IR 05000333/1992027

Insp Rept 50-333/92-27 on 921225-930123.Violations Noted But Not Cited.Major Areas Inspected:Plant Operations, Radiological Controls,Maint,Surveillance,Engineering & Technical Support & Qa/Safety Verification

ML20034F608
Person / Time
Site:	FitzPatrick
Issue date:	02/19/1993
From:	Eselgroth P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20034F604	List: IR 05000333/1992027 ML20034F603
References
50-333-92-27, NUDOCS 9303040044
Download: ML20034F608 (23)

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U.S. NUCLEAR REGULATORY COMMISSION Region I

Report No.:

92-27

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Docket No.:

50-333 l

License No.:

DPR-59 Licensee:

New York Power Authority P.O. Box 41

Lycoming, New York 13093

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Facility:

James A. FitzPatrick Nuclear Power Plant r

Location:

Scriba, New York Dates:

December 25,1992 through January 23,1993

.1 Inspectors:

W. Cook, Senior Resident Inspector

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J. Tappert, Resident Inspector

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P. Harris, Resident Inspector, Vermont Yankee L. Scholl, Reactor Engineer.

B. McCabe, Project Manager j

R. Urban, Project Engineer

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R. Lorson, Reactor Engineer

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Approved by:

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Peter W. Eselgif6th, Chief Date.

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eactor Projects Section IB, DRP 1'

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INSPECTION SUMMARY: Routine NRC resident inspection of plant operations, j

' radiological controls, maintenance, surveillance, engineering and technical support, and '

j quality assurance / safety verification.

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RESULTS: See Executive Summary

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E3_ECUTIVE SUMMARY James A. FitzPatrick Nuclear Power Plant NRC Region I Inspection Report No. 50-333/92-27 12/25/92 - 01/23/93 Plant Onerntions At the beginning of the inspection period, the plant was in cold shutdown making final preparations to startup. The startup commenced on January 2 and the reactor was taken critical at 1:52 a.m. on January 3,1993. NYPA implemented their startup and power ascension plan throughout the remainder of the period. The inspectors concluded that the startup was conducted in a slow, deliberate and safe manner. Tours of the plant and primary

containment identified no significant deficiencies. Inspector followup satisfactorily closed two unresolved items regarding the battery room ventilation.

Rndiological Controls The changing radiological environment was emphasized to plant workers and highlighted at shift turnover and management meetings. The inspector noted that no specific guidance was given to the radiation protection (RP) staff concerning suspected radon gas contaminations.

RP management subsequently provided written guidance to the staff. Inspector review'of the guidance found it acceptable.

Mnintennnce The inspectors noted acceptable performance in the area of maintenance. Some minor deficiencies with the diesel fire pump maintenance procedure were identified and corrected.

NYPA successfully repaired numerous leaks in the main condenser. Good initiative was i

used in identifying and correcting a longstanding deficiency with the RCIC turbine overspeed

trip mechanism. An unresolved item concerning the protective tagging program was closed.

Sairveillance The inspectors noted acceptable performance in the area of surveillance testing. NYPA did

not meet a procedural requirement and NRC commitment to cycle the safety relief valves -

within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of reaching 940 psig. At the conclusion of the inspection period, NYPA was still formulating their corrective actions. Pending NRC review of these actions, this is an unresolved item (URI 92-27-01). The inspector noted an error in an "information only" plant computer display concerning average jet pump loop differential pressures. After the error was corrected, the inspector determined that the change to the computer software was made without procedurally required management reviews and approvals. NYPA took prompt i

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corrective action in response to this observation and based upon the minor safety significance, this was a non-cited violation. The inspectors noted that the H/0 containment

monitors were not operable during initial unit startup. The licensee was informed of this observation and entered the limiting condition for operation and initiated appropriate corrective actions. Although a reactor mode change was made with these monitors

inoperable, contrary to TS 3.0.D, this violation was not cited due to the minor safety significance and prompt corrective action by NYPA.

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Dngineerine and Technical Support The inspectors conducted a survey to determine FitzPatrick's vulnerability to a loss of ECCS pumps due to insulation clogging of suction strainers. The inspectors concluded that FitzPatrick is not susceptible to this failure mode principally due to the SRV relief to the torus vice directly to the drywell.

Safetv Assessment /Oualitv Verification a

The inspectors noted good communications between site and corporate engineers at the monthly project and engineering meetings. Management oversight during the startup and

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power ascension was comprehensive. NYPA's self assessment plan during startup was well..

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executed.

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s TA.BLE OF SUMMARY OF FACILITY ACTIVITIES

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1.1 NYPA Activities

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At the beginning of the inspection period, the plant was in cold shutdown making final preparations to commence startup. With identified exceptions, NYPA documented their l

readiness to restart FitzPatrick by letter dated December 17, 1992. The exceptions were

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addressed in supplemental submittals dated December 22, 23, and 24,1992. On December -

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29, 1992, the NRC concluded that FitzPatrick was capable of safe plant operation and l

concurred with NYPA's assessment that the plant was ready for restart.

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On January 2,1993, the reactor mode switch was placed in "Startup/ Hot Standby" and the reactor was taken critical at 0152 on January 3. NYPA implemented their startup and power ascension plan throughout the remainder of the assessment period. At the end of the period, the plant was at 20% and preparations were being made to place the generator on line.

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l.2 NRC Activities The inspection activities during this report period included inspection during normal,z

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backshift and weekend hours by the resident staff. There were 132 hours0.00153 days <br />0.0367 hours <br />2.18254e-4 weeks <br />5.0226e-5 months <br /> of backshift -

(evening shift) and 104 hours0.0012 days <br />0.0289 hours <br />1.719577e-4 weeks <br />3.9572e-5 months <br /> of deep backshift (weekend, holiday and midnight shift)

inspections during this period.

