50-244/97-01-02).

The fire protection program has been effective in maintaining the integrity of fire barrier penetration seals, and good controls of combustible materials were developed and implemente 'ABLEOF CONTENTS EXECUTIVE SUMMARY TABLE OF CONTENTS IV I. Operations

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Conduct of Operations..........

01.1 General Comments.................................

01.2 Summary of Plant Status Operator Knowledge and Performance..

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04.1 (Updated) IFI 50-244/97-10-01:Weak Configuration Control Operator Training and Qualification 05.1 Equipment Out-of-Service (EOOS) and Autolog Training,......

Quality Assurance in Operations 07.1 Corrective Action Program 07.2 Self-Assessments 07.3 Operational Experience Feedback 07.4 Onsite and Offsite Safety Review Committees 07.5 Quality Assurance Audits and Surveillances 07.6 Review of INPO Evaluation Miscellaneous Operations Issues......

08.1 (Closed) URI 50-244/97-02-01: Intermediate Range Nuclear Instrument Operability 08.2 (Closed) LER 97-006, Revision 1: Verification of Boron-Concentration Not Performed Due to Misinterpretation of Event Sequence, Resulted in Condition Prohibited by Technical Specifications

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08.3 (Closed) LER 97-007, Revision 1: Reactor Trip Instrumentation Would Have Been in a Condition Prohibited by Technical Specifications 08.4 (Closed) LER 96-003, Revision 1: Both Pressurizer Relief Valves Inoperable, Due to Weakness in Procedure Change Process, Results in Condition That Could Have Prevented Fulfillment of a Safety Function

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II. Maintenance

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M1 Conduct of Maintenance........

M1.1 General Comments on Maintenance Activities...

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M1.2 General Comments on Surveillance Activities M2 Maintenance and Material Condition of Facilities and Equipment...

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M2.1 Auxiliary Feedwater (AFW) Pump Recirculation Line Air Operated Valve (AOV) Controller Replacement

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E2 Engineering Support of Facilities and Equipment E2.1 Spent Fuel Pool Boraflex Degradation

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E2.2 Charging System Performance...

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Table of Contents (cont'd)

E5 Engineering Staff Training and Qualification.....................

E5.1 Maintenance Rule Training for System Engineers.... ~........

IV. Plant Support...........

R8 Miscellaneous RP&C Issues..

R8.1 (Closed) VIO 50-244/97-01-02:

Inadequate Control of Radiological Boundaries F2 Status of Fire Protection Facilities and Equipment F2.1

.Fire Suppression System Walkdown............

F2.2 Fire Barrier Penetration Seals........

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F2.3 Fire Hose Hydrostatic Tests F2 4 Fire Main Loop Repairs..........................

F3 Fire Protection Procedures and Documentation F3.1 Combustible Material Control....... ~........

F3.2 Hotwork Controls...

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F3.3 Station Staff Responsibilities for the Fire Protection Program F7 Quality Assurance in Fire Protection Activities F7.1 Quality Assurance Audits of Fire Protection Program.....

F8 Miscellaneous Fire Protection Activities F8.1 ACTION Reports on Fire Protection Discrepancies

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X3.1 Deputy Division Director Visits L2 Review of UFSAR Commitments

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ATTACHMENTS Attachment 1 - Partial List of Persons Contacted

- Inspection Procedures Used

- Items Opened, Closed, and Discussed

- List of Acronyms Used

Re ort Details I. 0 erations

Conduct of Operations'1

~ 1 General Comments Ins ection Procedure IP 71707 The inspectors observed plant operations to verify that the facility was operated

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safely and in accordance with licensee procedures and regulatory requirements.

These reviews included tours of the accessible areas of the facility, verification of engineered safeguards feature (ESF) system operability, verification of proper control room and shift staffing, verification that the plant was operated in conformance with the improved technical specifications (ITS) and appropriate action statements for out-of-service equipment were implemented, and verification that logs and records accurately identified equipment status or deficiencies.

01.2 Summar of Plant Status a.

Ins ection Sco e (71707)

The inspectors monitored the status of plant power, the availability of plant safety systems, and the condition of plant equipment on an ongoing daily basis.

b.

Observations and Findin s The plant was operating at approximately 100% power at the beginning of the inspection period.

On January 8, 1998, equipment restoration procedure ER-SC.2,

"High Water (Flood) Plan," was entered for approximately four hours as a precautionary measure due to heavy rains that caused an unusually high water level in the stream (Deer Creek) directly adjacent to the plant.

During January 14-16, 19-21, and 26-28, 1998, plant power was reduced to 98%

while the licensee performed calibrations of reactor protection channels one, three, and four, respectively.

The power reductions were performed prior to the calibrations to provide an increased margin to reactor protection system setpoints while the calibrations were performed.

This action was taken as a precaution for ongoing intermittent spikes in the reactor coolant system (RCS) average temperature (Tave) caused by "streaming" of RCS fluid temperatures detected in channel two (see IR 50-244/97-12). Additionally, on January 18, 1998, power was reduced to 98% for approximately twelve hours during troubleshooting and repair of a failed power range channel (N41). The severity of the RCS temperature streaming later increased, and on February 21, 1998, operators placed the rod control system in the manual mode to prevent inadvertent rod motion due to the streaming affect.

Also, a rod control urgent failure occurred on February 14, 1998, due to a logic-Topical headings such as 01, M8, etc., are used in accordance with the NRC)

standardized reactor inspection report outline.

Individual reports are not expected to address all outline topic e

card failure. Following troubleshooting by IRC technicians, the logic card was replaced and the urgent failure was cleared the same day.

On January 21, 1998, the licensee performed a flush of the residual heat removal (RHR) system in accordance with procedure 2.2Q, "Residual Heat Removal System-Quarterly."

The flush significantly reduced high local radioactivity levels in the RHR loops that had increased during the recent refueling outage as a result of failed fuel during the previous operating cycle.

On January 27, 1998, a leak test of the technical support center (TSC) diesel fuel oil tank indicated a small loss of tank level after filling, The licensee subsequently performed a second leak test on February 4, 1998. That test also failed, and the licensee notified the New York State Department of Environmental Conservation of a potential spill in the ground surrounding the tank.

The licensee also issued a 10CFR50.72 four notification to the NRC. Troubleshooting revealed a crack in the

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vent fitting line into the tank.

The leak was subsequently repaired and the tank satisfactorily passed a leak test on February 10, 1998.

On February 4, 1998, the B-charging pump was taken out of service due to a failed speed increaser bearing.

The A-charging pump was also out of service for scheduled maintenance at the time, and the licensee entered a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> limiting condition for operation (LCO) for two charging pumps out of service in accordance with the technical requirements manual section 3.1.1.A. The A-charging pump was returned to service after repairs were completed, and the LCO was exited on February 5, 1998. The licensee machined a new speed increaser bearing and the B-charging pump was returned to service on February 16, 1998.

During the performance of periodic test PT-12.1, "Emergency Diesel Generator A,"

on February 16, 1998, the A-emergency diesel generator (EDG) day tank

'verflowed through the tank's breather line due to a leaking check valve (CV-5960A) in the fuel oil transfer pump recirculation line. The licensee briefly entered ER-SC.5, "Hazardous and Mixed Waste Management and Control," in response to the fuel oil spill. However, only 4-5 gallons were lost, and all of the spillage was contained within the plant.

On February 17, 1998, the licensee placed offsite power in a 100%/0% lineup on offsite circuits 767/751 while tree trimming maintenance was performed in the vicinity of circuit 751. The 100%/0% lineup was maintained through the end of the report period.

Conclusions Plant operators responded quickly and effectively to equipment deficiencies during the inspection period, and the deficiencies did not significantly affect safe and stable plant operation.

Overall, the Ginna Station was operated well and has sustained safe operational performance for an extended perio Operator Knowledge and Performance 04.1 U dated IFI 50-244 97-10-01:Weak Confi uration Control a.

