IR 05000220/1992012

Insp Repts 50-220/92-12 & 50-410/92-14 on Stated Date.No Violations Noted.Major Areas Inspected:Plant Operations, Radiological Controls,Maint,Surveillance,Emergency Planning, Security & Safety Assessment/Qualilty Verification

ML17056B921
Person / Time
Site:	Nine Mile Point
Issue date:	06/19/1992
From:	Larry Nicholson NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML17056B920	List: IR 05000220/1992012 ML17056B919
References
50-220-92-12, 50-410-92-14, NUDOCS 9207010090
Download: ML17056B921 (52)

Text

U.S. NUCLEAR REGULATORY COMMISSION REGION I-Report Nos.:

Docket Nos.:

License Nos.:

92-12; 92-14 50-220; 50-410 DPR-63; NPF-69 Licensee:

Facility:

Location:

Dates:

Niagara Mohawk Power Corporation 301 Plainfield Road Syracuse, New York 13212 Nine Mile Point, Units 1 and 2 Scriba, New York April 29 through May 23, 1992 Inspectors:

W. L. Schmidt, Senior Resident Inspector R. A. Laura, Resident Inspector W. F. Mattingly, Resident Inspector (in training)

C. D.

eardslee, ct spector, Intern Approved by:

Nicholson, Chief React r Projects Section No. 1A Division of Reactor Projects Alii':Thi i p

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f plant operations, radiological controls, maintenance, surveillance, emergency planning, security,

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and safety assessment/quality verification activities.

R~u~lt:

See Executive Summary.

9207010090 920619 PDR ADOCK 05000220

PDR

EXK IVE UIVHNARY Nine Mle Point Units 1 and 2 NRC Region I Inspection'Report Nos. 50-220/92-12 & 50-410/92-14 April29, 1992 - May 23, 1992 Plan rations NMPC operated Unit 1 and conducted refueling outage activities at Unit2 safely over the period.

At Unit 1 the operators performed well in response to a high neutron flux reactor scram.

Plant management showed good safety perspective by investigating a very small increase in drywell unidentified leakage rate, leading to the identification of cracking in emergency cooling system valve bodies. The identification by operators at Unit 2 ofa potential problem with the secondary containment isolation valves indicated a very good safety perspective.

The operations department self-assessment activities conducted over the last year were an excellent initiative, which required minor improvements to provide management with the necessary feedback on operator performance.

Radiolo ical Controls

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Generally good radiological conditions were observed, except for minor problems with the disposal ofanti-contamination clothing. Unit 2 radiation protection personnel responded well to a large area contamination event in the reactor building.

Maintenance and Surveillance Maintenance activities were conducted well at both units.

The identification at Unit 1, of leakage from an emergency cooling system valve body during a drywell inspection was noteworthy.

At Unit 2, the conduct of reactor vessel head stud tensioning and motor operated valve testing was good.

The Unit 2 training program for mechanics was found to be comprehensive.

The maintenance on the Unit 2 Division I and II emergency diesel generators was very good, following the identification of tin smear on cylinder liners.

NMPC effectively tracked the work request backlog and implemented the preventive maintenance program.

Surveillance activities were conducted properly at both sites.

At Unit 2 the reactor shutdown margin determination and the local leak rate testing program were satisfactory.

The testing of the generator for current carrying capability, on the Division I and II emergency diesel generators, was conducted to a load slightly higher than the loss of coolant accident load Executive Summary {Continued)

Emer nc Pl nnin At Unit 2, with the unit in cold shutdown, the site operations review committee made a

conservative determination not to enter the emergency plan during preventive maintenance to a non-safety-related uninterruptible power supply.

This maintenance de-energized computers, which would have affected the emergency assessment capability of the unit during operation.

En in rin nd T hnical u

Unit I plant management and engineering methodically reviewed the cracking identified in the emergency cooling system condensate return valves.

NMPC'.s initial repair plan including both ASME,Section XI {code) and non-code repairs, had some technical merit. However, the NRC staff found the use of the non-code repairs unacceptable.

NMPC's response to NRC Bulletin 88-98 and its supplements, which addressed such cracking due to thermal fatigue, willbe reviewed to determine why this condition was not identified and corrected before it led to a through wall leak,

{220/92-12-01)

A previously unresolved item (220/92-07-02)

on inadequacies in the testing of the secondary containment leak rate and the reactor building emergency ventilation system was closed based on timely and effective corrective actions by NMPC.

At Unit 2, the involvement by the system engineer in the identification of the tin smear on the Division I'and II emergency diesel generators was excellent.

Review of NMPC's response to a GE service information letter, dealing with the ability of the main steam isolation valves to close or stay closed during a loss ofcoolant accident, resulted in an unresolved item (410/92-14-02).

ecnrit Adequate actions were when an individual was identified, for the second time, with a blood alcohol level above the fitness for duty limit Executive Summary (Continued)

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ualit V rifi i n Numerous licensee event and special reports were found satisfactory.

The deviation/event report system was found to be an effective method of identification and correction of safety-related

,concerns.

Failures ofbalance-of-plant equipment at Unit 1 caused five of the six reactor scrams over the last year.

NMPC management has taken action to review the turbine control system, which caused three ofthese scrams, forreliability-centered maintenance issues.

The DER system was being utilized by all site and corporate departments.

Apreviously unresolved (items 220/92-07-03 and 410/92-08-03) dealing with an observation that the technical specification corrective action system audit had not been effective was close TABLE F NTE 1.0 SUMMARYOF FACILITYACTIVI'HES..............

1.1 Niagara Mohawk Power Corporation Activities.......

1 r2 NRC Actlvltles

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2.0 PLANT OPERATIONS (71707,93702).....................

2.1 Plant Operations Review - Unit 1.... '................

2.2 Operational Safety Verification - Unit 2................

2.3 Operation Department Self-Assessment - Units 1 and 2..:....

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3.0 RADIOLOGICALAND CHEMISTRY CONTROLS 3.1 Routine Observations - Unit 1 and Unit 2...................