A team of region based, headquarters, and resident inspectors provided 24-hour coverage from December 28,1992 through January 7,1993, to monitor NYPA's restart (backshift coverage was relaxed on New Years Day). After 24-hour coverage was discontinued, region based inspectors continued to provide augmented inspection coverage throughout the remainder of the assessment period. Specific observations of the team are detailed later in the report.

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2.0 PLANT OPERATIONS (71707,71710,93702)

2.1 Routine Plant Oncrations Review

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During the inspection period the inspectors observed control room activities including

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operator shift turnovers, shift crew briefings, panel manipulations and alarm response, and routine safety _ system and auxiliary system operations conducted in accordance with approved operating procedures and administrative guidelines. The inspectors made independent verification of safety system operability by review of operator logs, system markups, control y

panel walkdowns and component status verifications in the field. Discussions were held with j

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i operators and technicians in the field to assess their familiarity with current system status and personnel response to events during the inspection period. In addition, during plant tours,

inspectors reviewed routine radiological control practices. The activities inspected were acceptable.

2.1.1 Operational SaStv Verification The inspector conducted partial control room and in-plant walkdowns of the following systems:

Emergency diesel generators

Reactor feed pumps

Reactor core isolation cooling

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High pressure coolant injection

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Main condensers Battery room ventilation systems

The inspector noted no deficiencies and the above listed systems were properly aligned for

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operaticn or in a standby configuration.

2.2 Plant Tours l

On December 28,1992, the inspector toured the drywell to verify the adequacy of the-

material condition prior to resumption of power operations. The inspector was accompanied by the General Manager of Operations and a licensed senior reactor operator. No significant i

housekeeping or material deficiencies were noted. Minor items (i.e., flashlights, temporary -

j hoses, drop lights, electrical extension cords, and the like) were identified during the tour

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- and were itemized by the NYPA representatives for removal prior to final drywell closeout.

On January 8,1993, the inspector toured accessible portions of the top of the torus. At the'

i time of the tour, the radiation protection staff was overseeing the decontamination _of this

area. In general, the material condition of this area was good. Approximately 50% of the a

decontamination effort was completed and results achieved were satisfactory. The inspector identified no deficiencies.

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2.3 Event Followup 2.3.1 Engineered Safety Feature Actuation On Decembe-r 28,1992, NYPA notified the NRC per 10 CFR 50.72 that an inadvertent engineered safety feature (ESP) actuation occurred. Specifically, while conducting turbine electro-hydraulic control (EHC) testing an unanticipated primary containment isolation valve Group 1 isolation signal occurred. The Group 1 (main steam isolation valves-8, mam steam drain isolation valves-2, and reactor water sample isolation valves-2) isolation signal teru1ted when the turbine reset push button was depressed. The turbine stop valves went open (should not have) and with no condenser vacuuri this satisfied the Group 1 isolation logic.

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After several days of troubleshooting, the EHC circuit problem which caused the turbine stop valves to open was traced to a couple'of suspect circuit boards. These circuit boards were

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replaced and the EHC system was successfully retested. Bench testing was planned to

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identify the specific component failure on the circuit boards.

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The inspector noted good pre-planning and troubleshooting to identify the EHC system i

problems. Troubleshooting was carefully controlled through the Work Control Center and i

by the control room operators to minimize interference with unit startup activities.

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<t 2.3.2 Breaker Coordinator Concern Identified by NYPA

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On January 20, NYPA notified the NRC per 10 CFR 50.72 that the plant was found to be m a condition that is :outside the design basis. Specifically, during an engineering re-evaluation j

of electrical bus 12600 time current characteristic curves for feeds to motor control centers

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(MCCs) 262 and 264, it was determined that the breaker to the main turbine turning gear oil'

pump (71 MCC 262/OA2) did not coordinate with the supply breaker (12608) for all potential fault currents. For example, a fire induced fault could potentially result in a race to trip between breaker 12608 and 71 MCC 262/OA2. This lack of coordination adversely -

affected 10 CFR 50, Appendix R (safe shutdown) compliance in the cable' spreading room and the battery room corridor.

The inspector determined that the engineering re-evaluation was performed by a contractor after a breaker coordination package review by a NYPA engineer determined that the

architect / engineer study did not necessarily use the worst case circuit breaker. The i

contractor's review identified seven potential circuit breakers with coordination concerni,

however, only the turbine turning gear oil pump breaker compromised the Appendix R

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analysis. The inspector verified the adequacy ofinterim corrective actions until a smaller

amperage circuit breaker could be installed. The interim action consisted of a revision to i

AOP-43, Plant Shutdown fiom Outside the Control Room, which specifies to open 71 MCC

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262/OA2 and verify closed or reset breaker 12608, if it has tripped. This action does not -

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compromise the time line verification for AOP-43. At the conclusion of the inspection period, the breaker was replaced with a smaller amperage circuit breaker. The inspector

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concluded NYPA's response to this identified problem was timely and appropriate.

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2.4 Control Room Observations

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2.4.1 Annroach to Criticality On January 3, the inspectors observed that the conduct of plant operation during the approach to criticality was very good. This was accomplished, in part, because NYPA management directed that the shift turnover be conducted prior to reactor startup allowing an uninterrupted ascension to power. This conservative management decision reflected an excellent safety perspective and contributed significantly to safe plant. operation.

A good shift brief was held which involved operators, as well as, management and technicians.

