IR 05000267/1987009

Insp Rept 50-267/87-09 on 870401-30.No Violations or Deviations Noted.Major Areas Inspected:Operational Safety Verification,Previously Identified Insp Findings, Surveillance,Maint,Esf Walkdown & Security

ML20215L386
Person / Time
Site:	Fort Saint Vrain
Issue date:	05/06/1987
From:	Bennett W, Dubois D, Farrell R, Jaudon J, Michaud P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20215L379	List: IR 05000267/1987009 ML20215L376
References
50-267-87-09, 50-267-87-9, NUDOCS 8705120222
Download: ML20215L386 (10)

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APPENDIX U. S. NUCLEAR REGULATORY COMMISSION

REGION IV

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NRC Inspection Report: 50-267/87-09 License: .DPR-34'

Docket: 50-267 Licensee: Public: Service Company of Colorado (PSC)

, Facility Name: Fort St. Vrain Nuclear Generating Station Inspection At: Fort St. Vrain (FSV) Nuclear Generating Station,

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Platteville, Colorado and PSC Offices, Denver, M Colorado

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Inspection Conducted: April 1-30, 1987

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Inspectors:

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R. E.'Farrell, Senior Resident-Inspector.(S.RI)

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0, f210 P. W. Michaud, Resident Inspector (RI) Date N YOkb D. L. DuBois, Reactor Inspector 'Da te W/E& R. Bennett, Project Engineer 5/6lf?

Date Project Section A, Reactor Projects Branch Approved: . M I h J Chief, Project Section A Date J . P/.R6rPacto (udonBranch ects kD

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Inspection Summar Inspection Conducted April 1-30, 1987 (Report 50-267/87-09)

-Areas Inspected: ~ Routine,-unannou'ced n inspection of operational. safety

. verification, previously identified inspection findings, surveillance,

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maintenance,. ESF walkdown, and securit '

Results: .Within the six areas inspected, no violations or deviations were'

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DETAILS ~ Persons Contacted Principal Licensee Employees D. Alps, Supervisor, Security

• F. Borst, Manager, Support Services / Radiation Protection L. Brey, Manager, Nuclear Licensing and Fuels R. Burchfield, Superintendent, Betterment Engineering Nuclear Results

• R. Craun, Manager, Nuclear Site Engineering D. Evans, Superintendent, Operations

• C. Fuller, Station Manager

• M. Ferris, Manager, QA Operations W. Franek, Superintendent, Plan / Scheduling & Stores D. Goss, Coordinator,' Nuclear Licensing and Fuels

'*J. Gramling, Supervisor, Nuclear Licensing Operations

• M. Holmes, Manager, Nuclear Licensing F. Novachek, Manager, Technical / Administrative Services

• T. Prenger, QA Services Manager

• P. Tomlinson, Manager, QA R. Walker,-Chairman of~the Board and CEO D. Warembourg, Manager, Nuclear Engineering-

• Williams Jr. , Vice President, Nuclear Operations The NRC inspectors also contacted other licensee and contractor personnel during the inspectio * Denotes those attending the exit interview conducted April 28, 198 . Previously Identified Inspection Findings (Closed) Open Item (267/8429-01): Shift Supervisor Approval on Changes to Standard Clearances. The NRC inspectors reviewed Administrative Procedure P-2,-" Equipment Clearances and Operation Deviations," Issue 13,

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dated September 10, 1985, and Station Manager Administrative Procedure SMAP19, " Processing Equipment Clearances and Operation Deviations," Issue 6, dated December 1, 1986. These procedures require the shift supervisor to initial any modification to clearance The NRC inspectors sampled clearances-in the control room and found they were all in compliance with the above procedures. This item is considered close (Closed) Open Item (267/8415-05): Procedural Corrections to Procedure RP-5. The NRC inspectors reviewed Procedure RP-5, " Con ., Rod Drive Checkout," Issue 13, dated May 6, 1985. The procedure incorporated the proper changes. This item is considered closed.

