IR 05000267/1989003

Insp Rept 50-267/89-03 on 890201-0318.Violations Noted. Major Areas Inspected:Lers,Reserve Shutdown Matl Removal, Operational Safety Verification,Radiological Controls, Monthly Surveillance Observation & Maint Observation

ML20245C834
Person / Time
Site:	Fort Saint Vrain
Issue date:	04/13/1989
From:	Westerman T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20245C810	List: ML20245C808 ML20245C831 ML20245C834
References
50-267-89-03, 50-267-89-3, NUDOCS 8904270410
Download: ML20245C834 (17)

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' APPENDIX B

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LU.S. NUCLEhR REGULATORY COMMISSION H REGION IV?

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NRC Inspection Report: 50-267/89-03- Operating License: DPR-34

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7 Docket: 50-267

'Licenseei Public Service Company of Colorado (PSC)

P.O. Box 840 v,.. ,

Denver, Colorado . 80210-0840:

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

Inspection At: FSV, Platteville, Colorado

,. .. Inspection Conducted: LFebruary .1 through March 18, 1989 >

, Inspectors: lR. E. Farrell, Senior Resident' Inspector (SRI)

P. W. Michaud,' Resident Inspector (RI)

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LApproved: *

M/3[ff T. F. Westerman,. Chief, Project Section 8 Datfe Division of Reactor Projects

Inspection Summary Inspection Conducted February 19through March 18, 1989 (Report 50-267/89-03)

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Areas Inspected: Routine,: unannounced inspection of onsite. followup of licensee event reports-(LERs),_ reserve' shutdown material removal, operational-safety verification, radiological controls,. monthly surveillance observation,

-monthly maintenance observation, and a coastdown and defueling' meetin Results: The licensee completed.the reserve shutdown material removal during this inspection period. Several equipment malfunctions resulted in various delays _during this effort, but support by all licensee organizations

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. contributed to the' completion of_this effort in as timely a manner as possibl Numerous abnormal or first-time operations and configurations existed during

'this. inspection period. An appropriate-level of training, pre-job briefing,-

and monitoring contributed to-the. licensee's success in recovering from the

inadvertent actuation of the reserve' shutdown system and the high moisture

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levels in the primary coolan One' violation was identified by the NRC resident inspectors (paragraph 6),

which involved incorrectly posted radiological boundaries. The licensee's health' physics organization performed in an effective manner in support of-numerous evolutions which included radiological. concerns, with the exception of this: violatio J414

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PDR ADOCK 05000267

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. .s The licensee's preventive maintenance program.was reviewed and found to contain some improvements, but is still not existing at a level which is desirable.

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-3-DETAILS Persons Contacted D.' Alps, Supervisor, Security

• Block, System Engineering Manager

• L. Brey,' Manager, Nuclear Licensing and Resources

• Coppello, Central Planning and Scheduling Manager R. Craun, Nuclear Site Engineering Manager

, *D. Evans, Operations Manager

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• M. Ferris, QA Operations Manager

• C. Fuller, Manager, Nuclear Production

• B. Gares, Executive Secretary to Vice President, Nuclear Operations

• J. Gramling, Supervisor, Nuclear Licensing Operations M. Holmes, Nuclear Licensing Manager

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• R. Hooper,' Nuclear Technical Training Supervisor i

• J. Johns, Supervisor, Nuclear Licensing Engineering 'j M. Niehoff, Nuclear Design Manager '

F. Novachek, Nuclear Support Manager

• H. O'Hagan,' Outage Manager q

• J. Reesy, Nuclear Support Engineering Manager i

• D. Rodgers, Nuclear Computer Services Manager

• R. Sargent, Assistant to Vice President, Nuclear Operations

• L.. Scott, QA Services Manager

• V. Snyder, Maintenance Dept. Manager

• P. Tomlinson, Manager, Quality Assurance

• D. Warembourg, Manager, Nuclear Engineering R. Williams, Jr. , Senior Vice President, Nuclear Operations l The N9C inspectors also contacted other licensee and contractor personnel during the inspectio * Denotes those attending the exit interview conducted March 21, 198 . Plant Status The reactor remained shut down throughout this inspection period. The licensee's efforts were directed towards recovery from the unplanned ;

actuation of the reserve shutdown material and the removal of moisture from the reactor coolant syste The reserve shutdown material removal effort was hampered with equipment malfunctions, but was completed satisfactorily on March 2, 198 Moisture removal from the reactor coolant system was being accomplished by an. evacuation of the prestressed concrete reactor vessel (PCRV) at the end of the report period.