Overall, there was very good management oversight. The Vice President, Nuclear Support, observed the approach to criticality in the control room. In addition, managers experienced l

in reactor operations were assigned to each shift as a Senior Nuclear Manager (SNM). The SNM was responsible, in part, _to, rovide operational experience, timely management direction and decision making, and prioritization of startup acdvities. This initiative was considered effective.

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The shift supervisor and nuclear control operator displayed effective command and control by managing minor maintenance, startup surveillance, and system lineup changes so as to not.

i distract the reactor operator during the approach to criticality. Very descriptive and accurate communications contributed to the proper performance of in-plant operations and l

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maintenance. Communications between the reactor engineer (RE) and the reactor operator, responsible for the control rod manipulations, was also good. The inspector noted, however,.

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that the level of participation by the RE gave the appearance that the rod sequence was controlled by the RE vice the licensed reactor operator. This observation was recognized by

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the Vice President, Nuclear Support, who promptly and effectively re-emphasized that the l

licensed operators were responsible for reactor operation.

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Overall, the approach to criticality was slow, deliberate, and safe.

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2.4.2 Torus Water Level

On January 4, the torus low water level annunciator alarmed (13.88 feet) due to control room

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operators having started the B residual heat removal (RHR) pump in the torus cooling mode -

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of operation in preparation for high pressure coolant injection system (HPCI) operability

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testing. The shift supervisor entered emergency operating procedure (EOP)-4, Primary Containment Control, and restored level. The total time the alarm condition was enabled was approximately 35 minutes.

The inspector determined that Technical Specification (TS) 3.7.A.1.a. and b. requires torus

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level between 53 and 51.5 inches, inclusive, which corresponds to an indicated level of 14.0 and 13.88 feet, respectively. This TS LCO does not have an action statement, therefore, TS-

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3.0.C applies. TS 3.7.A.1.b does allow torus level to be outside the level limits for four hours during required operability testing of HPCI, RCIC, RHR, CS and the drywell-torus

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vacuum system. Therefore, it was appropriate that the TS 3.0.C was not entered.

ihe FitzPatdek TS recognize that momentary torus water levels fluctuations outside of TS lindts may happen. Standard TS also recognizes this, but specifically allows this to happen for up to one hour, even when there is no operability testing in progress. The SS promptly enterxl EOP-4, logged this event, and restored level. The inspector concluded these actior.s were appropriate.

Followup by the inspector identified that surveillance procedure ST-40D, page 52 of 159,~

lists the allowable range of the torus water level instruments as 13.8-14.1 feet for 'all

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insuuments. However, only five of twelve of the instrumenis are of a scale accurate enough to determine if the torus water level is within the allowab!c band. Operations management

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took prompt action to revise and clarify ST-40D to ensure proper logging and once per shift verification of torus water level. The inspector reviewed this revision and found it satisfactory.

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2.4.3 Drywell Eauipment and Floor Drains Review In reviewing NYPA's compliance with reactor coolant pressure boundary leakage

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monitoring, the inspectors observed drywell equipment and floor drain sump surveillance i

methodohgy and results, as controlled by surveillance test ST-40D. The inspectors identified no deficiencies in the methodology or daily log keeping. However, in researching

the sources of input to the drywell equipment drain sump, the inspectors identified that the.

Updated Final Safety Analysis Report (UFSAR), Figure 11-1-2, Revision 0, states that the drywell coolers drain to the equipment drain sump. Further discussion with control room

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operata s and review of control room drawings did not support this UFSAR information.- To

resolve this documentation conflict, the drywell cooler drains were walked down during the

150 psig inspection of the drywell. Operators confirmed that the drywell coolers drain to the e

drywell floor drain sump. NYPA has initiated a revision to UFSAR Figure 11-1-2 to correct this documentation error. The inspector had no additional concerns.

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2.5 Previously Identi6ed Items 2.5.1 (Closed) Unresolved Item (91-01-03): Adequacy of Battery Room Ventilation System Testing

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This unresolved item addressed a concern that the weekly surveillance test for 'he battery room ventilation system (ST-19) was inadequate to demonstrate proper performance of the air handling units to maintain battery room temperature between 60 and 110 F. Inspector followup of this concern identified that specific periodic performance testing of the battery -

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room ventilation system is not required or warranted. The basis for this conclusion includes recognition of the following:

The weekly performance of ST-19 verifies that the ventilation systems for both battery

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rooms function as designed. Upon a loss of the running exhaust fan, the standby

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exhaust fan starts. Their associated dampers close and open, respectively, to continually exhaust air from the battery rooms to the outside atmosphere. ST-19 also verifies the proper flow of air through the air handling units and associated recirculation fans. Consequently, ST-19 ensures that there is sufficient air turnover to prevent an explosive hydrogen buildup due to battery discharging or charging.

Battery room temperature monitoring is performed by the auxiliary operators (per

ODSO-17, Auxiliary Operator Plant Tour and Operating legs) to verify battery room

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temperature is within allowable limits every eight hours. This ensures proper battery performance within the allowable temperature band.

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NYPA's Technical Specification Interpretation No. 6, dated August 8,1988, appropriately clarifies the operability requirements for the battery room ventilation l

system to ensure proper and consistent operator response, in the event a component or sub-system of the battery room ventilation system is out of service.

Based on the above, this unresolved item is closed.

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2.5.2 (Closed) Unresolved Item (91-01-04): Inadequate Control of Temocrary Modifications This unresolved item was assigned to followup on NYPA improvements in the area of temporary modification controls and on the control of battery room ventilation system fire damper inspection covers. With respect to control of temporary modifications, this area of concern was reviewed in detail in two earlier inspection repons (reference 92-82 and 92-23)

and substantial improvement was noted by the inspectors. Regarding the control of fire damper inspection covers, the inspector conducted a walkdown of the battery room

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t ventilation system and verified that all inspection covers were properly secured. The inspector will continue to monitor this system and other ventilation' systems during routine plant walk-throughs to ensure proper integrity of these systems is maintained. This unresolved item is closed.