3.2 Contamination Event - Unit 2

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4.0 MAINTENANCE(62703)

4.1 Observation of Maintenance Activities - Unit 1 4.2 Observation of Maintenance Activities - Unit 2 4.3 Review of Work Request Backlog and the Program - Units 1 and 2

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Preventive Maintenance

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5.0 SURVEILLANCE(61726, 61707).... ~.......................

6.0 Emergency Cooling System Valve Body Cracks

- Unit

(71707, 93702, 6 1726)

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10 7.0 Cooper-Bessemer Emergency Diesel Generator Engine Tin Smear

- Unit 2 (62703, 93702)......,..............:..................

8.0 EMERGENCY PREPAREDNESS (71707)....................

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8.1 Review ofEmergency Planning forUninterruptible Power Supply Outage-U

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9.0 SECURITY AND SAFEGUARDS (71707).........-..............

9.1 Second Positive Alcohol Test on a Radiation Protection Supervisor - Unit

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13 10.1 ENGINEERING AND TECHNICALSUPPORT (71707, 92703, 37700)

0.1 Unit 1 1.

1.0.2 Unit 2

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Table of Contents (Continued)

11.0 SAFETY ASSESSMENT AND QUALITYVERIFICATION(71707, 92700)

11.1 Review of Licensee Event Reports (LERs) and Special Reports......

11.2 Station Operations Review Committee, Meeting - Unit 2 11.3 Balance of Plant Equipment Failures - Unit 1..............'...

11.4 Review of the Deviation/Event Report System - Units 1 and 2 11.5 (Closed) Unresolved Item 220/92-07-03 and 410/92-08-03:

Corrective

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ction Audits.....................................

A 12.0 MANAGEMENTMEETINGS...............................

• The NRC inspection manual procedure or temporary instruction that was used as inspection guidance is listed for each applicable report sectio DETAILS 1.0 SUMMARYOF FACILITYACTIVI'HES

Ni Mh k

n i'i The Niagara Mohawk Power Corporation (NMPC) conducted activities at Nine MilePoint Unit 1 (Unit 1) safely over the period. The unit operated at power until May 1 when an automatic high neutron flux reactor scram occurred due to a failure of the electrical pressure regulator (EPR)

in the turbine control system.

Based on a small increase in drywell unidentified leakage, maintenance mechanics found a 0.5 gpm leak from the body of an emergency cooling (EC)

system condensate return manual blocking valve during a drywell inspection followingthe scram.

Further investigation by NMPC identified cracking in the valve body castings in the return lines for both loops.

The unit remained shutdown during the remainder of the period, while NMPC developed a repair plan to address the cracking.

NMPC conducted refueling outage activities at Nine Mile Point Unit 2 (Unit 2) safely over the period.

Major refueling outage work included core reload, reactor pressure vessel (RPV)

reassembly, and inspection/maintenance on the Division II emergency diesel generator (EDG).

At the end of the period, Unit 2 was preparing for the RPV leakage test.

1.2

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Resident inspectors conducted inspection activities during normal, backshift and weekend hours over this period.

There were 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> of backshift (evening shift) and 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> of deep backshift (weekend, holiday, and midnight shift) inspection during this period.

A routine radiation protection inspection was conducted during the week of April 27.

The findings of this inspection will be documented in Combined Inspection Report 220/92-13 and 410/92-16.

On May 21, 1992, NRC Region I issued a Notice of Violation with a proposed Civil Penalty of $200,000 to Unit 1, as a result of: the isolation from the ultimate heat sink on January 23, 1992, as documented in Augmented Inspection Team Report 220/92-80, and the inadvertent bypassing of the anticipatory turbine trip reactor scram, as documented in Combined Inspection Reports 220/410/92-02.

NRC held two meetings with NMPC over the period in NRC Headquarters:

A May 18 meeting to discuss NMPC's proposed repair plan for the cracking in the EC valve bodie A May 19 meeting to discuss NMPC's.failure analysis on the Vnit2 'B'hase transformer that failed on August 13, 1991.

The presentation materials and a list of attendees from both meetings, were issued from NRC Headquarters by letters dated May 26, 1992.

2.0 PLANT OPERATIONS (71707,93702)

2.1 Pl nt rations Revi w -

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The inspectors observed and verified safe plant operations in accordance. with approved NMPC procedures and regulatory requirements.

Operators did well during the unit startup on April29 and in response to an automatic reactor scram on May 1. An operator on rounds showed good safety perspective by identifying that the No. 122 core spray topping pump motor heater was not:

working properly, allowing timely corrective actions.

The conduct of a drywell inspection following the May 1 scram showed a good safety perspective by NMPC management.

Regular tours were conducted of the following areas:

control room security access point core spray pump rooms reactor building radiological control point intake structure electrical switchgear rooms diesel generator rooms containment spray pump rooms

turbine building During the tours the inspector verified operability of engineered safety features and other safety-related systems and power sources (on-site and off-site). Operators responded according to alarm response procedures.

The operators effectively used the equipment'status log to track technical specification action statements for out-of-service equipment.

The station shift supervisors (SSSs)

consistently maintained well written logs, containing sufficient information to allow an independent reader to understand plant conditions, ongoing evolutions, and problems encountered by the shifts. The operators effectively used the turnover sheets, the markup log, and temporary modification log. Plant housekeeping controls, including the storage offlammable material, were adequate.

Control room and shift manning met technical specification requirements.

The SSSs properly-controlled access to and maintained a professional atmosphere in the control room.

2.1.2 t matic R ct r-cr m Operators were observed to have responded well to the May 1 reactor scram.

The automatic high neutron flux reactor scram, from 97.6% power, occurred on May 1 following a reactor pressure fluctuation resulting from a failure of the main turbine control system.

The turbine control valves momentarily traveled 10% in the close direction causing a 40 psig reactor pressure increase.

All other systems performed as designed and no safety or electromatic relief valves lifted. Reactor water level decreased to a minimum of 23 inches and the feed system started in

its high pressure core injection (HPCI) mode restoring water level to the normal operating band.

Operators initiated a controlled cooldown and placed the unit in cold shutdown.