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-4-3. Onsite Followup of Licen*ee Event Reports (LERs) (92700)

The NRC inspectors reviewed selected LERs to determine whether corrective actions as stated in the LERs are appropriate to correct the cause of the event and to verify these corrective actions have been implemente LER 87-15 reported a Loop 1 shutdown (ESF actuation) when the interlock

- sequence switch (ISS)lwas moved from " Low Power" to " Power." The loop shutdown was caused by an improper wiring configuration which provided a trip signal to both "A"' and "B" Helium Circulator's logic circuitry when the ISS was placed in the " Power" position. The improper wiring was a result of incorrect electrical drawings which showed certain cables to be determinate and some used as spares that actually were not. These incorrect drawings were used to ground spare ccaductors to reduce electrical noise during corrective actions in response to LER 86-28- The licensee's investigation of the discrepancy between electrical drawings and the as-found configuration concluded the cables in question should have been, but apparently were not, determinate during original plant construction. Corrective actions included determinating the two conductors, which corrected the discrepancy with the electrical drawings and also removed the false trip signal from the'"A" and "B" Helium Circulator's logic circuitr A special test (T-360) was also performed prior to startup to ensure the plant protective system (PPS) had no abnormal trip signals. These actions are acceptable to close this LE LER 87-22 reported a. loop shutdown (ESF actuation) with the reactor shut dow This occurred during a surveillance test due to the failure of an

electronic logic chip. The failed chip placed one logic channel in a trip l condition. When a technician tripped another channel as part of the surveillance test, the 2-of-3 logic was satisfied and caused a loop shutdown. The failed logic chip was replaced and the surveillance test was subsequently completed satisfactorily. The component failure resulted in a conservative (trip) condition as designed, and thus did not present an unanalyzed configuration. The licensee's corrective actions are considered suffkient to close this LE LER 87-28 described a loss of offsite electrical power event during performance of a postmaintenance test with the reactor shut down. The postmaintenance test was to be performed on the reserve auxiliary transformer (RAT) firewater deluge control relay. Part of the test procedure included lifting leads so the RAT would not actually trip. The procedure incorrectly epecified the leads to be lifted, and when the RAT deluge system was actuated per the test procedure, the RAT breakers tripped. The root cause of'this event was identified as personnel error by the licensee, in that the test preparer specified incorrect leads to disable the RAT trip function as intended. Additionally, administrative  !

procedures did not require an independent review of postmaintenance test '

procedures. The licensee corrected the deficient test procedure and reperformed it satisfactorily. The engineer involved was counseled on the accuracy required of test procedures and the need to ensure erroneous results of this sort are preclude The licensee's Administrative

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Procedure SMAP-23, " Post Maintenance Testing,"; Issue 6, includes  !

I Step 3.3.4, which requires an independent review for accurac These actions provide a sufficient basis to close this LE LER 88-04 reported a manual reactor scram from 74 percent power due.to an upset on the offsite electrical power grid. This grid disturbance caused ,

L several western power plants;ta shut down. The reactor operator manuallyJ scrammed the reactor just "; the turbine automatically tripped due to load swing The plant subse b .1tly experienced an unplanned radioactive gas release from the core support floor vent system, and a Notification of Unusual Event was declare The velease occurred when a safety valve in the core support floor vent system lifted due to flow restrictions in the'

downstream' piping. The total activity released was approximately 15 percent of technical specification (TS) limits and resulted in a total dose of 3.67 E-5 rem at 'the exclu: ion area boundary. The licensee performed a special test (T-383), on the core support floor vent system after cleaning strainers and replacing downstream filte,s. The results of this test indicated the relief valve setpoint (5 psig) sas too low, and that it should be changed to 10 psig. This setpoint '.nange was reviewed by the NRC inspectors and verified to be acceptable, both for protecting the-system from overpressurization and to provide a margin against unplanned releases. -The new setpoint has been verified to be successful in preventing unplanned releases from this path during subsequent reactor shutdowns. Additionally, wiring deficiencies found in the turbine control circuitry following this event have been corrected. This LER is close LER 88 2 06 reported a manual scram actuation from 71 percent power when all circulating water flow was lost. The loss of circulating water was due t an expansion joint failure on the "1A" Circulating Water Pump which flooded the circulating water pump pit. The expansion joint failed due to its age of approximately 15 years. The service life of these joints is approximately 10 years. The root cause of this event was the lack of a preventive maintenance program for rubber expansion joints. As a result of 'this event, all eight expansion joints in the circulating water pump pit were replaced. In addition, the licensee inspected and evaluated the condition of all expansion joints in the plan A number of expansion joints have been replaced as a result of these inspections, including those on tne condenser water boxes, diesel generator heat exchangers, and condensate pumps. The licensee has developed and implemented a preventive maintenance program for the inspection and periodic replacement of expansion joints throughout the plant. These actions are sufficient to close this LE . Reserve Shutdown (RSD) Material Removal (6071(