Ins ection Sco e (92901)

The inspectors reviewed the licensee's ongoing activities to identify and resolve plant system configuration control deficiencies.

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Observations and Findin s IFI 50-244/97-10was,opened after configuration control problems resulted in the inadvertent discharge of reactor makeup water into the waste holdup tank in September 1997. Additionally, the licensee identified multiple occurrences of configuration control problems during the recent refueling outage (see IR 50-244/97-11).

The licensee generated a comprehensive ACTION Report (98-0025) addressing human performance and configuration control issues that occurred during the fourth quarter of 1997, in order to determine whether any common cause contributed to these problems.

Their investigations into these issues were ongoing at the end of the inspection period.

On December 23, 1997, the licensee discovered valve V-348B (emergency borate inlet isolation valve to charging pumps) in the closed position when it should have been open.

This discovery was made while performing motor-operated valve stroke testing in accordance with periodic test PT-2.3, "Safeguard Motor Operated Valve Operation," and the licensee activated their security plan to determine if the mispositioned valve resulted from a malicious act.

The operations and security personnel concluded that the mispositioned valve was not a malicious event and was apparently the result of a human error prior to the periodic test.

The licensee returned V-348B to the open position and generated an ACTION Report (97-2207)

to further address the issue.

The licensee also performed a comprehensive system line up verification in accordance with all existing system (S) procedures.

No other valve position discrepancies were discovered during that verification.

The licensee's training department also generated a case study on the inadvertent discharge of reactor makeup water to.the holdup tank that occurred in September 1997. Training on the case study was scheduled to be presented to all operating crews during the next training cycle.

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Conclusions The inspectors concluded that the licensee has made progress in identifying and resolving plant and equipment configuration control problems.

Although plant configuration control problems were ongoing, the licensee's generation of an encompassing ACTION Report, their performance of system line-up verifications, and their planned case study training were positive initiatives toward resolving a general configuration control problem.

Pending further inspection in this area, this item will remain open (IFI 50-244/97-10-01).

Operator Training and Qualification 05.1 E ui ment Out-of-Service EOOS and Autolo Trainin a.

Ins ection Sco e (71707)

The inspectors attended operator training on a control room modification that installed a computerized system for tracking out-of-service equipment and keeping control room operations logs.

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Observations and Findin s On February 13, 1998, the licensee conducted training for operators on a modification that installed a computerized tracking system for equipment out-of-service (EOOS), and for keeping an electronic record of control room logs.

The training focused on the practical use of the computerized programs.

Each student followed along with the instructor on a computer terminal while various example

'situations involving out-of-service equipment were discussed.

The computer automatically kept track of the overall increase from the baseline risk (probabilistic risk factor) when equipment was removed from service.

The program provided color codes (green, yellow, orange, and red) to flag operators to different levels of risk to safe plant operation.

The operators also performed similar exercises with the autolog program for electronically logging control room data and activities.

Some operators noted minor problems with the autolog system that needed to be resolved.

For example, the system could not alert operators if a log entry did not agree with the current status of plant equipment.

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Conclusions The inspectors concluded that the EOOS computer training effectively initiated operations personnel in the basic use and operation of the programs designed to calculate probabilistic risk factors and to record the operational status of plant equipment.

Quality Assurance in Operations Using NRC Inspection Procedure 40500, the inspectors reviewed RGRE's programs related to the identification and resolution of issues that affect the quality of operations at the Ginna station.

The review was to evaluate the effectiveness of the programs for correcting immediate concerns, and preventing recurrence of similar deficiencies.

The programs reviewed included:

the corrective action program (section 07.1),

self-assessments (section 07.2),

incorporation of operational experience (s'ection 07.3),

on-site and off-site safety review committees (section 07.4), and quality assurance activities (section 07.5).

07.1 Corrective Action Pro ram a.

Ins ection Sco e

40500 The inspectors reviewed the RGSE corrective action program and its implementation relative to identification and correction of deficiencies, and actions to prevent recurrence.

The review included operability determinations, root cause analysis, and safety evaluations; in addition, the. inspectors toured the plant and interviewed station personnel.

b.

Observations and Findin s The high-level guidance for the Ginna corrective action program was contained in the Nuclear Directive for the corrective action program (ND-CAP) that was approved by the Vice President for Nuclear Operations.

The details of the program were contained in Interface Procedure IP-CAP-1, "Abnormal Condition Tracking I~itiation or Notification (ACTION) Report." The ACTION Reporting (AR) system at Ginna was relatively new, i.e., just over two years old; however, the program was based on several established corrective action programs at other nuclear facilities.

In general, the AR program appeared effective.

In accordance with ND-CAP, the Production Superintendent was responsible for the overall management of the program, and he reviewed each AR. Each weekday morning, most of the ARs initiated are discussed at the morning meeting and, as necessary, were reviewed in detail by the Plant Operations Review Committee (PORC)

~ In 1996, 1260 ARs were written; and in 1997, 2092 ARs were written." An intentional move was made in 1997 to lower the threshold for the initiation of ARs, to ensure that all conditions adverse to quality, or "near-misses" were captured in the system for identification and trending.

However, the inspectors identified some program weaknesses:

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The program allowed ARs to be closed without the corrective actions being completed, by transferring the corrective actions to one of the other Ginna processes; such as a plant change request (PCR), or procedure change notice (PCN), or the commitment and action tracking (CAT) system.

ND-CAP, paragraph 3.10.2, stated that corrective action documents shall be assigned unique numbers and shall be tracked to completion.

Discussions with the AR coordinator determined that the other processes were not tracked with the same rigor as ARs were.

The inspector was concerned that the other processes could allow deficient issues to remain open, or even be canceled, without senior management being aware.

RGRE indicated they would conduct a special QA audit to review this concern and determine if the practice has allowed corrective actions to be missed.

RGSE also stated they would complete the audit and provide written results to the NRC by the end of June 1998. The licensee had initiated the audit at the end of the current inspection period.

This item will be tracked as an inspector follow item (IFI 50-244/98-01-01).

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The AR procedure allowed for the AR to be assigned a priority that determines the required timeliness for the disposition, usually based on safety significance, and the depth of the root cause investigation.

The highest priority was "A"which required prompt attention, and normally required a formal root cause analysis.

The lowest priority were "D" ARs, which were intended to document near-.

misses or items of minimal significance.

Several of the Priority "D" ARs reviewed by the inspectors did not appear to be consistent with the original intent of the concept.

There were two examples where Priority "D" was applied to deviations from procedural requirements; however, the deficiencies were immediately corrected, and so noted on the AR.

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The inspectors were concerned that the day-to-day involvement of the Production Superintendent in the administration of the AR process may be detracting from his other duties associated with the oversight of the operations, chemistry, radiation protection, and other areas.

The Vice President, Nuclear Operations, informed the inspectors that this had already been recognized by the Production Superintendent and several options were being considered to reduce the amount of time he spent administering the AR system.

The inspectors conducted an independent tour of the facility and identified several deficiencies, most of which were of minor significance.

However, some plant areas exhibited material condition deficiencies, such as housekeeping in the B-emergency diesel generator room, rusted piping in the component cooling water system, and groundwater inleakage in the area by the volume control tank room.

In addition, the inspectors noted several long-standing conditions that had not been identified by RGSE and therefore not entered into any of the systems for correction; such as a tygon tube attached to the fire suppression system in the ventilation room, and a temporary power feed panel in the auxiliary building that had been explicitly excluded from the temporary modification program.

During the inspectors'ours of the plant, they noticed that many components had maintenance identification (MID)

tags attached to the equipment.

Discussions with the maintenance manager indicated that this has been recognized by RG&E, and they are in the process of reviewing all the existing MIDtags for either repair or acceptance of the condition as-is.