The drywell inspection completed during the shutdown identified a 0.5 gpm leak from an EC condensate return line blocking valv'e, EC 39-02 (see section 6.0 below for further details).

The NMPC post scram review team methodically troubleshot, identified the most probable root cause, and directed the repairs to the main turbine control system.

An independent engineering review team analyzed the failure mode of the turbine control system.

NMPC concluded that the most probable source of the intermittent, partial valve closure was a momentary loss of signal from one of the four differential transformers (DT) in the EPR.

These transformers input the following information to the EPR:

turbine bypass valve position, control valve position, main steam header pressure, and EPR servo valve position feedback.

Troubleshooting activities included tightening and replacement of loose or corroded connections, continuity and resistance checks of the Dts, replacing of the EPR servo valve, calibration of the system (including adjustment of mechanical linkages), and verification of the power supply performance.

The inspector reviewed the information compiled by both review teams, attended several meetings, visually inspected the control system, and discussed the repetitive nature of this failure with the system engineer, The investigation by NMPC of the turbine control system failure was methodical and thorough.

Several discrepancies were identified and repaired.

A'pecial test to monitor the turbine control system willbe performed during the next startup.

Further, NMPC compiled historical data on turbine control problems'to allow further evaluation.

2.2 There was no safety consequence of this event because the failure of the turbine control system resulted in an automatic shutdown which placed the unit in a safe condition. Although the cause of this reactor scram was similar to the one that occurred on April 18, 1992, the failure mode of the turbine control system was different (EPR vs. mechanical pressure regulator failure).

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afet Verificati n -

ni 2 NMPC safely performed refueling outage activities in conformance with approved procedures and regulatory requirements.

Control room activities, including shift turnovers and crew briefings, panel manipulations, and operator response to alarms, were observed.

Regular tours of the plant were conducted to assess equipment conditions, radiological conditions, fire protection, security, general housekeeping practices and personnel safety.

In general, the inspector observed an acceptable level ofperformance and generally good conditions throughout the plan The inspector observed portions of fuel transfer from the spent fuel pool to the reactor vessel, from the refueling bridge and the control room, including: plant conditions, fuel handling and accountability, core reactivity monitoring, refueling equipment operability verifications,

'ousekeeping and loose object control, communication practices, and radiological controls. The inspector assessed that qualified and knowledgeable individuals conducted the refueling activities well, as required by the technical specifications and the approved procedure.

2.2.2 eview f ec nd n inmn l Drin Operations department personnel demonstrated a very good safety perspective and a questioning attitude when they identified that secondary containment integrity technical specification requirements were unclear.

Specifically, they questioned the requirements for primary containment isolation valves which would have been necessary, to function as secondary containment isolation valves during the outage.

This was possible with the primary and secondary containment free to communicate (i.e., the drywell hatches removed) when piping downstream of a primary containment isolation valve passed out of the secondary containment.

In this case the primary containment function of the valves was not required but they would be required to function as part of secondary containment.

Technical Specification Definition 1.3.8 for secondary containment integrity required that all secondary containment penetrations required to be closed during an accident are either capable ofbeing closed by an operable reactor building automatic isolation signal or closed by at least one manual valve, blind flange, or deactivated automatic valve or damper secured in its closed position.

Operations management wrote a deviation/event report (DER) and took appropriate actions to ensure that the requirements of secondary containment were met on these valves. Further, a technical specification interpretation was revised to give the operators clearer guidance on this situation.

The inspector assessed that these actions were appropriately completed before secondary containment was required for movement of irradiated fuel in the reactor building.

2.3 0 erati n De artment elf-A sessment -

nits l and 2 The inspector reviewed operations department self-assessment program used at both units.

An operations department instruction (ODI), at each unit, controlled the program requirements.

The inspector reviewed numerous completed self-assessment data sheets and interviewed several operations personnel who perform the inspections.

The programs were assessed to be an excellent initiative to provide NMPC management with more meaningful data to detect and correct performance issues.

However, areas for improvement in the implementation were noted, as discussed below,'hich made the program less than fully effectiv The ODIs were generally compatible with each other and contained guidance, in attachments, on performing assessments in the following areas:

control room activities shift turnovers surveillance testing operator rounds backshift operations training activities previous corrective actions Each attachment contained a number of specific, effective observation and evaluation assessment elements, each to be rated within a numerical scale.

Although the elements in each assessment area were good, the numerical rating scale did not allow adequate development of performance based assessments.

Operations management reviewed each completed attachment.

Each operations department implemented the self-assessment programs differently, with varying degrees of effectiveness.

AtUnit 1, on-watch senior reactor operators (SROs) and shift technical advisors performed the assessments in all areas, for their shifts, each calendar quarter.

Operations management reviewed this data at the end of the quarter and discussed any performance trends.

Quarterly, the plant manager received a report containing the final collective assessments.

The inspector reviewed numerous completed Unit 1 self-assessment data sheets and interviewed several personnel involved in the process.

The assessment data was generally weak, because it did not contain crew performance issues, but captured minor problems that confronted the crew.

Several assessors stated that there was a reluctance to fully develop and document negative performance issues, in order to avoid tension between crew members.

The inspector assessed that the use of on-watch personnel to assess themselves contributed to the problem.

Several completed quarterly reports to the plant manager were reviewed.

The 1991 third quarter report contained effective assessments; while, the 1991 fourth quarter report was rather vague.

Through discussions with operations management personnel, the inspector found that because of a lack of good performance data from the crew assessments, management supplemented the assessments with their own observations ofperformance.

During this inspection period, the Unit 1 operations manager suspended the performance of self-assessments due to the increasing backlog and poor quality of the data requiring review. A major revision to the self-assessment process was being prepared in an effort to make the assessments more performance based.

Although a positive initiative, the Unit 1 self-assessment process required some refinement to become more effective in identifying performance trend At Unit 2, the operations department manager, supervisors, and selected off-watch SROs performed the assessments.

All assessments were required to be completed once per calendar quarter for the whole department rather than each individual crew.

The operations manager reviewed the assessment data, but no quarterly meeting or report was utilized to identify performance trends.