3.0 RADIOLOGICAL CONTROLS (IP 71707)

I The inspector observed routine radiological work practices during observation of various maintenance activities and in routine tours of the plant. In general, radiological workers seemed to be well-trained and were observed to be using appropriate radiological work.

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practices (i.e., bagged tools and other items, as required, maintained work areas clean, removed protective clothing properly, dosimetry worn properly, and all radiological postings

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obeyed). The health physics technicians were observed to give good pre-job briefings and

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maintained close surveillance over the work activities in their assigned areas. The '

l radiological work areas, in general, were well-maintained (i.e., clean r.ifn app.mpriate radiological postings). The inspector concluded that the workers and health phpics technicians were working well together to ensure safe and appropriate rdiologbl work practices.

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The changing radiological environment was emphasized to workers. A memo was issued

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describing expected radiation levels in the plant. Work that was subject to changing radiological, as well as industrial safety conditions, was highlighted at shift turnover and management meetings.

i 3.1 Noble Gas and Radon Personnel Contaminatiotl On January 7, upon exiting from the restricted area, the inspector alarmed the portal monitor which performs a whole body frisk to detect personnel contamination. A hand frisk of the

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suspected contaminated area (pant legs) with a DT-304 probe by the attending radiation protection (RP) technician failed to identify any fixed or high energy contamination.

Suspecting radon gas contamination, the RP technician attempted to disperse the slightly charged radon gas with a dry muslin cloth and tape. The technique was repeated until after a

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fourth portal monitor frisk, the inspector was determined free of contamination and permitted

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to exit the restricted area.

Inspector followup of this event identified that the site guidance for personnel contamination (Radiation Protection Procedure, RPP-2, para. 5.5) specifically excludes situations where.

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thraugh radioactive decay in a brief time period. Consequently, no written guidance was provided to RP tcclmicians for situations encountered on January 7. The inspector

considered the RP technician's actions to have been appropriate, however, based upon other observed radon gas personnel contaminations and different RP technician responses, a i

uniform approach to resolving these types of contamination events appeared warranted to j

ensure proper identification and handling of the contamination.

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After discussions with RP management, written guiddace war provided to the site RP staff '

defining the necessary actions to be taken in the everd e noble gas or radon personnel r

contamination. The inspector reviewed this guidance with the Radiological and Environmental Services manager and found it to be acceptable.

4.0 MAINTENANCE (IP 62703)

4.1 Observation of Maintenance Activities The inspector observed and reviewed selected portions of preventive and corrective maintenance to verify compliance with codes, standards and Technical Specifications, proper use of administrative and maintenance procedures, proper QA/QC involvement, and appropriate equipment alignment and retest. The following activities were observed:

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4.1.1 Review of Diesel Fire Pumo Maintenance The inspector reviewed the activities in support of and involving planned preventive maintenance on the diesel fire pump, performed in accordance with maintenance procedure

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MP-76.1. The inspector also witnessed the post-maintenance testing conducted per surveillance test MST-76.01. The following observations were noted:

Technical Specification (TS) 3.12.A limiting condition of operation for diesel fire pump inoperability was entered at 1:40 a.m., on January 5 per the shift supervisor's log. The inspector determined that the basis for entry into the limiting condition of-operation (LCO), at that time, was the receipt of the diesel fire pump trouble alarm

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caused by the tagging of the local control switch to "off". However, the inspector.

noted that a similar tag was hung earlier on the control room switch. After discussion -

with the shift supervisor and operations department management, the inspector.

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concluded that the guidance provided to shift supervisors, per Plant Standing Order

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62, was not explicit with respect to when the TS LCO time clock should be started when tagging a system out-of-service. NYPA resolved to declare the system TS inoperable upon issuance of the protective tags for hanging, and start the LCO time

clock at that time. Written guidance was issued to the operations staff to document management's expectations. - The inspector found this approach satisfactory.

Detailed review of MP-76.1, by the inspector, identified that the procedure lifts 'the i

diesel fire pump battery leads to provide personnel protection, but does not explicitly restore the leads following the maintenance. This observation was conveyed to the maintenance department management who conducted a detailed review of MP-76.1 and corrected this and numerous other minor procedural deficiencies. The inspector reviewed the revised procedure and found it to be much improved.

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The inspector reviewed surveillance procedure MST 76.01 and the preventive -

maintenance package (PM 503980), for this LCO maintenance. The documentation was complete, properly reviewed and approved, and accurately reflected the safety-l class and function of the diesel fire pump. The most recent revision of the surveillance procedure was in use in the field and personnel were knowledgeable of precautions and requirements.

Overall, the conduct of this maintenance activity was good. Coordination and planning of the diesel fire pump outage was appropriately done. Personnel involved proceeded carefully

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i and cautiously through the various evolutions. Housekeeping in the diesel fire pump room was commensurate with the maintenance performed.

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4.1.2 Main Condenser Leaks On January 3, '993, the reactor was taken critical. Power was raised into the power range l

and temperature and pressure were raised. As the reactor began steaming, a high conductivity was noted in the condensate and both mechanical vacuum pumps were needed to maintain vacuum. This indicated both circulating water and air in-leakage into the condensers. The water boxes were isolated sequentially. The B2 water box was identified as the most significant contribution to leakage since a large drop in conductivity and vacuum

occurred when it was isolated and drained. Concurrent with this effort, inspections were i

being made in the plant to identify any air in-leakage. Several minor air leaks were

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i identified and corrected.

l Several tube leaks were identified in the B2 waterbox. The inspector observed various aspects of the condenser tube inspection and repair and found the job pre-brief to be q

thorough and good communications and coordination were used throughout the process.