These items were also being addressed through the licensee's rotating work schedule.

Although the number of catch containments in the plant was not excessive, there was also no formal system to track or authorize the installation and removal of catch containments, nor to initiate repair of the affected components.

The general housekeeping and material condition of the Ginna station was good; however, several deficiencies were identified by the inspectors that were communicated to RG5E supervisors and managers for follow-up.

The inspectors reviewed the quality of the licensee's root cause evaluations.

The review consisted of interviews with cognizant personnel and review of two formal (Level A) root cause evaluations and five apparent cause evaluations.

The inspectors'eviews were based on the ACTION Report procedure (IP-CAP-1), and

Root Cause Analysis procedure (IP-CAP-2). Procedural guidance outlined in!P-CAP-1 required the cause of abnormal events or conditions to be determined.

Procedure guidance outlined in IP-CAP-2, required additional cause investigations to be performed for significant issues.

Procedural guidance was clear, and provided adequate administrative guidance for the performance of root cause and apparent cause investigations, Formal root cause investigations were required for issues considered to have a high impact such as a significant condition adverse to quality; issues with lesser impact received an apparent cause evaluation.

Of the ARs written in 1997, two formal root cause investigations and eight apparent cause investigations were performed.

The inspectors considered that root cause and apparent cause evaluations were thorough, technically well based, and resulted in identification of root cause(s)

and corrective action recommendations.

Conclusions The corrective action program at Ginna appeared effective in identifying and correcting conditions adverse to quality, and generally in preventing them from recurring.

However, some weaknesses were identified, which could affect the overall usefulness of the system; most notably was the practice of closing ARs before the corrective actions were completed (IFI 50-244/98-01-01).

Root cause evaluations were thorough, technically well based, and aided in the identification of the appropriate corrective actions.

In addition, housekeeping and material condition of the station was generally good, but several deficiencies were identified by the inspectors.

Self-Assessments Ins ection Sco e 40500 The inspectors performed a review of the use and effectiveness of self-assessments.

Information was gathered by a review of several peer-assisted self-assessments and the licensee's responses to the associated findings. The inspectors interviewed Ginna personnel, including department managers and a self-assessment team leader, and reviewed responses to self-assessment findings.

Observations and Findin s In an attempt to enhance overall performance, RGBcE senior management committed the major organizations involved with the operation and support of the Ginna station to perform a series of peer-assisted self-assessments using criteria selected from an industry standard (the Institute of Nuclear Power Operations Publication 96-006,

"Performance Objectives and Criteria for Operating Nuclear Electric Generating Stations" ).

In early 1997, teams were assembled and self-assessments were conducted in the functional areas of operations, maintenance, engineering, radiation protection, and chemistry.

Also, assessments were conducted in the "softer" areas of safety culture, self-evaluation, human performance, and organizational effectiveness.

The assessment results identified programmatic strengths and

weaknesses, and were described to senior management in a self-evaluation report.

Overall, the inspector found the quality of the assessments to be very good.

The self-assessments included industry peers, were generally well written and critical, and contained recommendations for improvement.

Line organizations responded to the self-assessments by evaluating the findings and recommendations, implementing corrective actions or program enhancements, and, if necessary, initiating ARs.

Examples of program enhancements initiated as a result of the self-assessments included:

the operations group lowered the threshold for managing reactivity changes; the shift supervisor became more involved in plant meetings such as the morning priorities action required (MOPAR) meeting and plant operations review committee (PORC) meeting; the radiation protection and chemistry organizations increased the use and publication of station goals; and the maintenance department initiated various plant cleanups and painting projects.

• The inspectors noted that the operations group maintained a self-monitoring program that appeared to gain performance insights and improve performance.

However, both the operations and maintenance organizations did not provide a team member for their respective assessments; without support from the line organizations, the inspector considered that some recommendations were not well founded.

As a result, there was less acceptance of the self-assessment findings/recommendations.

In addition, the organizations missed a training opportunity for the performance of self-assessments.

The inspectors were informed that full support was not provided primarily due to a lack of available personnel at the time of the self-assessment.

However, the operations and maintenance managers revealed that some of their staff perceived the peer-assist as an "outside-audit" rather than as a true self-assessment, and that the assessments were conducted only to identify weakness rather than as a process or tool to improve performance.

c.

Conclusion The inspector concluded that the peer-assisted self-assessments were good; they included a strong'independent perspective, numerous findings for improvement were identified and evaluated, and valuable corrective actions and program enhancements resulted from the assessments.

However, the overall quality and value of the assessments would have been greater had the operations and maintenance organizations more fullyparticipated in the proces.3 0 erational Ex erience Feedback Ins ection Sco e (40500)

The inspectors reviewed the Ginna program for identifying industry events and disseminating that information to the plant staff.

b.

Observations and Findin s The inspectors reviewed Interface Procedure IP-SEP-4, "Operating Experience Program," and the administrative procedures and guidelines that provided direction for processing information on industry operational experience.

The licensee collected information from vendor reports, the industry nuclear network and significant operating experience reports, daily reports for other facilities, and NRC Bulletins, Generic Letters, and Information Notices.

The events were prioritized by determining their significance and their applicability to Ginna.

Routine events were tracked using the Ginna Commitment and Action Tracking System (CATS)

procedure (IP-LPC-1). However, significant events that were applicable to Ginna were tracked using the AR system.

Once events were identified, a training committee, with members from the applicable departments, screened the events for determination of appropriate training. The inspectors noted that a representative from the operational experience group discussed the significant industry events during the daily MOPAR meeting.

c.

Conclusions The inspectors concluded that the licensee effectively identified applicable industry events, and appropriately incorporated those events in the corrective action system.

The operating experience inputs to the morning management meeting routinely provided worthwhile information to the plant staff.

07.4 Onsite and Offsite Safet Review Committees Ins ection Sco e (40500)

C The inspectors attended meetings of the onsite plant operations review committee (PORC) and the offsite nuclear safety audit and review board (NSARB). Information was gathered by a review of the applicable procedure, the technical specifications, and interviews with members of the review committees.

b.

Observations and Findin s Administrative procedure A-205, "Plant Operations Review Committee Operating Procedure," provided a clear structure and guidance for PORC membership and quorum requirements, responsibilities and authority, conduct of meetings, review process, documentation of minutes,-and required effectiveness reviews.

The PORC chairman ensured compliance with procedural requirements, ensured that the

significance of issues was evaluated, and kept the meeting running smoothly.

PORC members exhibited a questioning attitude, healthy discussions were observed, and good insight was observed during their reviews of ARs.

PORC documentation, including agenda and meeting minutes were well detailed, and the inspector verified that self-assessments of PORC were performed.

The inspectors attended portions of the NSARB meeting conducted on January 28-29, 1998. The board membership consisted of RGSE management, and management representatives from the Dresden and Nine Mile Point nuclear stations.

Interface procedure IP-NPD-6, "Nuclear Safetygudit and Review Board," described the function of the NSARB to direct the establishment of an audit program for safe station operation.

The board discussed corrective actions associated with the QA program as identified in the quality assurance/quality control (QA/QC) subcommittee meeting and the current NRC corrective actions inspection.

The board concluded that more attention needed to be placed on ensuring that all holds (tag-outs) sent to work control were appropriately filled out. Additionally, the board determined that AR titles could be improved to better indicate the event they were describing.

The board also noted that cleanliness in the containment during the refuelihg outage did not meet management expectations, and more attention needed to be placed on housekeeping.

The inspectors noted that the representatives from the other facilities had a more questioning attitude and focused their questions at a higher level. They were also able to bring insights on similar problems from their own facilities. The board discussed the area of QA auditor frustration when inspecting

"soft" issues, such as human performance (see Section 07.5). The board concluded that QA auditors should be encouraged to raise issues that identify areas for improvement, but that management reserved the right to make interpretations when necessary.