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.(Closed) Violation (267/8415-08): Performance of Test Without Shift

~ Supervisor _'s Signature. This violation was considered as corrected in NRC

' Inspection Report _50-267/84-15. No. further action was required.. This item

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is considered close (Closed) Violation (267/8414-01): SysteminDeviationStatusWithout 0perations Knowledge. Corrective actions were reviewed and found to be incomplete in NRC Inspection Report 50-267/85-14. Further corrective actions, including a : revision to the equipment operator's-log ~to check the System 146 surge tanks pressures, have been completed.- This item is considered

close (Closed) Violation (267/8414-03): Failure to Follow Startup Procedures.-

This violation involved'a shift supervisor and a senior reactor operator-failing to properly fill-.out .the master check list contained in-Procedure OPOP Corrective actions have been completed and no furthe infractions in this area were noted. This item is considered close '(Closed)Open. Item (267/8413-03): . Identification of Safety-Related Work

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on PTRs. Specification SR-6-2 identifies all safety-related structures,.

systems, equipment, and components. This specification is used by reactor operators to identify-safety-related work. Operators are trained and

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knowledgeable on Specification SR-6-2. This item is considered close . 0perational Safety Verification The NRC inspectors reviewed licensee activities to ascertain that the facility is being operated safely and in conformance with regulatory requirements and that the licensee's management control system is effectively discharging its responsibilities for continued safe operatio The review was conducted by direct observation of activities, tours of the facility, interviews and discussions with licensee personnel, independent verifications of safety system status and limiting conditions for operation, and review of facility record Logs and records reviewed included:

. ~ Shift supervisor logs

. . Reactor operator logs

.. Equipment operator logs

. Auxiliary operator logs

. Technical. Specification compliance logs

.. Operations order book

. Operations deviations reports

. Clearance log

. Temporary configuration reports

. Station service requests (SSR) ,

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During tours of accessible areas, particular attention was directed to the following:

. Monitoring instrumentation

. Radiation controls

. Housekeeping

. Fluid leaks

. Piping vibrations

. Hangers and seismic restraints

. Clearance tags

. Fire hazards

. Control room manning

. Annunciators During the inspection period, the licensee was authorized and commenced restart of the plant after an 11-month outage to meet the requirements of 10 CFR Part 50.49, " Environmental Qualification of Equipment." No simulator exists for this plan Consequently, the operators did not have an opportunity to practice at power control manipulations during the outag Although oth nuclear plants have experienced outages of this duration, FSV entered tne most recent outage very shortly after a control rod drive overhaul outage of even longer length and only operated 30 full-power equivalent days since June of 1984. Consequently, NRC regional management deemed increased NRC attention during this startup was warranted.

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Regional management anticipated some problems with startup because of the length of the outage and the extent of the plant equipment rework during the outage. Augmented inspection coverage was plannec to monitor the lin see's handling of problems as they arose and to confirm satisfactory ret . or operator performance. There were equipment problems encountere Thes a e, obiems were handled expeditiously and appropriately wi'.h a comrvati'.e approach to safeguards. Engineers, technicians, ar.d plant merv ,sieat worked arnund the clock to solve problems in order to expedite a smooth plant startu Operator performance, as observed by the NRC inspectors, has been very good with a high degree of control room decorum observed and excellent operator morale noted. Operator morale appeared to be improving as plant power level increase During restart activities, the NRC resident inspection staff maintained an almost 24-hour a day presence onsite with a very strong emphasis on control room activity. The NRC inspectors observed, among other things:

. Very high management presence in the control roo . Close attention to detail, including repeated checks of LCO compliance beyond the normal frequency of checks required by Technical Specifications and procedure _

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. Operators coming in on their own time or hours early for shift to participate in startup activitie . Strong engineering support of operations from the various engineering group . A very conservative approach to plant operation, which was a marked improvement over previous plant operations witnesse . An~ apparent team spirit and cooperation not previously noted at FS . An orderly and controlled startu The plant startup, which commenced on April 17, 1987, was conducted in a manner that allowed NRC regional management on April 26, 1987, to release the plant from a 10 percent thermal power hold so that the plant is now authorized to operate at 35 percent power (at FSV the turbine generator is not put on the line until the plant reaches 28 percent of full power).