NYPA identified several discrepancies with the condenser tube map. Many tubes that were indicated as plugged were not and many tubes that were indicated as unplugged were

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plugged. NYPA plugged all tubes that were indicated as plugged and noted all other l

discrepancies. The three other waterboxes were entered sequentially and discrepancies with their respective tube maps were resolved. To resolve any future problems with tube map

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discrepancies the maps will be kept by the technical scavices department on site to ensure j

positive control. The inspector identified no concerns with NYPA's activities to address the I

condensor tube leakage or tube map concerns.

4.1.3 Recirculation Pump Seals On January 4,1992, during initial plant pressurization to 150 psig, it was noted that the B recirculation pump inner seal pressure was approximately equal to the outer seal pressure.

This was indicative of a failure of the inner seal to seat properly. This indication was not

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conclusive at low pressure, since the seal may seat as operating pressure increases. The licensee made preparations to replace the seal by ensuring all required parts were'on. site and

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paperwork to support the replacement'was in order. As reactor pressure was increased, it became evident that the seal would not reseat.

The inspector determined that both inner and outer seals were replaced during the outage.

The inner seal failed to seat during the reactor hydrostatic test and was replaced again.

There is no technical specification for seal operability and identified leakage is well below the allowable limits. The licensee has reviewed industry experience with only one seal in service and has determined that the unit can continue to be safely operated withouti jeopardizing damage to the recirculation pump. However, NYPA plans to operate for about one month before conducting a unit shutdown to replace the seal. Additionally, NYPA is reviewing their maintenance procedure to determine if enhancements can be made to reduce

the seal failure rate. The inspector has concluded that NYPA's handling of this problem during unit startup has been prudent and appropriate.

4.1.4 Egaetor Core Isolation Cooling (RCIC) Oversneed Test

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From January 8 to January 10, 1993, NYPA performed ST-24G, RCIC Turbine Overspeed '

Test. After three successive failures where adjustments to the overspeed mechanism were unsuccessful in providing consecutive satisfactory trips, the maintenance engineer identified a small boss on the governor cover that was interfering with the governor weight as it approached the tappet (trip mechanism). (The boss in the casting could be used to tap a bolt hole, but was not used in this configuration.) The governor weight would occasionally strike -

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the boss and cause oscillations in the governor weight. This caused the lack of repeatability.

j of the test. Subsequently, the boss was ground out and the turbine was satisfactorily'

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overspeed tested. The approach to finding and correcting the root cause of the problem was in sharp contrast to the previous outage when ST-24G was run 29 times to achieve a satisfactory result.

(Closed) Unresolved item (90-09-Olh Equioment and Personnel Protective Taeging The item 90-09-01 identified that NYPA's implementation of the Protective Tag Record-(PTR) Program was incomplete with respect to electrical and mechanical interfaces, and also that job specific guidance which involved mechanical and electrical interfaces was weak.

The inspector reviewed procedure WACP-10.1.2, revision 5 which establishes the protective

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tagging requirements that are used to enhance personnel and equipment safety during the performance of maintenance activities. The inspector concluded that the program was

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comprehensive and adequate to provide for personnel and equipment safety. The inspector reviewed the implementation of this program by examining selected PTR forms, program.

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records, and existing tags in the plant. No deficiencies were identified on any of the PTR f

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records or tags reviewed. Activities involving electrical and mechanical interfaces were observed to be properly addressed on all PTR records reviewed. Based on the' records review and field observations, the inspector concluded that NYPA's performance in this area was acceptable. This unresolved item is closed.

5.0 SURVEILLANCE (61726)

The inspector observed and reviewed portions of ongoing and completed surveillance tests to

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assess performance in accordance with approved procedures and Limiting Conditions for Operation, removal and restoration of equipment, and deficiency review and resolution. The following activities were reviewed:

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5.1 Intermediate Range Monitor (IRM) Instrument Trip Function Calibration. ISP-71 ISP-71 was observed on January 5. This IRM surveillance test was well controlled. The technicians performing the surveillance exhibited excellent system knowledge and were fully cognizant of the effects of procedural steps on IRM system operation. Communications were accurate, formal, and in accordance with management expectations. Very good plant control was exhibited by the shift supervisor when he temporarily secured the IRM surveillance so that he could fully understand the cause of a fire panel alarm.

5.2 Reactor Safety Relief Valve Testing On January 18,1993, NYPA was reviewing ST-22B, Manual Safety Relief Valve (SRV) ~

Operation and Valve monitoring System Functional Test (IST), in anticipation of performing i

the test. The procedure review revealed a test frequency requirement that the surveillance be started within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reaching a pressure of 940 psig. This time limit had expired since 940 psig had been reached at 4:30 p.m. on January 16. The procedural requirement was a specific corrective action following an event (reactor scram) in December 1989.

Neither the governing procedure, OP 65, Startup and Shutdown Procedure, nor the FitzPatrick 1992 Startup Plan identified any time requirements associated with SRV testing.