The QA manager agreed with the board.

Overall, the presentations made to the board, and its evaluation of agenda items, were detailed and comprehensive.

Board members generally exhibited a questioning attitude that contributed to the discussions of root cause analyses.

At the end of the formal discussions, the NSARB chairman requested feedback from the board members as to the performance of the board during this meeting.

The inspectors considered the discussions and evaluations during the NSARB meeting to be thorough and critical.

As part of the evaluation of the NSARB, the inspectors attended the quarterly QA/QC subcommittee meeting.

Topics discussed included the corrective action program, performance trending, and QA audits, The discussions focused on potential solutions to deficiencies identified in recent QA audits, including a computer program that was under development to identify trends in human performance problems.

The committee also discussed reducing the number of cause codes used in the AR system, such that similar performance and equipment problems would be more readily identified for repeat occurrences.

The subcommittee concluded that the threshold for problem identification had been appropriately lowered over the past year, based on the fact that the number of ARs significantly increased from 1996 to 199 c.

Conclusions The inspectors concluded that PORC and NSARB meetings were effective and conducted in accordance with ITS requirements.

The QA/QC subcommittee

, effectively analyzed QA audits to identify areas for improvement in the corrective action program.

07.5 Qualit Assurance Audits and Surveillances a.

Ins ection Sco e (40500)

The inspectors reviewed several QA audits for problem identification findings.

b.

Observations and Findin s The inspectors reviewed the QA audits of the corrective action program performed in 1996 and 1997. The inspectors noted that the auditors identified that the AR process lacked a tracking and trending program, and that AR cause codes were not consistently applied for similar deficiencies.

For example, the audit referred to an instance in which four different cause codes were applied to similar failures of the electronic dosimetry used in the radiation protection department.

The inspectors also noted that resolutions to these deficiencies had been discussed at the QA/QC subcommittee meeting.

The audits revealed that the QA inspectors performing the audits were having difficultyevaluating "soft" issues, such as human performance.

One of the audit reports implied that management expected a certain number of human performance errors to occur; but without an established goal, the effectiveness of the corrective action could not be accurately assessed.

It appeared to the inspectors that QA auditors were more concerned about auditing to management expectations instead of to procedures and established standards, c.

Conclusions The inspectors concluded that the quality assurance staff was effective in identifying deficiencies.

However, the inspectors were concerned that some QA auditors had difficultyassessing the "softer" human performance issues.

07.6 Review of INPO Evaluation 71707 The inspectors reviewed the report from the Institute of Nuclear Power Operations (INPO) for the onsite evaluation of Ginna activities, conducted from May 5 through 16, 1997.

The evaluation examined the overall operation of the Ginna site, and was performed by peer evaluators from other nuclear facilities. The report identified no issues that the NRC was not already aware of, and no additional follow-up by the NRC is warrante Miscellaneous Operations Issues 08.1 Closed URI 50-244 97-02-01:Intermediate Ran e Nuclear Instrument 0 erabilit URI 50-244/97-02-02was opened on May 5, 1997, after the licensee discovered that a reactor startup was performed on August 21, 1996, with an improperly compensated intermediate range nuclear instrument channel (see IR 50-244/97-02).

Of specific concern to the inspectors was the possibility that improved technical specification (ITS) operability requirements for the intermediate range instrument, (N-36) as well the permissive P-6,.may have been violated.

The ITS operability requirements for N-36 and P-6 were reviewed by the Office of Nuclear Reactor Regulation (NRR) instrumentation and control (ISC) staff for applicability. The NRR staff concluded that although the undercompensated intermediate range channel and the P-6 interlock exhibited degraded performance, the appropriate ITS required actions were met, and no violations of ITS requirements occurred.

The NRR staff also concluded that starting up the plant with an undercompensated intermediate range channel was not desirable; however, it did not violate the plant's licensing basis.

During a review of the updated final safety analysis report (UFSAR), the inspectors discovered that section 7.2.2.4.5, "P-6 Permissive," stated that "Ifboth intermediate range channels drop below 1 X 10'~ amps, the permissive will automatically be defeated."

This was contrary to the actual setpoint of 5 X 10" amps and ITS bases 3.3.1 which stated "On decreasing power, the P-6 interlock automatically energizes the NIS source range detectors and enables the Source Range Neutron Flux reactor trip at 5 X 10" amps."

The licensee indicated that the UFSAR was incorrect, and initiated a UFSAR change notice (14-058) to correct the deficiency.

Based upon all of the above actions, this item is closed (URI 50-244/97-02-01).

08.2 Closed LER 97-006 Revision 1: Verification of Boron Concentration Not Performed Due to Misinter retation of Event Se uence Resulted in Condition Prohibited b Technical S ecifications LER 97-006, Revision 1, was submitted to the NRC on February 6, 1998, after the licensee discovered that an ITS requirement to verify boron concentration following an inoperability of the source range audible count rate may not have been performed within the required four hour time limit (see IR 50-244/97-12).

The licensee determined that control room operators did not declare the audible count rate instrument inoperable at the proper time. The licensee reviewed the event sequence with the shift supervisor, and subsequently generated a training work request addressing the need to apply the most conservative LCO requirements when equipment is removed from service.

Based on in-office review and in-plant follow-up, the inspectors concluded that the LER revision adequately described additional follow-up corrective actions for this event.

This LER is closed (LER 97-006, Revision 1) ~

08.3 Closed LER 97-007 Revision 1: Reactor Tri Instrumentation Would Have Been in a Condition Prohibited b Technical S ecifications LER 97-007, Revision 1, was submitted to the NRC on February 6, 1998, after the licensee discovered that there was no requirement in instrumentation and control procedures to place the neutron flux low range trip circuitry in the tripped condition for a power range channel removed from service for physics testing in MODE 2 as required by the ITS (see IR 50-244/97-12).

The licensee made the discovery while in MODE 3, and thus prevented any ITS discrepancy.

The LER indicated that instrumentation procedures would be revised to ensure the neutron flux low range trip circuitry would be placed in the tripped condition when required by the ITS.

Additionally, the licensee generated a training work request to sensitize operations personnel on the need to place affected trip circuitry in the tripped condition when necessary.

The LER also described previous occurrences where physics testing was conducted in MODE 2 without the low range trip circuitry in the tripped condition.

Based on in-office review and in-plant follow-up, the inspectors concluded that the LER adequately described the event, and appropriately addressed the root causes and corrective actions.

The licensee's instrumentation and control procedures did not appear to satisfy the requirements of 10CFR50, Appendix B, Criteriorl V,

"Instructions, Procedures, and Drawings." However, since this item was identified by the licensee and corrective actions initiated, this item is a Non-Cited Violation in accordance with Section VII.B.1 of the NRC Enforcement Policy and is closed (NCV 50-244/98-01-02).

This LER is also closed (LER 97-007, Revision 1).

08,4 Closed LER 96-003 Revision 1: Both Pressurizer Relief Valves Ino erable Due to Weakness in Procedure Chan e Process Results in Condition That Could Have Prevented Fulfillment of a Safet Function LER 96-003 was submitted to the NRC on April 8, 1996, and Revision 1 was submitted on May 17, 1996, after the licensee discovered that they had authorized work that physically rendered both pressurizer power operated relief valves (PORVs)

inoperable.

This event also resulted in a notice of violation (VIO 96-01-01). The licensee's corrective actions taken in response to the notice of violation included procedural enhancements and training for all plant operators.

Based on extensive in-plant review of this event, the inspector subsequently closed the notice of violation (see IR 50-244/97-11).