The plant manager was generally aware of the assessment findings.

The assessment data at Unit 2 was generally good.

The inspector attributed this to the use of experienced off-watch SROs and operations department management which provided an objective view. The corrective action self-assessment was not performed three of the last four calendar quarters.

The operations manager stated that this was an oversight. In summary, the assessment

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data was generally good; however, the lack of a quarteily meeting and report to identify overall trends and necessary corrective actions, was not c'onsistent with the Unit 1 program.

3.0 RADIOLOGICALAND CHEMISTRY CONTROLS (71707)

3.1 R utine Observation

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nit 1 and nit 2 During a plant tour of the recycle pump room located in the Unit 1 turbine building, the inspector observed that a full anti-contamination clothing disposal basket, at a step-off pad, had overflowed onto the floor inside the contaminated area.

Unrelated to this, a used set of anti-contamination clothing coveralls was observed laying on the floor inside a posted contaminated area in the turbine front standard area.

These two observations were discussed with Unit 1 radiation protection management who initiated appropriate corrective actions.

These minor observations showed of a lack of attention when disposing of anti-contamination clothing.

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nit 2 Radiation protection (RP) personnel responded promptly and effectively when several floor and equipment drains backed-up contaminating most ofreactor building elevation 215. 'Access to the floor was immediately restricted to only those RP and operations'personnel necessary for the event response.

The quick response by RP prevented the contamination problem from developing into an airborne concern.

No personnel were contaminated during this event or the subsequent clean-up.

The cause of the drain system back-up was subsequently determined to be a design deficiency in the uncontaminated reactor building closed loop cooling (CCP) system.

Specifically, the CCP expansion tank was drained by siphon-action into the contaminated equipment drain system following a CCP system perturbation.

The amount of water drained exceeded the drain system capacity, resulting in the back-up, and subsequent area contamination.

The operations department initiated a DER to correct the CCP design deficiency that allowed siphoning of the expansion tank into the drain syste.0 MAINTENANCE(62703)

4.1 bserv tion fMin n n A ivi'

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The inspector observed selected portions ofthe preventive and corrective maintenance activities listed below, to verify:

adherence to regulations, use of administrative and maintenance procedures, conformance with codes and-standards, proper quality assurancel quality control (QA/QC) involvement, and proper equipment alignment and retest.

Work Request 197022, Replace the No. 102 diesel generator air compressor power supply breaker.

Work Request 197474, Replace the No. 102 diesel generator air compressor motor.

Maintenance personnel conducted these activities properly.

4.2 bservati n fM inten nce Activities -

nit 2 4.2.1 eac r V l

T i nin Contract maintenance personnel conducted the reactor vessel head stud tensioning correctly, with adequate oversight from NMPC mechanical maintenance, radiological protection, and quality assurance personnel.

The inspector also observed proper radiological practices to minimize personnel exposure and the spread of contamination.

4.2.2 Electrical Preventive M inten nce on Limi r V lve Electricians conducted a calibration of the valve operation test and evaluation system (VOTES)

sensor for containment spray isolation valve 2RHS~MOV25A following an approved preventive maintenance procedure.

The permanently mounted VOTES sensor and a dedicated computer system determine valve stem thrust by measuring yoke tension and compression through the entire length of a valve stroke cycle.

VOTES testing replaced motor operated valve actuator testing (MOVATs) as part of the inservice testing program for all safety-related valves.

A U-clamp stem calibrator measured the stem distortion during static and dynamic tension and compression.

The stem distortion measurements were then used to calibrate the VOTES sensor.

The inspector assessed that the maintenance personnel were knowledgeable of the VOTES technology and performed the procedure properly, using good electrical safety precautions and radiological procedure.2.3 R view f M h ni 1M in n

T inin Pr m

The mechanical maintenance training program functioned well.

This program consisted of numerous modules (i.e., pumps, valves, rigging, heating and ventilation, etc.) in which a person could become qualified. The qualification for each module consisted of classroom training, in-plant work under the supervision of a qualified mechanic, followed by an in-plant qualification by a supervisor who watches the person perform specific tasks.

A qualification sheet clearly outlined the requirements (tasks to be demonstrated) for each module.

A board posted in the maintenance shop tracked the progress of each individual by showing qualifications in specific modules.

The maintenance training coordinator tracked the progress toward qualification in specific modules.

Overall, the inspector found the program satisfactory.

4.2.4 Review of he cr nh u e Fish Return Lin Pi in During an earlier inspection period the inspector identified potential leakage for a weld in the screen house fish return line, unisolable from the lake.

During a plant tour the inspector identified rust around the weld just upstream of the screenhouse penetration.

The inspector, discussed this observation with the system engineer and determined that a work request (WR) had been written on this weld previously.

As part of this work the weld was cleaned and leakage was not apparent.

The engineer stated that there could be very slow leakage from the weld, possibly due to contamination during the welding process.

Further, the engineer stated that there.

was not reason to suggest that the weld was going to fail. The inspector walked down the piping and screenhouse structure and determined that if the weld were to crack there would be no significant impact on the safe operation of the unit. The only components that would be affected were the non-safety-related traveling screen wash pumps and the fish-jet motive pump. Ifthe traveling screens could not be washed, debris buildup'on the screens would eventually limitthe water flow to the SW bay and bay level would drop. When the level dropped a low level alarm would be initiated in the control room. Ifoperators did not take actions to restore the SW bay level, two safety-related (one powered from each electrical safety division) valves would automatically open to bypass the traveling screens and restore level.

The inspector concluded that NMPC had properly dispositioned this issue.

4.3 Reviewof W rkR tB kl an h

Pr v niv M ntenan Pr m-nit

/lid 2 The inspector determined that NMPC effectively tracked and managed the WR backlog at both units.

Management used goals and a good safety perspective to determine the timeliness in which Wrs were performed.

In addition, Unit 2 Maintenance took the initiative to perform a surveillance of Wrs to determine repetitive equipment failures.