The original commitment was made in LER 89-23, dated December 12, 1989, as the result of an automatic scram during SRV testing at lower pressures with the mode switch in

"Startup/ Hot Standby". With the mode switch in "Startup/ Hot Standby", the reactor scram-high flux setpoint is 15%. As stated in LER 89-23, the pressure and power transient associated with cycling the SRV's was sufficient to reach this high flux setpoint and the reactor scram resulted. To preclude this from happening in the future, NYPA committed to complete this test within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reaching 940 psig when used to demonstrate I

operability of the SRV's. This LER 89-23 commitment was consistent with the standard Technical Specifications (TS) for boiling water reactors and would allow the reactor mode-switch to be in "Run" and thus raise the high flux scram setpoints. FitzPatrick TS 3.5.D

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requires the automatic depressurization system (ADS) to be operable whenever reactor pressure is greater than 100 psig and there is irradiated fuel in the vessel. The surveillance

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requirement (4.5.D) requires during each operating cycle that all SRVs be manually opened

while bypassing steam to the condenser to verify the SRVs open. However, 4.5.D provides no further guidance on when this must be accomplished.

In response to the missed requirement, NYPA researched the source of the commitment and continued with preparations to perform the surveillance. At the conclusion of the inspection l

period, NYPA was still formulating their corrective actions for failure to meet the procedural

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time requirements. Pending NRC review of these actions, this item will remain open (Unresolved Item URI 92-27-01). The inspector noted that NYPA's implementation of.

t commitments was weak, in this case, in that the commitment was incorrectly transcribed from the LER to the ST (completing test within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> versus starting test within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />)

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and that the commitment was not incorporated into higher tier' operating procedures.

The actual performance of this surveillance was observed on January 19, 1993. The test was conducted in a deliberate and professional manner. Excellent command and control was evident throughout the performance of the test.

5.3 Reactor Recirculation System Jet Pumo Computer Points On January 4,1993 the inspector reviewed the daily jet pump surveillance test, ST-23C and.

observed the jet pump instrumentation in the control room. The inspector noted that the B Loop jet pump differential pressures indicated less than one psid, while the A Imop jet pump differential pressures were all greater than three psid. This difference was apparently the result of a slight difference in the reactor water recirculation (RWR) pump speeds (both pumps were operating at thebr minimum speed control settings). While discussing the jet

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pump performance with the reactor operator, the jet pump differential pressure information was observed on the Emergency and Plant Information Computer (EPIC) display. The inspector observed from the jet pump display that although all of the ten Loop B jet pumps were operating at less than one psid, the computer was displaying an average jet pump differential pressure for Loop B of 2.13 psid. Also, with all of the Loop A jet pumps

indicating greater than three psid, the displayed average was 2.09 psid. Further investigation revealed that t! e inputs which were being averaged were incorrect. Five inputs from the A Loop jet pumps and 5 inputs from the B Loop jet pumps were being used to generate the A loop average differential pressure and the remaining inputs (five from each loop) were being i

utilized to generate the B IAop average differential pressure. These average values are not used in the surveillance test and therefore did not affect the results of the test.

l The control room operator agreed that the computer averages were incorrect and contacted a technician in the computer group to inform him of the problem. The computer technician l

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acknowledged the problem and then left the control room to correct the computer software.

At the end of the shift, the inspector was informed that the problem had been corrected. The t

NRC inspector on the following shift verified that the problem had been corrected and, at that time, inquired how computer software changes are administratively controlled. The inspector determined that procedure EPIC-2.2, EPIC Mimic and Algorithm /C Point, was the

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governing station procedure, however, it was not used by the technician. The technician used an unapproved form to document computer changes and the form did not ensure all the requirements of EPIC-2.2 were satisfied.

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The inspector informed station management about this observation and a critique of the event was conducted by the computer department staff. The computer staff confirmed that proceduce EPIC-2.2 was not followed. The individuals involved were counseled on the need to follow existing plant procedures. The importance of proper management review and approval of EPIC computer changes was emphasized. All computer changes made during the 1992 refueling outage were also reviewed to ensure that they comply with EPIC-2.2 and that

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all tracking and approval forms were properly used. Additionally, the EPIC-2.2 procedure will be revised to make it more detailed, but less cumbersome to follow.

This event constitutes a violation of plant Technical Specification 6.8, which requires written procedures to be implemented. However, based on the criteria specified in Section V.A of the Enforcement Policy (i.e., the violation was not willful, it was of minor safety significance, and the licensee initiated appropriate corrective action prior to the conclusion of the inspection period), this violation was not cited.

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5.4 Reactor Feed Pumo Turbine Mechanical Oversoeed Test On January 13, 1993, the inspector observed the performance of the A reactor feed pump.

turbine (RFPT) overspeed test which was conducted in accordance with procedure MP-034.11, revision 01. Operators, both in the control room and at the feed pump, were

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attentive and followed the procedure. Good control of turbine speed was maintained throughout the test.

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Minor material deficiencies were noted during the test. The A RFPT vibration readmgs registered on the B vice A remote vibration indicator. Also, the A RFPT tripped

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approximately two revolutions per minute (RPM) below the specified trip range (5445-5555 RPM). Appropriate actions were initiated by the NYPA staff to address both of these deficiencies. The vibration detector leads were found swapped and the turbine vendor accepted the 2 RPM low trip setting. The acceptance range was properly revised in the test procedure.

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On January 15, the inspector observed the performance of the B RFPT overspeed test in accordance with procedure MP-034.11, revision 01. The RFPT failed to automatically trip at the high revolution per minute setpoint and the turbine was appropriately manually tripped.

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Control of the test was good and communications were clear. Operator response to a high RFPT vibration alarm was prompt and consistent.with the alarm response procedure.

Following the aborted test, the inspector reviewed the work instruction developed to adjust

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the RFPT overspeed trip setpoint. The inspector noted that the work instructions were appropriately detailed. The inspector observed a portion of the adjustment and noted that the workers were lmowledgeable about their task, and performed the adjustment in a professional manner.