The plant management plans to operate up to 35 percent power until NRR completes the review of safe shutdown cooling, at which time it is anticipated that the plant will be released to operate at power levels up to 82 percent power. Modifications necessary to support the 82 percent power level were completed during the recent outage, and these modifications have been inspected by the NRC inspector No violations or deviations were identifie . Surveillance Because of the length of the most recent outage,11 months, and the long time since the plant was refueled (June 1984), the NRC inspectors met with plant management and requested that plant management provide a technical justification for any refueling interval surveillances which were not going to be performed prior to plant restart. Plant management did perform this review and did evaluate each refueling interval surveillanc These surveillances were either performed or an acceptable technical reason, precluding the necessity for reperformance of the surveillance, was provide The NRC inspectors witnessed performance of SR 5.6.lb-SA1 and -SA2, " Loss of Outside Power with the Main Turbine Generator Tripped (Part One and Two)." The first run of this test showed Emergency Diesel Generator (EDG) 1A failed to meet a requirement to reach 480V 148V and 60 HZ *1.2 HZ within 10 seconds. In addition, EDG "1B" failed to maintain 60 HZ 11.2 HZ when loaded. Adjustments were made to the engine governors, which had recently been rebuilt by the vendor, and the test was

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reperformed. During this second run, EDG "1A" met all its acceptance criteria but EDG "1B" again failed to maintain frequency as it was loade EDG "1B" started and obtained the required voltage and frequency within 10 seconds, but as loads were automatically sequenced, frequency continued to decrease to a value of approximately 56.5 HZ.

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The governor vendor's representative was brought in to troubleshoot the

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problems with FDG "1B", and found the speed droop signal was not being removed as it should have been in the independent mode of operation. The control circuit. board for EDG "1B" was replaced and the new board calibrated satisfactorily. A third " Loss of Outside Power" test was performed and all acceptance criteria were met, with EDG "1B" maintaining approximately;60.5 HZ.' No other discrepancies were note During the. loss of offsite power test, a wire is lifted in the diesel generator load sequence cabinet to assure that the train to be tested is loaded first with the larger loads. Normally this cabinet logic selects the first diesel generator set to come to speed as the one receiving the first loads. There was a step in the test procedure to restore the lifted lead following completion of the test. There was, however, no requirement for independent verification of the wire retermination. The plant has successfully completed its independent verification requirement according to NUREG 0737 and is not committed to the 1976 version of ANSI N18.7, which requires independent verification in instances such as this. The plant is, however, in the process of adding independent verification requirements to surveillance procedures. These requirements have already been added to maintenance procedures, system lineup procedures, and clearance procedures. The lack of independent verification in this was case considered a bad practice and identified as such by the NRC inspectors. Plant management agreed and immediately added an independent verification requirement in this case. Other surveillance procedures are being reviewed by the licensee, and independent verifications are being added as cases are identified that require such verification. This is considered to be an open item pending followup by the NRC inspectors (267/8709-01).

Prestartup surveillance testing of control rod drive and orifice assemblies (CRD0As) was monitored by the NRC inspector SR 4.1.1.D-X,

" Full Stroke Scram Test," SR 4.1.3.B-R, " Channel Functional Test of the Rod Pair Redundant In and Out Limit Switches and the Analog and Digital Rod Position Indicating System," and SR-0P39-X, " Control Rod Drive and Orifice Operability" were performed concurrently. These tests checked and verified the following:

. Digital and analog rod position indications and rod in light lit prior to withdrawa . Rod in light clear and proper outward rod motion indicate . Rod out light lit and further outward motion inhibited at full out positio . Digital and ar.alog rod position indications at full out and after scra _ _ _ . . _ _ _ _ .___ _ _ . _ . . _ _ _ ._ _ _ _ . _ _ _ _ _ - _ _ _ _ . _ . __

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. Scram tim . No slack cable indicatio Also performed in conjunction with the scram test were back-EMF measurements on each CRD0A. All tests were completed satisfactorily. The NRC inspectors reviewed test data, and no discrepancies were note At approximately 5 percent reactor power, the NRC inspectors observed calibration of the logarithmic power instruments. SR 5.4.1.1.5c-M/