This surveillance was initiated as a result of a similar beneficial surveillance conducted at Unit The inspector reviewed the preventive maintenance (PM) program and the process used tojustify deferring a PM. The maintenance management at both units were knowledgeable of the backlog in PM and the process used to track these items. The deferral process was adequately described in the applicable site procedure.

During the review, the inspector found a few overdue PMs which had not been properly deferred.

NMPC management took immediate appropriate actions to correct this oversight.

5.0 SVRVEILLANCE(61726, 61707)

5.1.1 bservation f urv ill n A ivi 'e -

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The inspector observed and reviewed portions of completed surveillance tests to assess performance in accordance with approved procedures and limiting conditions of operation (LCOs), removal and restoration ofequipment, and deficiency review and resolution.

Observed portions of the following surveillance tests were properly conducted:

Liquid poison system pump and check valve operability test Reactor building close loo 5.1.2 h td wnMr'inD m n p cooling system pump and valve test in-ni 2 The inspector reviewed reactor engineering shutdown margin (SDM) demonstration surveillance.

procedure N2-RESP-009 and its supporting technical references.

Technical specifications require a determination'f sufficient SDM before or during the first startup after each refueling.

A sufficient SDM ensures that:

the reactor can be made sub-critical from all operating and shutdown conditions, and the reactivity transients associated with postulated accident conditions are controllable within acceptable limits. The inspector reviewed calculations and assessed that the SDM demonstration method was technically adequate and verified that sufficient SDM was ensured for Fuel Cycle 3 following the refueling outage.

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ni 2 The local leak rate test (LLRT) program implemented during the current refueling outage met

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technical specification and 10 CRF 50, Appendix J requirements.

The I&Cdepartment tracked and scheduled LLRTs using the preventive maintenance/surveillance test (PM/ST) data base.

Individual component leak rate results were verified to have been properly recorded in the LLRT summary used to determine the total containment leakage.

The total containment leakage calculated for this outage was within the allowable limits.

The DER system was being effectively used to document LLRTfailures. The resolution to several DERs were reviewed and found adequate.

Overall, it appeared that the program was being carried out and tracked according to approved procedure.

The maintenance department took appropriate actions to return to service those components that failed an LLR.1.4 eview fEmer n

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ni 2 The inspector reviewed the testing of the Division I and II EDGs with respect to the generator total load (KVA) (the vector sum of the real (KW) plus reactive (KVAR) loads).

During a surveillance test run of the Division I EDG, the inspector identified that the diesel was adequately loaded to the continuous rating of 4,400 KW. However, the generator was loaded to 4400 KW, with 900 KVAR(0.97 power factor (PF)) producing a total load of 4491 KVA.

Per the Updated Safety Analysis Report (USAR), the continuous real load of 4400 KWwith the design PF of0.8 would produce a reactive load of3,300 KVAR,with a total load of5500 KVA.

This concerned the inspector since the generator testing was not conducted at the designed PF and thus, the design total current load was not being placed on the generator windings and voltage regulating components.

The inspector discussed this with the lead Unit 2 electrical design engineer who stated 'that a DER had previously been written on this issue.

The inspector reviewed DER 2-92-0293 initiated in February 1992 and discussed it with the manager of technical support.

The DER stated that the loss of coolant accident (LOCA) loads, per the USAR, would be 4,287 KW with a.87 PF

'reactive load 2,429 KVAR, total load 4,927 KVA).

In February 1992, the operations department performed a monthly operability test with a revised load of 4,400 KW with a 0.87 PF (reactive load 2,500 KVAR, total load 5,060 KVA)on the Division I EDG.

The DER closure was based on the 110% technical specification required EDG testing being conducted at a higher total load than the USAR LOCA analysis.

During the cyclic twenty-four hour run, the EDG is loaded to its two hour 110% rating (4,840 KW) with a.97 PF (1,213 KVAR, 4,989 KVA), which was slightly greater than the total LOCA load of 4,927 KVAR (4,287 KW, 2,429 KVAR). The inspector found this acceptable, but noted that the margin to the total LOCA load was small.

6.0 Emergency Cooling System Valve Body Cracks - Unit 1 (71707, 93702, 61726)

NMPC took aggressive actions to bound the problem following the identification of the cracking in the upstream portions of the EC loop 12 condensate return line blocking valve (39-02) body (isolable from the reactor by the valve).

During the investigation process, grinding exposed a two inch long crack through the 1.25 inch thick valve body wall. NMPC developed a repair plan to excavate the crack using air arching methodology and then do a weld repair.

The upstream check valve (39-04) was disassembled to ease welding. Internal inspection of the drain line revealed other cracks, This prompted NMPC to conduct further non-destructive examinations on this and the other EC loop drain connection on the manual blocking and check valves.

Several other cracks at drain line connections, including a 90% through-wall-crack in the unisolable side of valve 39-02, were identifie NMPC developed a repair plan and made a formal presentation to the NRC staff on May 18.

The plan included a combination ofASME Section XI(code) and non-code repairs to the various cracks.

The NRC staff found the use of non-co'de repairs unacceptable because of uncertainties in the ultrasonic testing and the repair techniques.

NMPC subsequently decided to pursue code repairs to the valves or to replace the valves.

Initial NMPC review determined that stresses due to thermal cycling induced the cracks. This cracking phenomena was discussed in NRC Bulletin No.,88-08 and its three supplements.

The inspector reviewed NMPC's responses, dated September 29, 1988, and December 16, 1991, to this bulletin.

In the 1988 response NMPC identified that the condensate return lines were susceptible to thermal cycling fatigue and that a non-destructive examination of the unisolable piping welds, heat-affected zones and high stress locations, would be conducted from the reactor coolant system piping to the manual blocking valve.

Unrelated to this, NMPC requested a

schedular exemption from 10 CFR 50, Appendix J testing of the normally shut EC return line check and containment isolation valves.

Major modifications were required to install these new valves that could be tested per Appendix J. NMPC also stated that these lower leak rates could resolve the potential for thermal cycling and therefore, took no other actions at that time

=-

regarding Bulletin 88-08.

The NRC staff initially approved this schedular exemption until the 1992 refueling outage and subsequently extended the exemption to the 1994 refueling outage.