5.5 High Pressure Coolant Iniection (HPCI) Ouick-Start Transient Monitorine Test On January 21 the inspector observed a portion of the HPCI Quick-Start Transient Monitoring Test which was conducted in accordance with surveillance test procedure ST-4P, revision 5. Specifically, the inspector observed portions of the test designed to demonstrate the ability of the HPCI system to perform a " quick start" and also the portion that monitored the HPCI flow control system stability. The inspector noted that test personnel properly coordinated and controlled the test. The inspector observed that the HPCI response time was

within the acceptance criteria. The test crew identified a procedural problem in that the data'

recorder chart speed specified in the test procedure was too slow to analyze the control system response. The procedure was promptly revised to require a faster chart recorder speed. The inspector was satisfied with the test crew's identification and corrective actions taken regarding this issue.

5.6 Containment Atmosphere (HJOJ Monitors At 1:30 p.m. on January 2,1993, the mode switch was placed in "Startup/ Hot Standby" to

- commence a normal reactor startup. The reactor was taken critical on January 3 at 1:52 a.m. Shortly after criticality, the inspector noted that the O, function of the H/0 monitors

was inoperable. When questioned, operations personnel responded that the monitors were not required until containment inertion (within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of mode switch being place in

"Run"). Inspector review determined that technical specification Table 3.2-8, Accident Monitoring Instrumentation, requires one of two channels operable while in "Run" and

"Startup/ Hot Standby". Technical specification 3.7.A.9 requires that the drywell atmosphere be continuously monitored when containment integrity is required. Technical specification 3.7. A.2 requires containment integrity when the reactor is critical. These requirements were brought to NYPA's attention. The plant staff determined that the instruments were inoperable because the surveillances were not performed yet (they were scheduled to be performed after the containment was inerted) and entered the limiting condition for operation

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(LCO) action state-ment (3.7.A.9.a). The LCO action statement requires a grab sample to be taken each 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period from and after the time primary containment atmosphere monitoring instruments are found inoperable when containment integrity is required. In that, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> had not elapsed between achieving criticality (when primary containment was

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required) and when this issue was identified to the plant staff, NYPA was able to satisfy the requirements of the action statement.

Technical specification (TS) section 3.0.D prohibits changing modes when the conditions for'

a TS limiting condition of operation are not met. NYPA changed modes with both H/O, monitors inoperable on January 2. This was not in adherence with the approved facility

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technical specifications. The safety significance of this violation was minor and

administrative in nature. Standard technical specifications exempt these monitors (and all accident monitoring instrumentation) from the requirements of TS 3.0.D in that tne unit could start with these monitors not operable as long as the required grab samples are taken.

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Also, the monitors were functional and power level was low when the monitors were -

recalibrated. Based on the criteria specified in Section V.A of the Enforcement Policy (i.e.,

the violation was not willful, it was of minor safety significance, and the licensee initiated appropriate corrective action prior to the conclusion of the inspection period),- this TS

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violation was not cited.

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Following entry into the LCO action statement, the Instrumentation and Controls (I&C) staff modified the H/0 calibration method to accommodate the higher levels of containment -

oxy;,.n (the procedure was written for an inerted environment). The B H/0 monitor was

calibrated and the LCO action statement was exited on January 4. The licensee's immediate corrective actions to restore a monitor to an operable condition were prompt and well

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executed.

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A critique was conducted to determine the cause and recommend corrective actions for this.

- i event. The critique identified the apparent cause as I&C's failure to properly identify the -

impact TS Amendment 181 had on H/0 operability. Amendment 181 changed the format

of Table 3.2-8 to include the reactor modes when the accident monitoring instruments are required (this was done to more closely parallel standard TS). The new table required the

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H/0 monitors to be operable when in "Startup/ Hot Standby" or "Run". However, the

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inspector did not agree with this apparent cause, because TS 3.2.9.A, which was unchanged,

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requires containment monitoring whenever primary containment is set. This observation was discussed with the I&C supervisor who concurred with the inspector's assessment that the TS

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amendment did not precipitate the event. Notwithstanding, the inspector concluded that the recommended corrective actions in the critique were comprehensive and included revising the test procedure and adding H/0 operability to the pre-startup checklist to preclude a

recurrence.

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6.0 ENGINEERING AND TECHNICAL SUPPORT (93702)

6.1 Drvwell Insulation Survey

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On July 28,1992, a Swedish plant, Barsebeck II, experienced an event where a pilot

operated relief valve was inadvertently opened and the discharge jet from the valve dislodged and fragmented some mineral wool thermal piping insulation. The discharge from the valve

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caused a high drywell pressure condition which actuated the drywell sprays and the sprays were allowed to continue their operation to reduce drywell pressure.

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trapped on the ECCS and spray system pump suction strainers, causing a high differential

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pressure across the strainer within approximately 50 minutes of system initiation.

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Differential pressures across the suction strainers of the containment spray pumps became i

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high enough to cause a reduction in NPSH on the containment spray pumps and lead to a failure of one spray pump.

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The NRC issued an Information Notice (IN 92-71) describing the event and requested the

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BWROG to confirm that the U.S. BWRs were not as susceptible to strainer clogging. In.

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order to assess FitzPatrick's vulnerability, the inspector conducted an insulation and strainer survey. The inspector determined that FitzPatrick is replacing all of its high density fibrous insulation (mineral wool) with low density fibrous insulation. This modification is

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approximately 80% wmplete. The plant has no s'eam valves which discharge directly into the drywell. The ECCS suction :, trainers in the suppression pool are designed to provide j

minimum NPSH in a 50% clogged state. Additionally, there are numerous means of -

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providing make-up water to the reactor, independent of the suppression pool. The plant has

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no installed backwash system for the strainers, but based on the above information, the inspector concluded that FitzPatrick is not vulnerable to the Barsebeck incident.