Since the licensee's corrective actions to address the notice of violation were consistent with the identified root causes and corrective actions identified in the LER, this LER is closed (LER 96-003, Revision 1) ~

II. Maintenance M1 Conduct of Maintenance M1.1 General Comments on Maintenance Activities a 0 Ins ection Sco e (62707)

The inspectors observed portions of plant maintenance activities to verify that the correct parts and tools were utilized, the applicable industry codes and technical specification requirements were satisfied, adequate measures were in place to ensure personnel safety and prevent damage to plant structures, systems, and components, and to ensure that equipment operability was verified upon completion of post maintenance testing.

b.

Observations and Findin s The inspectors observed all or portions of the following work activities:

Emergency Diesel Generator (EDG) diagnostic testing; observed on January 22, 1998 A-AFW pump recirculation line air operated valve controller replacement; observed on February 3, 1998 (see section M2.1)

Repairs to relief valve 4654 (service water outlet from B-component coolirig water heat exchanger); observed on February 13, 1998 c.

Conclusions The inspectors concluded that the observed maintenance activities were performed in accordance with procedural requirements.

Equipment received adequate post-maintenance testing prior to its return to service.

Good personnel and plant safety practices were observed during the maintenance work.

M1.2 General Comments on Surveillance Activities a.

Ins ection.Sco e (61726)

The inspectors observed two surveillance tests to verify that approved procedures were used, procedure details were adequate, test instrumentationwas properly calibrated and used, technical specifications were satisfied, testing was performed by knowledgeable personnel, and test results satisfied acceptance criteria or were properly dispositione b.

Observations and Findin s The inspectors observed portions of the following surveillance activities:

~

PT-12.1, "Emergency Diesel Generator A"; observed on January 22, 1998

~

PT-12.2, "Emergency Diesel Generator B"; observed on January 22, 1998 C.

Conclusions The inspectors confirmed that the procedures used were current and properly followed. The shift supervisor properly authorized the surveillance work to proceed.

The licensee confirmed the qualifications of the surveillance test personnel involved in the tests.

Both EDGs met all of the acceptance criteria specified in the test procedures for operability.

M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Auxiliar Feedwater AFW Pum Recirculation Line Air 0 crated Valve AOV Controller Re lacement Ins ection Sco e (62707)

The inspectors observed instrumentation and control (l&C) personnel install a new controller for the A-AFW pump recirculation line AOV.

b.

Observations and Findin s The controller replacement took place on February 3, 1998. The reliability of the AFW recirculation line AOV controller had been a recurring problem in the recent past, and had also been designated as an operator workaround (see IR 50-244/97-06).

The B-AFW recirculation line AOV controller had also experienced similar functional problems and was replaced on December 30, 1997 (see IR 50-244/97-12).

The inspectors reviewed procedure CPI-CV-4304, "Calibration of Auxiliary Feedwater Pump A Recirculation Air Operated Valve 4304 and Pressure Controller PC-2033." Attachment 1 to that procedure provided instructions for the controller replacement; however, the inspectors noted that the attachment did not contain specific instructions to perform some parts of the replacement.

For example, step 1 stated, "Adjust air to diaphragm of valve 4304 to desired position,"

and step 5 stated, "Repair copper air lines to valve 4304 as necessary."

The IRC technicians considered that step 1 meant to isolate air to valve 4304, and that step 5 meant to ensure no kinking or deformities were present in the installed copper tubing. The inspectors observed the technicians perform bench testing of the new controller in accordance with vendor manual instructions, then observed the actual installation in the A-AFW system.

Although some of the steps in Attachment 1 of CPI-CV-4304 appeared vague, the technicians demonstrated a high level of expertise during the controller replacement.

After installation, the controller was

tested satisfactorily during the performance of PT 16-Q-A, "AFW Pump A-Quarterly."

C.

Conclusions The inspectors concluded that the replacement of the recirculation line AOV controllers in the AFW system should effectively resolve previously noted problerrls with.AFW recirculation line AOV operability.

The observed IRC technicians displayed a good working knowledge of component operation and installation.

However, the procedure used to replace the controller was somewhat vague in that specific instructions for performing the replacement were not included, relying instead on technician expertise to successfully accomplish the work, III. En'neerin E2 Engineering Support of Facilities and Equipment E2.1 S ent Fuel Pool Boraflex De radation a.

Ins ection Sco e (62707)

The inspectors reviewed the licensee's response to observed degradation in installed Boraflex panels in the spent fuel pool (SFP).

b.

Observations and Findin s On February 9, 1998, the licensee discovered that some Boraflex panels were significantly degraded while performing special testing specifically designed to examine the integrity of the panels.

The criticality analysis for the SFP assumed a

maximum gap of four inches in each Boraflex panel

~ The licensee discovered that some panels had experienced significantly more degradation, i.e., up to one hundred inches.

However, the degradation was not uniformly distributed among the SFP storage cells.

Some panels (eight of the first sixteen analyzed) showed no degradation at all ~ The panels that did exhibit degradation had all experienced gamma exposures greater than 2 4 X 10~ rads.

Based upon the observed degradation, the licensee declared the spent fuel pool inoperable and issued a one hour notification to the NRC in accordance with 10 CFR 50.72 section (b)(1)(ii)(B).

Although there was no ITS limit on gap size in the Boraflex panels, ITS 3.7.13 required that assemblies stored in Region 1 of the SFP (fresh fuel) have a K-infinity less than or equal to 1 458, and that assemblies in Region 2 be above the levels specified in ITS Table 3.7.13-1.

The licensee confirmed that both of these conditions were met, and that no ITS LCO applied.

However, the degradation of Boraflex by more than 4 inches in local fuel racks invalidated the assumptions of the SFP criticality analysis, and the licensee could not positively confirm that the entire pool was within the limits of the current analysis.

Additional analysis will be required to account for the missing Boraflex, including untested regions with known

exposures greater than 2.4 X 10'ads, to show that the requirement to have K,<< in all regions less than 0.95 is satisfied.

The NRC does not allow Ginna to take credit for dissolved boron in the SFP analysis, and their re-analysis must also make that assumption.

The licensee currently considers the pool to be in a safe condition because of the existing high level of dissolved boron, greater than 2500 parts per million (ppm). The licensee considered that there is no design basis event that could cause the boron concentration to go to zero.

A catastrophic failure of the weir gate could reduce the water level.by several feet, but would not reduce the boron concentration.

The licensee incorporated several short term compensatory measures to assure that the SFP remained in a safe condition. These included maintaining the water temperature less than 70 degrees Fahrenheit (as permitted by lake temperature),

maintaining the SFP water above the low level alarm point (20 inches below the pool ledge), and trending the weekly sample of boron concentration.

Longer term measures will be considered that include a major revision to the SFP criticality analysis after the 1998 rerack modification, removing fuel from racks that are known to have severely degraded Boraflex, reinserting rodded assemblies into the racks that have shown severe degradation in the current tests, and installing poison insert rods into high exposure assemblies.

The licensee indicated that an interim analysis would be performed to further confirm that the SFP is not in an unanalyzed condition, and that continued operability is justified.

Conclusions The licensee's discovery that Boraflex degradation had occurred in the spent fuel pool was timely and should permit corrective actions prior to the 1998 SFP rerack.

The licensee's immediate corrective actions to ensure subcriticality and their intention to perform analysis to determine longer term corrective actions were appropriate.

Char in S stem Performance Ins ection Sco e (40500)

The inspectors interviewed the system engineer for the charging system and reviewed the system's recent maintenance history and corrective actions for performance problems.

Observations and Findin s The inspectors reviewed seven ARs generated during the past year regarding the charging system performance and maintenance problems.

Specifically, the A-charging pump experienced leakage that could have occurred due to excessive vibrations in the charging system suction and discharge lines. Additionally, all three charging pumps appeared to have undergone packing replacements due to excessive primary leakage through the pump packing.

The inspector interviewed the system engineer (SE) for the charging system.

The SE indicated that much

P

progress had been made in the charging system in the past year, citing that pump packing performance had not deteriorated since the last outage.