In the 1991 response NMPC reported that inspections of the subject piping revealed no rejectable indications or evidence of cracking in the bottom third of the pipe circumference.

This examination did not include the manual blocking valve body, where the cracking was subsequently found.

The potential for thermal cycling was reevaluated and concluded that the

='nisolable sections of the return lines were not subject to temperature distributions that would result in unacceptable thermal stresses during normal plant operation.

The basis for this was the existence ofturbulent penetration, the mixing that occurs due to turbulent eddies in the unisolable piping, due to flowin the recirculation pump suction line, which would result in mixing ofwater in the pipe, thus preventing the development of thermal stratification.

Maintenance personnel identified the leak during a drywell inspection conducted to determine the source of increased unidentified drywell leakage.

The investigation to bound the extent of the cracking was thorough and reflected a good safety perspective.

Although the initial repair plan had some technical merit, the NRC staff found it unacceptable.

The NMPC response to NRC Bulletin 88-08 is being reviewed.

This was considered unresolved pending completion of the review.

(220/92-12-01)

7.0 Cooper-Bessemer Emergency Diesel Generator Engine Tin Smear - Unit 2 (62703,-

93702)

NMPC management took aggressive action, after obtaining new industry information, to investigate a possibly safety significant failure mode of the Division I and II EDGs.

On April 30, 1992, NMPC system engineering received a draft of the "Inspection Manual for Cooper-Bessemer (CB) Model KSV Diesel Engine Cylinder Liners, Pistons, and Bearings" dated April 8, 1992.

This guidance, from the CB users group (seven utilities responsible for 14 nuclear power plants), provided information on degradation ofcylinder liners, pistons, main and connecting rod bearings, and camshafts.

These specific components were selected due to the potential for a significant reduction in EDG reliability if degradation was not identified, or because the mode of degradation was somewhat unique to these engines.

The guidance for cylinder liner tin smear detection included two critical pieces of information, provided in the

.technical manual; specifically, identification ofthe thrust and non-thrust side ofthe cylinder liner and a description of tin smear, including diagrams and color photographs.

Tin smear is the transfer of the piston skirt's tin plating (which provides a bearing surface between the cast iron piston and the steel liner) into the microscopic pores of the chrome-plated cylinder-liner.

This apparently occurs when the oil film between the piston and liner breaks down or is insufficient to sustain the imposed side thrust loads and tin wipes offthe piston onto the liner. The draft inspection manual concluded that the most likelyroot cause oftin smear was excessive starts and rapid loading common to diesel engines used in nuclear applications.

Minor amounts of tin smear are expected (and have not caused problems) on the thrust side of the cylinder liner.

When the system engineer received the above information the Division I EDG was operable following a satisfactory five-year m'aintenance inspection and the Division IIEDG was two days from completion of its five-year inspection.

These five year inspections, conducted by NMPC and CB representatives, without the new guidance, had not identified cylinder liner tin smear.

However, the system engineer visually reinspected the liners on the Division II EDG from the crankcase, using the new guidance, and concluded that tin smear existed on the non-thrust side of at least four liners.

Further inspections, both boroscopic and visual from the cylinder head, concluded that nine of 16 cylinders had non-thrust side tin smear.

Tin smear on the non-thrust side of the cylinder liner was a precursor to the crankcase lube oil explosions at Pennsylvania Power &Light's Susquehanna nuclear power plants.

The probable chain of events was as follows:

tin transfers from the piston skirt to the non-thrust side of the chrome-plated cylinder liner this tin fills the microscopic pores of the liner degrading the lubrication of the cylinder assembly

increased heat due to friction causes crankcase lube oil explosion or piston lock-up NMPC replaced the piston, cylinder liner and rings for the nine affected cylinders in the Division.

IIEDG. Also, the piston wrist pin end caps and the lower oil scraper ring for all 16 cylinders were removed.

This change willallow more lubrication between the piston and cylinder liner from the crankcase lube oil splash and from lube oil directed to the underside of the piston crown through passages drilled in the connecting rod and piston wrist pin. At the end of the inspection period, four DivisionI EDG liners exhibited signs oftin smear based on a visual inspection from the crankcase.

Further inspections and similar corrective actions willbe performed as required on the Division I EDG.

The inspectors observed various portions of the corrective maintenance performed by CB representatives on the Division II EDG and the visual inspections of the Division I EDG. The

.safety evaluation for the piston modification was also reviewed. The followingconclusions were made:

system engineering identification, ownership, and focusing of management attention on the issue of the tin smear problem was excellent and indicative of a good questioning attitude; NMPC oversight of the corrective maintenance performed by CB was adequate, and the corrective actions taken by NMPC, with CB's concurrence, were appropriate.

General Motors manufactured the Division III EDG and both Unit 1 EDGs which were not subject to this phenomenon.

8.0 EMERGENCY PREPAREDNESS (71707)

8.1 Review of Emer en Pl nnin f r nin il P r

u

nl 2 The Unit 2 site operations review committee (SORC) conducted an adequate review of a planned PM on non-safety-related uninterruptible power supply (UPS) 1G.

SORC determined that the emergency plan did not need to be entered, due to a planned loss of assessment capability, with the unit in cold shutdown.

Disabling UPS 1G caused. the loss of the process, emergency response facility and safety parameter display system, and the GE transient analysis recording system computers.

NMPC planned adequate compensatory measure while the computers were de-energized, including enhanced operator tours throughout the plant. The inspector found this approach acceptable.

9.0 SECURITY AND SAFEGVARDS (71707)

9.1 Second Positive Alc h 1Teston a Radi ti n Pr

'

u rvisor-ni

On May 20, an employee tested positive for alcohol (blood alcohol ).05%) during a random fitness for duty test.

In October 1991 this individual tested positive for alcohol during another random'est.

Security investigated on both occasions and determined that the alcohol had been consumed outside the work area at least five hours prior to the supervisor's scheduled working tou After the initial confirmed positive test, the individual's unescorted access was temporarily suspended and he was referred to the employees assistance program (EAP). He met au his initial obligations in compliance withNMPC's nuclear division directive - fitness forduty (NDD-FDD),

continued to receive treatment through the end of March, and took several follow-up drug and alcohol tests, for which the results were negative.