5.4.1.4.3c-M, " Wide Range Channel Heat Balance Calibration" was performed to correct these power indications using the primary gas heat balance as the reference power level. Two of the three wide range channels were ddjusted, with all channels in agreement after the calibration. This surveillance will be performed again when power is high enough to perform a secondary heat balance calculatio SR 5.4.3-A, " Core Region Outlet Temperature Instrumentation Calibration,"

was observed by the NRC inspectors. This surveillance inserts a calibrated reference thermocouple into the reactor outlet plenum. This calibrated thermocouple is positioned in proximity to each of the 37 core outlet region thermocouples to check their accuracy. The NRC inspectors witnessed moving of the calibrated thermocouple to several positions in the core, data taking at each location, and health physics controls and monitoring of this operatio No violations or deviations were identified in this inspection are . Maintenance The NRC inspectors witnessed testing of DC circuit breakers in the control power supplies to the 480 VAC vital switchgear. Six 100 amp and twelve 50 amp breakers supply DC control power through an automatic throw-over (ATO) relay. arrangement to the three 480 VAC vital buse These DC circuit breakers were tested as part of a PSC commitment in Letter P-87008 of January 9,1987, to verify operability of the ATO scheme and to ensure a fault on one DC bus is not automatically transferred to the other DC bus. The test included long time delay and instantaneous overcurrent trips on both phases of each breaker. All breakers passed these tests with the exception of one 100 amp breaker which failed the instantaneous trip on one phase. This breaker was repaired and retested three times satisfactoril No discrepancies were noted by the NRC inspector During the loss of outside power test, the NRC inspectors observed portions of FT2191-1, a functional test to verify correct operation of relays which had been replaced or upgraded during the outage. All relays performed satisfactorily and no discrepancies were note The motors on P-2109 and P-2110, " Emergency Water Booster Pumps," were replaced with EQ motors during the outage. These pumps are used to boost

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= [ firewater pressure to drive the helium circulator pelton wheels if l required. A functional test of these pumps caused them to trip on motor l overload. An engineering evaluation showed the pumps to be operating above their design capacity. This pump runout resulted in excessive current in the motor. The pumps were disassembled and the impeller wear rings were found to be within tolerance but extremely worn. This allowed excessive recirculation flow within the pump, which coupled with the system flow caused the pump to operate above its design curves and thus overloaded and tripped the motor. The wear rings were replaced and the NRC inspectors observed the functional test, which was completed satisfactoril .

The Loop I Feedwater Control Valve, HV-2205, has an associated circuit which monitors electrical continuity through the valve's solenoids. This valve is part of the SLRDIS system and as such is safety-related. The monitor for one of the solenoid valves indicated an open in the circuitry, although tests showed continuity and operability of the solenoi Troubleshooting observed by the NRC inspectors indicated the problem was due to the solenoid valves which were replaced as part of the EQ program had a higher impedance than the previous, non EQ solenoids. The trickle current through the new solenoids used by the monitor circuit indicate circuit continuity was insufficien A permanent modification was made to the monitor circuitry to allow a sufficient trickle current to pass through the new solenoids. The NRC inspectors reviewed this modification for technical adequacy and observed the installation and functional test No discrepancies were note No violation or deviations were identified in this inspection are . ESF Walkdown In preparation for the plant's restart following an extended outage, plant management, as observed by the NRC inspectors, conducted redundant walkdowns and reviews of all system status in configurations both safety-related systems and systems important to operation. Critical valve lists were walked down, verified, and reviewed. Equipment operability was verified by system configuration, functional tests, and surveillance test All engineered safety features were verified to be in correct configuration and operable, both by plant management and the NRC inspectors prior to plant startu No violations or deviations were identified in this inspection are _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ._ _ _ . __

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10' ' Security-During the inspection period, the NRC inspectors observed compliance with the facility's security pla Perimeter security, detection aids,

' personnel access, compensatory measures, area patrols, and security event reporting were observe No violations or deviations were identified in this inspection are . Management and Exit Meetings During the inspection period, the NRC inspectors conducted several meetings with the licensee senior management to review progress towards outage completion and preparation for plant startu Exit interviews were conducted on April 28, 1987, attended by those indicated in paragraph 1. At this time, the NRC inspectors reviewed the scope and findings of the inspectio . .

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