Previously, the pumps required repacking after about 9,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of operation; currently, pump repacking is not expected to be needed until at least 14,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of operation.

The SE indicated that system piping vibrations have been reduced (though not eliminated), by varying the pump speeds to minimize resonance between the pumps.

Although not a formal self-assessment, the inspector walked down the charging system with the SE.

The SE was able to point out the location of all the major system components and was aware of the status of all out-of-service equipment.

The SE stated that his goal was to walk down his system twice per week.

C.

Conclusions The inspectors concluded that charging system deficiencies were appropriately identified, and charging system performance was considered to have improved over the past year.

Also, the system engineer exhibited a good knowledge on the operation and status of his system.

E5 Engineering Staff Training and Qualification E5.1 Maintenance Rule Trainin for S stem En ineers Ins ection Sco e (37551)

The inspectors attended the licensee's training on the Maintenance Rule (10CFR50.65) for system engineers.

b.

Observations and Findin s The licensee conducted Maintenance Rule training for system engineers on February 10, 1998. The purpose of the training was to familiarize system. engineers with the purpose of the Maintenance Rule, focusing on performance data and risk analysis to monitor and evaluate the effectiveness of maintenance on their systems.

Emphasis was also placed on the importance of system engineers to have ownership of their systems.

Allfunctional failures of safety related equipment, as well as non safety related equipment important to safety, were intended to be documented using the ACTION Report system and routed to the appropriate system engineer for resolution.

Each system engineer received a binder containing Maintenance Rule basis documentation for their respective systems.

The binders also contained information on Maintenance Rule scoping requirements, performance criteria, and definitions regarding probabilistic safety assessments.

The inspectors noted that the engineers in attendance directly participated in the training as material was presented, routinely asking questions and providing examples from their own experiences.

However, some engineers indicated that it may be difficultfor them to distinguish between functional failures and maintenance preventable functional failure c.

Conclusions The inspectors concluded that the Maintenance Rule training for system engineers was effective with good student participation on the subject matter.

The written training material contained good information that should effectively aid the systems engineers in their assessments of system performance.

IV. Plant Su ort R8 Miscellaneous RP8cC Issues R8.1 Closed VIO 50-244 97-01-02: Inade uate Control of Radiolo ical Boundaries (92904)

In February 1997, the NRC observed that the licensee had produced ineffective corrective actions in response to long standing and ongoing poor radiological work practices and contamination boundary controls that led to several incidents the uncontrolled release of contaminated materials to uncontaminated areas (see IR 50-244/97-01 ).

The inspectors reviewed RGSE's actions in response to the violation. The licensee revised rules for work in and around radiological boundaries, including: training to clarify expectations for performance, and interactive and mockup training to improve communications and work practices.

In addition, a causal analysis was performed to identify contributing factors and develop corrective actions, personnel accountability was enforced during the recent refueling outage, and effectiveness reviews were performed.

Also, plans were made to emphasize boundary controls during future outages.

Trending records showed that the number of boundary control deficiencies had significantly decreased.

The overall response to this violation was adequate; therefore, this item is closed (VIO 50-244/97-01-02).

F2 Status of Fire Protection Facilities and Equipment F2.1 Fire Su ression S stem Walkdown ao Ins ection Sco e (64704)

The inspector performed a walkdown of the fire suppression systems for the relay room and "mux" cabinet room, both independently and in company with the fire protection engineer (FPE).

The inspector observed the conditions of the agent supplies, piping and discharge heads for both the Halon 1301 and water suppression systems.

b.

Observations and Findin s The piping for both Halon systems was readily accessible and appeared to be in good repair.

The discharge heads for both systems were unobstructed.

The Halono

nozzles were located in the overhead of the room, while the water discharge heads were located in the room overhead and in proximity to the cable trays.

The manual actuation stations for both systems were located outside the relay room, on the 271'- 0" elevation of the turbine building.

Both actuation stations were clearly marked to show the system identification and the area the system protected.

The inspector verified that the pressure gages on the Halon bottles all indicated

.

within the allowable range.

The FPE pointed out to the inspector that an additional larger bottle had been added to the Halon system bank, with an orifice to provide a slow bleed of the agent into the room after system activation.

This was installed to resolve a'slow dissipation of the agent identified during initial acceptance testing.

Conclusions Based on the observed conditions, and discussions with the FPE, the inspector concluded that the suppression systems were in good repair and were well maintained.

Fire Barrier Penetration Seals Ins ection Sco e (64704)

In company with the FPE, the inspector observed the condition of penetration seals in the accessible portions of south wall of the Intermediate Building. The inspector reviewed the seal installation specification and the penetration seal qualification

'eport to determine the qualified seal configurations and the allowable limits for degradations of the seals.

In addition, the data sheets in engineering work request EWR-4941 showing the specific configuration of the individual seals were reviewed to determine what should be installed in the penetrations.

Observations and Findin s Due to the Intermediate Building south wall being in the radiologically controlled area, some of the penetration seals were not accessible since they were located inside contamination areas.

Those seals that were accessible were inspected to determine if their configuration matched the design, and if there was any degradation.

For the seals inspected (which are listed at the end of the seal qualification report), the installed configuration matched the design, and no degradation was observed.

In addition, the general condition of the fire rated assembly (wall) was noted to be good, with no unsealed breaches, and all the fire.

doors functioned correctly.

Conclusions Based on the specific inspections documented in this section, and general observations made during facilitytours, the inspector concluded that the fire barrier penetration seal program was functioning well and the seals were in good conditio F2.3 Fire Hose H drostatic Tests a 0 Ins ection Sco e (64704)

During tours of the facility, the inspector noted the condition of the hoses installed on the hose reels, and looked for hydrostatic test dates on the hoses.

The inspector also reviewed the information in the database used to track the hydrostatic tests of the fire hoses.

Observations and Findin s During plant tours, the inspector observed that the hoses on the hose reels were in good condition, with fog nozzles installed.

The inspector also noted that the hoses were not marked with hydrostatic test dates, but that identifying marks were stamped into the couplings at the hose ends.

When the inspector inquired as to the dates of hydrostatic testing, the RGSE fire protection staff produced a hard copy of the database used to track the hose hydrostatic tests.

The hoses are listed by serial number, and the database shows the date of the last hydrostatic test of the hose, and on which hose reel it is installed.

The serial numbers of the hoses were stamped on the coupling at the female hose end.

The inspector was able to verify that the hoses in question had been tested within the three year interval specified for hoses installed indoors.

C.

Conclusions Based on the information in the hose database, the inspector concluded that the fire hoses were being properly tested and tracked.

F2.4 Fire Main Loo Re airs

Ins ection Sco e (64704)

The inspector reviewed ACTION Report (AR) 97-2197, "Yard Hydrant Loop Leakage," initiated December 19, 1997, and discussed the problem identified with the FPE.

The inspector did not review the associated corrective maintenance work orders, which were in review for closeout at the time of the inspection.

b, Observations and Findin s In response to a past problem with reduced residual pressure during flow testing of the yard hydrant loop of the fire mains (fed from the offsite municipal water supply),

RGSE performed a video camera inspection of the interior of the piping. Nodular corrosion of unlined carbon steel portions of the system was found to be creating blockages.

The obstructed sections of piping were replaced with cement-lined pipe, since cement-lined portions of the system had not exhibited nodular corrosion.

Removed portions were being evaluated by an offsite laborator I I

On December 19, 1997, RGSE identified leakage from the yard hydrant loop, and documented the condition in AR 97-2197. The affected portion of the system was excavated and replaced.

Video camera inspections of the in-plant loop of the fire mains, fed from the onsite fire pumps and supplying the power block and safety-related portions of the plant, showed they did not exhibit nodular corrosion.

RGKE personnel stated that they

.

believe this was due to the chlorination of the plant intake bays.

C.