NMPC took all the necessary actions required by NDD-FDD and management noticed no unusual behavior by the individual.

After the confirmed positive follow-up test on May 20, NMPC made a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> report to the NRC in compliance with 10 CFR 26.73.

In accordance with NDD-FDD, the individual's unescorted access has been suspended for a minimum of three years.

NMPC was performing a quality check of his work and willhold further discussions with the individual.

10.1 ENGINEERING AND TECHNICALSUPPORT (71707, 92703, 37700)

10.1 (/nit 1 10.1.1 nr lved lcm 22 2-7- 2 n

inment k T in Inspection Reports 220/92-07 and 220/92-11 discussed an inadequate test methodology employed by NMPC during secondary containment leak testing with the reactor building emergency ventilation system.

NMPC documented their review of this condition in LER 92-06, dated April 28, 1992.

This LER adequately discussed the specifics involved in a 1989 change to the testing procedure that resulted in the secondary containment being tested for leakage with the reactor building emergency ventilation system operating outside of its design basis.

Based on previous reviews of this situation and on the information in the LER, the inspector considered that NMPC took adequate and prompt action to correct this testing deficiency.

The inspector noted that the technical specifications for the testing of the secondary containment and the operability ofthe reactor building ventilation system were confusing, and conflicted directly with the USAR. These conflicts led in part to the change to the test methodology made in 1989.

In the LER NMPC committed to submit a change to the technical specifications to correct these errors.

The inspector considered this item close.2

@nit 2 10.2.1 Review f P

R n

E rvi Inf rm

'

tter477 The inspector reviewed NMPC's closure of GE service information letter (SIL) 477, dated December 13, 1988, and found that the actions to be taken, as documented in an operational experience assessment, had not been fully completed.

This SIL concerned the ability of the inside containment main steam isolations valve (MSIVs) to close, and stay closed, following a design basis LOCA. The MSIVs close on a differeritialpressure across the valve actuator piston, developed by nitrogen pressure, with the assistance of the closure springs, On a close signal the actuator shuttle valve repositions porting nitrogen to the top of the actuator piston and vents the lower portion of the actuator piston to the containment.

The concern was that during a LOCA the containment pressurization to the maximum pressure (35 psig) could reduce the differential pressure holding the MSIVs shut, and the valves could open or not be able to fully close.

Recommendation 3 ofthis SILprovided information that NMPC should complete a force balance calculation on the actuators to determine what nitrogen pressure was required to ensure that the valve would stay shut.

Further recommended was a safety-related low pressure alarm, with a setpoint based in the calculation, for either each nitrogen accumulator or the nitrogen supply.

Changes were also suggested to the plant operating procedures to require closure of the MSIVs ifpressure dropped to this setpoint.

The initial NMPC review, dated January 31, 1989, stated that GE had performed a force balance on the inboard operators and found that the minimum nitrogen pressure required to ensure that the valves would go shut was 74 psig.

Further, it stated that when pressure dropped to 70 psig the valves would go closed, as designed, due to repositioning of the actuator shuttle valve. The response also noted that there was no pressure instrument, in the drywell, that monitored the nitrogen pressure being supplied to the MSIVs, and that a modification would be required to provide such an indication.

It was also stated that the operations department would make changes to the. operating procedures.

Subsequent NMPC evaluation, dated April 12, 1989, changed the evaluation such that a modification was not necessary.

The inspector walked down the nitrogen supply system to the inboard MSIVs and determined that the only pressure instrument that supplied a remote readout was a non-safety-related transmitter on the large, nitrogen receiver located upstream of the containment isolation valves.

This instrument supplied a computer point, a computer alarm, and an annunciated alarm at 90 psig decreasing.

On the sensing line there was also a non-safety related gauge that would allow local pressure readings'in the reactor buildin.0

It concerned the inspector that at the time of the inspection no changes to the alarm response procedure had been made, and that the only instrumentation available to the operators in the control room was a non-safety computer point and annunciator.

The inspector reviewed the alarm response for this annunciated alarm and noted no indication that the MSIVs might be inoperable or should be shut, ifpressure dropped to 74 psig.

Further review of the isolation function of the containment isolation valves for the nitrogen supply indicated that the MSIVs would be removed from communication with the non-safety instrumentation ifthe containment isolation valves were shut.

The MSIV actuators, because of their attached accumulators and check valves, would initially be at or above the pressure in the nitrogen receiver, when the isolation occurred; however, due to leakage the pressure could reach the 74 psig or 70 psig when the valves would go shut, without the operators being alerted to a degrading condition.

The inspector discussed this issue with the QA operational events assessment (OEA) supervisor for Unit 2. Based on this discussion the OEA was re-reviewing the adequacy of the SIL closure.

The inspector considered this issue unresolved (410/92-14-02)

pending review of the OEA investigation.

10.2.2 lo ed nr lv Item 41 1-2 -0 Rea tor re I

i n lin a le

~earation This issue was left unresolved pending inspector review of the safety significance of the reactor core isolation cooling (RCIC) system cable separation operability concern and the DER initiation delay.

The inspector reviewed NMPC's response to this issue and concluded that the ability of

'the RCIC system to do its intended safety function was not affected.

Also, NMPC's plan to prevent delays in the initiation of a DER following the identification of a problem was satisfactory.

The inspector considered this issue closed.

10.2.3 Clo ed Unre olved Item 41 / 1-17-03:

ervice W ter heck Valves This item dealt with the frequent failure of the SW system pump discharge check valves to properly seat, based on a review conducted as part of an NRC check valve audit. The inspector found that NMPC has taken adequate corrective actions to identify, trend and correct the reasons for the frequent failures.

This included changing the internal connection arrangement of the check valve disc to the shaft to minimize stress on the connecting components.

Further, the inspector reviewed why a non-conformance was not written when the maintenance department identified that an anti-rotation key had mistakenly not been installed in the subject valve during prior maintenance.