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r 7.0 SAFETY ASSESSMENT / QUALITY VERIFICATION (71707,' 93702)

7.1 Project Meeting i

On January 20, the inspector attended the monthly project meeting. The project meetings are.

for corporate and station managers to periodically meet together on site to discuss their current and planned workloads for the purpose of direct face-to-face communications and ensure appropriate coordination and prioritization of effort. The meeting was well-attended by both staffs, including the Executive Vice President and the majority of his direct reports.

Topics of discussion and presentations made during the January 20 meeting included: startup

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and power ascension status; proposed 13-week rolling schedule, proposed modifications for the 1993 maintenance outage; current licensing issues; and reracking of the spent fuel pool.

The inspector observed good discussions and broad participation by those in attendance.

a 7.2 Engineering Meeting On January 13, the inspector attended the monthly engineering meeting. The engineering'

i meetings are held on site for corporate and site engineering staffs to discuss their current and planned work activities. In conjunction with the monthly project meetings, the engineering

meetings further promote direct fact-to-face communications between the corporate and site -

technical and engineering staffs. Topics of discussion and presentations made during the _

January 13 meeting included: proposed revision to Section 16.5 of the Updated Final Safety

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Analysis Report; the planned 1993 maintenance outage; the proposed outage modifications and the prioritization methodology; the fire protection improvement program (long-term)

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action plan; and the cable separation (long-term) action plan. The inspector observed good discussions and broad participation.

7.3 Ouality Assurance Department Item Tracking Review

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During an earlier inspection (reference inspection report 50-333/92-15) the inspector reviewed the technical resolution of the inappropriate substitution of non-traceable all-thread

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material in a safety related application. This item was identified by a contractor quality assurance inspector during a routine surveillance. The NRC inspector noted that an adverse quality condition report (AQCR) number was assigned for this concern when it was

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identified, but that there was no apparent mechanism to ensure the AQCR was being

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processed or cancelled. In response to this administrative oversight, NYPA quality assurance

(QA) management conducted an audit of all their tracking logs (i.e., Procurement Deficiency Log, Work Activity Inspection Report Log, Surveillance Report and Audit Logs). No

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discrepancies were noted and the AQCR discussed above appeared to have been an isolated case. Since early 1992, all open AQCRs have been tracked and trended. Biweekly trend and summary reports are now distributed to senior site and corporate management.'.During the Wednesday 8:30 a.m. site planning meetings, the QA department representative reviews with all department managers the weekly AQCR summary report. Overdue or coming due AQCR responses are highlighted and escalated for senior management action. The inspector.

concluded that the AQCR tracking mechanisms currently in use are appropriate and.that the AQCR involving the all-thread material was an isolated case.

7.4 Industry Event Followun Prior tr unit restart, the inspector reviewed NYPA's internal assessment and corrective-actions associated with the Salem nuclear power plant turbine failure discussed in NRC Information Notice No. 91-83. NYPA's review of this event and a number of similar industry events was summarized in Operating Experience / Vendor Technical Information (OE/VI) Review Report No. 910791, dated March 17, 1992.' The inspector concluded NYPA's review of this event was satisfactory and notes that a re-evaluation was performed by the technical services systems engineering staff, as documented by memorandum JTS-92-0442, dated June 1,1992. The re-evaluation was performed to verify that the earlier NYPA

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assessment and corrective actions were reasonable compared to present (higher) J. A.

FitzPatrick performance standards. The JTS-92-0442 memorandum concluded the initial event review response was satisfactory. The inspector concurred with this assessment.

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7.5 Manacement Oversicht

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Management oversight during the startup and power ascension was comprehensive. Many of the management oversight functions were stipulated in the FitzPatrick 1992 Startup Plan. A senior nuclear manager (SNM) was assigned to each shift throughout the startup progra z a

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The SNM was a senior operations manager who in addition to providing oversight, assisted

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the shift supervisor in directing support for non-operational matters. Additionally, the corporate Vice President, Nuclear Support, who was assigned to the site throughout the

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startup, and the Resident Manager provided a management presence in the control room during several key events.

The Startup Plan divided the unit startup into seven milestones. Each milestone was defined by a different set of plant conditions. Prior to proceeding from one milestone to the next, the Plant Leadership Team (PLT), consisting of the Resident Manager and the three General-Managers, would convene a milestone review meeting. The PLT would receive reports and assessments from the department managers and their recommendations to proceed with the unit power ascension. The Resident Manager authorized each milestone and exercised positive control over the entire startup program.

Included in the reports to the PLT were the results of the Startup Self-Assessment Plan. The Self-Assessment Plan called for the department managers to perform frequent field observations during each milestone and report their findings to the PLT. The observations were frequently done on the back-shift or deep back-shift and appeared to be thorough. Most assessments were quite self-critical, and discrepancies were either corrected immediately or incorporated into the administrative or de'iciency tracking systems. - F@owing completion of the final milestone, a critique will be held to identify lessons leamed, strengths, and weaknesses. Overall, the management oversight of the startup has been very good and the -

self-assessment plan has been well-executed.

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8.0 MANAGEMENT MEETINGS At periodic intervals during the course of this inspection, meetings were held with senior facility management to discuss inspection scope and findings. In addition, at the end of-the period, the inspectors met with licensee representatives and summarized the scope and

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findings of the inspection as they are described in this report.

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