Conclusions The inspectors concluded that RGSE was taking appropriate actions to maintain the fire fighting capabilities of the plant based upon the inspection of the in-plant fire mains in response to the yard hydrant loop problem, and the replacement of the diesel fire pump driver in January 1996 (see IR 50-244/95-21).

The inspector also concluded that the hydrant loop had remained functional prior to the pipe replacement.

F3 Fire Protection Procedures and Documentation F3.1 Combustible Material Control a 0 Ins ection Sco e (64704).

The inspector reviewed administrative procedure FPS-16, Rev. 2, "Bulk Storage of Combustible Materials and Transient Fire Loads," observed the storage and use of combustible materials in the plant, and closely followed activities involving the delivery and unpacking of new furniture for the control room on January 8, 1998, conducted under transient combustible permit no. 98-0002.

b.

'bservations and Findin s Procedure FPS-16 defined restrictions imposed on bulk storage of combustible materials, defined limitations on the use of combustible materials, and provided requirements for controlling transient combustible materials.

The stated intent was not to eliminate all hazards, but to control materials in a manner which willsupport plant operation while considering good fire protection and loss prevention principles.

The procedure made the various job supervisors responsible for monitoring transient combustible material, and its acceptability, throughout the plant during their tours.

Any deficiencies were to be reported to the FPE, or shift supervisor in the absence of the FPE, for resolution.

The procedure further stated that materials are considered "stored" in an area if they were left unattended.

Requirements for stored materials are more stringent than those for transient material.

Bulk storage of combustible material in the plant is prohibited except in designated areas that were approved by the FPE.

Vehicles entering the plant required permits, unless the vehicles were entering nonsafety-related areas and would not be left unattende Specific guidance for combustible materials was provided in attachments to the procedure.

So long as the use of transient materials was in accordance with the guidance provided, permits were not required.

Examples of the guidance included:

All wood brought into the protected area (except 4x4 and larger dimensional lumber) shall be treated with fire retardant material; Only non-combustible containers shall be used for storage of combustible

.

trash; HEPA and charcoal filters were not to be removed from the warehouse in.

excess of daily usage.

HEPA and charcoal filters removed from plant systems were to be removed from the area on the same day; Storage of combustible liquids within areas or structures containing safety-related equipment was prohibited; Use of open containers or buckets to collect, transport or store combustible liquids was prohibited.

Only Underwriter's Laboratory (UL), National Fire Protection Association (NFPA), or Factory Mutual (FM) listed safety cans shall be used; There shall be no storage of flammable gases within any plant structure with the exception of designated storage areas, Bulk st'orage of flammable gasses within the protected area shall be located outdoors, a minimum of fiftyfeet away from important buildings, structures, and equipment; and, Flammable and oxidizing gas cylinders shall be stored at least 25 feet apart with no combustibles between unless separated by a one-half hour rated fire wall extending a minimum of five feet high.

A transient combustible permit (TCP) was to be used for any transient combustible material brought into the protected area in a manner that conflicted with any of the procedure's guidance.

The FPE would review the proposed usage and determine if it is acceptable.

Supplemental fire protection equipment may be specified, and fire watches may be established to compensate for the increased fire loading.

The originating organization and the fire protection staff were responsible for ensuring that any backup fire protection, fire watch, or special restriction listed on the permit were maintained.

A TCP could be cancelled at any time by the FPE or shift supervisor.

Examples of conditions warranting cancellation included:

~

Loss of existing fire protection in the area;

~

Excessive fire loading in the area due to other essential work activities; or

~

Change in MODE of operation.

Upon cancellation, all combustible materials which were not in compliance with the procedure's restrictions were to be removed.

New furniture for the control room was scheduled for delivery on January 7, 1998.

In all, six pallets of boxed furniture were expected.

The engineer in charge of the modification planned to unbox, assemble and install the new furniture during the day on January 8. The FPE specified that the pallets be stored in an area of the turbine building with detection and suppression systems during the evening of the

seventh.

Two additional water extinguishers were also required by the TCP.

Seven pallets actually were delivered the morning of January 8. The FPE specified that the boxes be opened and kept in the area covered by the control room wall water curtain deluge system.

In addition, a fire watch was stationed.

The boxes and pallets were monitored by the fire watch for the entire time they were on the turbine operating floor. The empty boxes and pallets were removed from the plant by the end of the work day.

During tours of the plant, the inspector noted that no area showed any evidence of accumulation of combustible materials.

Conclusions Based on the procedural controls, and observations of activities in the plant, the inspector determined that good controls over combustible materials had been developed and implemented.

Hotwork Controls Ins ection Sco e (64704)

The inspector reviewed administrative procedure A-905, Rev. 22, "Open Flame, Welding and Grinding Permit (Hot Work Permit)," and observed hot work activities during tours of the facility.

Observations and Findin s The procedure required a permit for any work requiring grinding, heating, welding, or cutting with welding equipment, or use of open flame sources (including combustion space heaters) within the protected area.

A special caution was included regarding the effect of welding vapors on charcoal adsorbers and weld spatter on high efficiency particulate air (HEPA) filters. Exceptions were made for the permanent shop areas, the training rooms in the warehouse, and the use of combustion type space heaters used with appropriate barriers in the west end of the screenhouse.

The permit would specify the work location, the type of work, and the duration for which the permit was issued.

Provisions were made for extending the time or scope of work as well as the identity of the personnel involved. The permit specified any additional fire protection equipment to be provided in the work area, and any special precautions to be taken.

Prior to commencement of the work, the work area must be inspected by fire protection personnel to ensure that all precautions have been properly implemented.

The procedure required a fire watch during the conduct of the work activities, and for at least thirty minutes after. the work has been complete Prior to issuing the permit, the FPE was required to assess whether any fire detection equipment may be adversely affected.

Detection only equipment was to be adequately protected, and automatic detection/suppression systems were to be disconnected and compensatory measures instituted as necessary.

During a tour of the plant on January 6, 1998, the inspector observed a hot work (welding) area in the control room ventilation equipment room. The fire watch was posted, provided with an appropriate portable extinguisher, and attentive to her duties.

The area had been cleared of combustible materials.

Conclusions Based on the procedural requirements for hot work permits, and observation of a hot work job site in the plant, the inspector determined that an appropriate control program had been developed and implemented.

Station Staff Res onsibilities for the Fire Protection Pro ram Ins ection Sco e (64704)

The inspector reviewed Revisions 12 and 13 to station administrative procedure A-202, "The Fire Protection Program and Ginna Station Staff Responsibilities for Fire Protection."

In addition, the inspector discussed with the FPE the background and events which led to the most recent revision.

Observations and Findin s Station Administrative Procedure A-202 described the program for fire protection at R.E. Ginna Nuclear Power Plant, and outlined the Ginna Station staff responsibilities to assure that activities involving potential fire hazards were properly controlled.

Revision 13 to A-202 was issued as part of the corrective actions for a series of instances of fire doors being blocked open or obstructed, which occurred in December 1997 (see section F8.1), and a prior event involving temporary leak sealant.

Revision 13 had an effective date of January 8, 1998.

Revision 13 contained an amplification to section 3.3.6, which specifically listed the fire protection equipment at the station, and maintained the requirement to keep it clear of obstructions.

In addition, instructions were provided for those cases in which fire doors must be held open, or if equipment must be placed in the areas below roll-up fire doors.

An additional change placed a requirement on the Ginna Station Fire Protection System Engineer to evaluate all temporary leak seal repair activities. This requirement was added due to a fire which occurred at the site when vapors driven off a leak seal compound during curing were inadvertently ignite c.

Conclusions The inspector determined that the A-202 procedure provided appropriate guidance for the conduct of fire protection activities at the station.

F7 Quality Assurance in Fire Protection Activities F7.1 Qualit Assurance Audits of Fire Protection Pro ram a.