In this instance, the maintenance supervisor adequately documented his thought process on the work request.

NMPC conducted a detailed root cause analysis on the failures and developed a schedule such that all the valves would be inspected for damage before the end of 1992.

The inspector found these corrective actions acceptable and closed this ite linit i 11.0 SAFETY ASSESSMENT AND QUALITYVERIFICATION(71707, 92700)

11.1 Review of Li n

Ev n R ERs i 1R 11.1.1 The inspector reviewed the following LER and found it satisfactory:

LER 92-08, Reactor scram on high neutron flux caused by failed mechanical pressure regulator servo motor position indicator.

11.1.2 Unit 2 The inspector reviewed the following Special Reports and LERs and found them satisfactory:

Special Report, dated March 9, 1992.

Reactor building ventilation gaseous effluent monitoring system (GEMS) and the main stack GEMS noble gas activity monitor both being inoperable for greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Special Report, dated May 8, 1992.

Failure to return the main stack radioactive GEMS to an operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of declaring it inoperable for a quarterly channel functional test.

LER 92-02, dated February 10, 1992.

Channel functional test and logic system functional test of the reactor protection system did not verify operation of the non-coincident high neutron flux trip logic.

LER 92-04, dated March 27, 1992.

Discrepancies with the instruments used to obtain vibration data during pump inservice testing.

LER 92-05, dated April 15, 1992, Failure to declare the Emergency DC busses inoperable while performing electrical surveillance tests on the Division I battery chargers.

LER 92-06, dated April22, 1992.

March 23, 1992, loss ofall off-site power and control room annunciators.

Details of this event are contained in IR 50-410/92-08 and the NRC Region I AIT report dated April 10, 1992.

LER 92-07, dated April27, 1992.

Inadvertent primary containment purge system isolation due to an uninterruptible power supply output breaker failure.- See IR 50-410/92-08 for further dlscusslo LER 92-08, dated April 25, 1992.

Inadvertent start of the Division IIIEDG and the HPCS, system during a refuelin'g cycle instrument calibration due to pers'onnel error. See IR 50-410/92-08 for further discussion.

LER 92-11, dated May 18, 1992.

Reactor building isolation due to a spurious high radiation level signal, See IR 50-410/92-13 for further discussion.

LER 92-12, dated May 12, 1992.

Improper testing of four RHR motor operated valves due to an inadequate technical review of test procedures.

Specifically, the test method used on the valves did not apply pressure differential in the direction of the design system flow.

11.2 tation erati n

R vi w mmi M

tin

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ni 2 I

The inspector attended the SORC meeting held to discuss, in part, a simple design change to'nhance the reliability of the nuclear steam supply shutoff system annunciator power supply.

Technical specification requirements for a quorum was met.

The depth of review and member participation demonstrated the proper safety'perspective and adequately fulfilled the SORC responsibilities.

11.3 Balance of Plant ui ment F ilur

-

nit

The mspector reviewed the plant scrams and shutdowns over the previous 12 months and identified that balance-of-plant (BOP) equipment failures initiated five of the last six reactor scrams.

This included three turbine control system failures, a current transformer failure, and a failed steam flow meter.

Further, BOP failures resulted in operation at less than rated power.

Some examples included a tube rupture in a high pressure feed water heater, mechanical vacuum pump drain tank isolation valve seat leakage, and a,feedwater flow control valve malfunction.

The Unit 1 plant and maintenance managers were aware of the negative trend and had initiated corrective actions.

A review of the effectiveness of the preventive maintenance conducted on plant equipment was initiated and scheduled to be completed over the next three to four years, Twenty systems, both safety-related and BOP, were selected.

The first system selected to be reviewed was the turbine control system.

Emphasis during these reviews willbe on performing increased reliability-centered maintenance.

The inspector assessed that this initiative to improve the quality and effectiveness of preventive maintenance was positiv '

11.4 vi w fth Devi n Ev n R m-ni

2 An audit conducted on the DER system found it functioning correctly to document adverse conditions.

The inspector selected and reviewed numerous DERs, both open and closed, generated by various departments on site and in the corporate headquarters.

With minor exception, the DERs were initiated for proper reasons and given sufficient priorities by plant management.

The closed DERs reviewed provided proper resolution to issues.

The inspector had the following observations:

Since its beginning in April 1991, the DER process has been changed several times to enhance its usability and efficiency. These changes appeared to be working to improve the process.

Plant workers understood the DER process.

The workers interviewed believed that the system functions to get adequate resolution to issues.

Operators identified many conditions that were appropriately documented on DERs, including entry into LCOs, personnel errors, and repeated equipment problems.

The quality assurance (QA) department management prepared weekly summaries of the outstanding DERs and tracks their completion; The safety review and audit board (SRAB) remains current on the DERs issued by routine briefings by the QA Vice-President.

Overall the inspector found that the DER system was functioning well.

11.5 losed nresolved Item 220/ 2-7- 3 and 41 / 2-0 -

rrec 've Ac i n A dits This item dealt with an inspector observation that the technical specification audits of the corrective action program were not fully effective.

The inspector identified that while the method chosen by NMPC to conduct these six month audits was acceptable, the implementation of this method was not effective.

In January 1991, NMPC decided to do this audit with the other normal audits.

The inspector reviewed several audits and determined that the only corrective actions that were being reviewed were those taken as a result of previous audit finding The QA Vice-President reviewed this concern and developed a checklist that willbe implemented as a part of the normal audits to ensure that the area of corrective actions is adequately audited, The inspector reviewed this checklist and found it acceptable.

Further, a recent audit of the nuclear engineering and licensing activities conducted in April 1992 showed adequate review of the corrective actions taken in that area.

The inspector considered that NMPC took adequate and timely corrective actions to address this weakness in the audit program and considered this item closed.

12.0 MANAGEMIPITMEETINGS At periodic intervals and at the conclusion of the inspection, meetings were held with senior station management to discuss the scope and findings of this inspection.

Based on the NRC Region I review of this report and discussions held with Niagara Mohawk repiesentatives, it was determined that this report does not contain safeguards or proprietary information.

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