Iris ection Sco e (64704)

The inspector reviewed the following reports of audits of the fire protection program conducted by the quality assurance (QA) organization:

~

AINT-1996-0008-HMG,Audit of Fire Protection and Safety

~

94-22:CJK, Biennial Audit of Ginna Station Fire Protection Program

~

93-22:CJK,.Triennial Audit of Ginna Station Fire Protection Program The inspector did not review annual loss-prevention audits of the program conducted by American Nuclear Insurers (ANI).

b.

Observations and Findin s The audits conducted were combined audits of the fire protection, personnel safety, and hazardous materials programs.

Each audit included an outside auditor from another utility. Discrepancies identified by the audits were identified in Quality Assurance Occurrence Reports (QAORs), Audit Finding Corrective Action Reports (AFCARs), or ACTION Reports.

The audits found that the program had been effective, with only minor discrepancies for the most part.

The 1996 audit identified that the Fire Hazards Analysis (FHA) had not been updated.

This same issue was identified during a recent audit of the design process conducted by Duke Engineering.

The inspector discussed this matter with the FPE.

The combustible loading tables were updated in 1985 in a database used in licensing.

The database had been maintained up-to-date by licensing, but had never been formally issued as a revision to the FHA. A full review and update of the FHA was currently in progress, with completion targeted for March 1998.

C.

Conclusions Based upon the reviews of the audits, and discussions with the FPE, the inspector determined that the audits had been effective in identifying deficiencies, and in causing them to be resolve F8 Miscellaneous Fire Protection Activities F8.1 ACTION Re orts on Fire Protection Discre ancies ar Ins ection Sco e (64704)

The inspector reviewed ACTION Reports issued during December 1997 relating to fire protection, and discussed fire protection concerns with the FPE, The inspector also reviewed a memo to all station personnel from the plant manager dated December 19, 1997, a proposed supplement to the General Employee Training (GET) on fire protection, and A-202, Rev. 13, "The Fire Protection Program and Ginna Station Staff Responsibilities for Fire Protection," effective date January 8, 1998.

b.

Observations and Findin s During December 1997, eleven ACTION Reports were issued to document fire protection deficiencies.

Immediate actions were taken on each occasion to restore the condition or establish the appropriate compensatory measures.

RGSE's review of the'events found that the personnel involved were both offsite RGRE maintenance personnel and Ginna Station personnel.

Long-term corrective actions included issuing an addendum to the GET, to reinforce employee responsibilities regarding fire protection program, including control of combustible material.

C.

Conclusions The inspector determined, based on discussions with the FPE and reviewing documents described, that the plant staff were promptly identifying and correcting fire protection deficiencies.

The long-term corrective actions could not be evaluated for their effectiveness due to the short time they had been in place.

V. Mana ement Meetin s X1 Exit Meeting Summary The results of the fire protection inspection were presented on January 9, 1998, and the corrective action inspection results were presented on January 23, 1998.

After the inspection was concluded, the inspectors presented the overall results of the whole inspection period to members of licensee management on March 3, 1998.

The licensee acknowledged the findings presented.

The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary.

No proprietary information was identifie X3 Management Meeting Summary X3.1 De ut Division Director Visits On February 5 and 6, 1997, Richard V. Crlenjak, Deputy Director, Division of Reactor Projects, Region I, conducted a tour of Ginna Station.

He was accompanied by Lawrence T. Doerflein, Chief, Reactor Projects Branch 1, Region I. Additionally,,

on February 18 and 19, 1998, Larry E..Nicholson, Deputy Director, Division of Reactor Safety, Region I, conducted a tour of Ginna Station.

The individuals observed systems and equipment at the Ginna facility and met with RGKE management and staff.

L2 Review of UFSAR Commitments

't While performing the inspections discussed in this report, the inspector reviewed the applicable portions of the UFSAR that related to the areas inspected.

The inspector verified that the UFSAR wording was consistent with the observed plant practices, procedure and/or parameters, with the exception of UFSAR section 7.2.2.4.5 which incorrectly stated that the setpoint for defeat of the P-6 permissive was 1 X 10'~ on the intermediate range nuclear instruments.

The setpoint is actually 5 X 10" amps (see section 08.1).

ATTACHMENTI PARTIALLIST OF PERSONS CONTACTED Licensee B. Flynn C. Forkell G. Graus A. Harhay J. Hotchkiss G. Joss R. Marchionda P. Polfleit R. Ploof J

~ Smith J. Widay T. White G. Wrobel Primary Systems Engineering Manager Electrical Systems Engineering Manager IRC/Electrical Maintenance Manager Chemistry 5 Radiological Protection Manager Mechanical Maintenance Manager Results and Test Supervisor Production Superintendent Emergency Preparedness Manager Secondary Systems Engineering Manager Maintenance Superintendent Plant Manager Operations Manager Nuclear Safety 5 Licensing Manager INSPECTION PROCEDURES USED IP 37551:

IP 40500:

IP 61726:

IP 62707:

IP 64704:

IP 71707:

IP 83750 IP 92700:

IP 92901:

IP 92904.

Onsite Engineering Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems Surveillance Observation Maintenance Observation Fire Protection Program Plant Operations Occupational Radiation Exposure Onsite Follow-up of Written Reports of Nonroutine Events at Power Reactor Facilities Follow-up - Operations Follow-up - Plant Support ITEMS OPENED, CLOSED, AND DISCUSSED

~Oened IFI 98-01-01 NCV 98-01-02 Closed NCV 98-01-02 VIO 97-01-02 Tracking the closure of corrective actions Inadequate Reactor Trip Instrumentation Procedure Inadequate Reactor Trip Instrumentation Procedure Radiological Boundary Controls

Attachment I

URI 97-02-01 LER 96-003, Rv

Intermediate Range Nuclear Instrument Operability Both Pressurizer Relief Valves Inoperable, Results in Condition That Could Have Prevented Fulfillment of a Safety Function LER 97-006, Rv 1 Verification of Boron Concentration Not Performed Due to Misinterpretation of Event Sequence, Resulted in Condition Prohibited by Technical Specifications LER 97-007, Rv 1 Reactor Trip Instrumentation Would Have Been in a Condition Prohibited by Technical Specifications Discussed IFI 97-10-01 Weak Configuration Control AFCAR AFW ALARA ANI AR CCW CFR EDG EOOS ESF EWR FHA FM FPE GET IFI IR ITS LCO LER MOPAR MOV NFPA Nl NRC NRR NSARB PCN PCR Audit Finding Corrective Action Report Auxiliary Feedwater As Low As Reasonably Achievable American Nuclear Insurers ACTION Report Component Cooling Water Code of Federal Regulations Emergency Diesel Generator Equipment Out-of-Service Engineered Safety Feature Engineering Work Request Fire hazards Analysis Factory Mutual Fire Protection Engineer General Employee Training Inspector Follow-up Item

Inspection Report

Improved Technical Specification

Limiting Condition for Operation

Licensee Event Report

Morning Priorities Action Required

Motor-Operated Valve

National Fire Protection Association

Nuclear Instrument

Nuclear Regulatory Commission

Nuclear Reactor Regulation

Nuclear Safety Audit and Review Board

Procedure Change Notice

Procedure Change Request

Attachment

I

PORC

PORV

ppm

PT

QA

QAOR

QC

RCA

RGS.E

RHR

RPSC

SFP

SI

TCP

TSC

UFSAR

UL

URI

VIO

Plant Operations Review Committee

Power-Operated

Relief Valve

parts per million

Periodic Test

Quality Assurance

Quality Assurance Occurrence Reports

Quality Control

Radiologically Controlled Area

Rochester Gas and Electric Corporation

Residual Heat Removal

Radiological Protection and Chemistry

Spent Fuel Pool

Safety Injection

Transient Combustibles Permit

Technical Support Center

Updated Final Safety Analysis Report

Underwriter's Laboratory

Unresolved Item

Violation