The seven-region group of the RSD system was inadvertently inserted into the reactor on January 19, 1989. The NRC inspectors closely monitored the licensee's preparation and implementation of activities to remove the RSD material'from the cor . . - _ - - _ _ _ _ - _ _ . _ _

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!Thehotservicefacility(HSF)wouldnormallyhavebeenutilizedtosetup

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,' the reserve shutdown vacuum too However, the HSF was configured for control rod drive' refurbishment. Since it was' desirable to leave the HSF '

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.in.this configuration ,and with the fact that it would have taken 2 weeks v .to' reconfigure the HSF, the licensee' decided to utilize the new fuel .

, loading' port (NFLP) to set up the RSD vacuum tool. A new Procedure,.

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! !MPF-1074, Issue 1, "In-Core Removal of RSD Material /NFLP," was written to perform this evolution. . The'NRC inspectors reviewed this procedure and found that it contained. adequate precautions and instructions.to ensure

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the, reactor was maintaird 'n refueling conditions and to specify responsibilities of personnel involved in its use. Cetailed steps were-

1. ' provided:to' move,sset.up, and tu t equipment. The licensee provided

, in-depth training on this procedure, as well as other procedures, and the overall-scheme of the RSD removal operation The NRC inspectors attended two of these training'sessiont and found them to be well organized and

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informativ The' vacuum tool consists of a 400 Hz motor attached to a vacuum impeller, a hopper to collect the RSD material, and a suction probe to be inserted into the RSD channel; all of which are housed inside the auxiliary transfer cask'-(ATC). The ATC provides shielding for personnel and a radiological boundary to handle control rod ~ drive and orifice assemblies and the~ reserve shutdown vacuum tool. An external motor generator provides' the 400 Hz power to the vacuum motor, and is connected via wiring inside the ATC. . Because of the configuration, the. Vacuum motor has minimal heat removal characteristics, and as such is limited to a 4 minute duty' cycle which must be followed by a minimum 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> cooldown perio The general sequence of activities to remove the RSD material for each of the'seven affected regions was: Perform a shutdown margin verification in accordance with Procedure SR 4.1.6.c/d-x, '" Shutdown Margin Evaluation. for In-Core Maintenance."

b.- Retract the control rods and insert a rewind tool to hold them after power is remove Install a reacter isolation valve (RIV). Remove the control rod drive and orifice assembly (CRD0A) using the AT l Transf er and store the CRD0A in an equipment storage wel Assemble the vacuum. tool, test, and place inside the AT Move the ATC to the RIV, open the RIV and vacuum the RSD material  ;

' from the region. This operation consists of lowering the vacuun

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probe until it contacts the_ RSD material (as indicated by a loss of weight"from the ATC grapple), then energizing the vacuum motor and slowly. lowering the probe into the RSD channel. When the end of the probe-is approximately 6' inches from the bottom of the RSD channel, a

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-7-mechanical stop prevents' further travel, and the probe is withdrawn into the AT h. . . Move the ATC to the NFLP, and extend and disconnect the probe and vacuum too Empty the vacuum tool hopper and weigh the RSD material to verify all the material was remove j.- Install a refurbished CRDOA.into the region using the AT The NRC~ inspectors attended shift briefings and observed good

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informational exchanges and overall coordination. Key personnel were interviewed to ensure their understanding of their responsibilities, and knowledge of procedures, administrative requirements, and actions to be taken under unexpected or abnormal conditions. Management involvement was observed throughout thisieffort. Overall, the RSD removal effort went well, despite problems which occurred at various stage Following ex-core testing, the first attempt to vacuum RSD material on February 5, 1989, was unsuccessful. Procedure MPF-1074 specified the starting. current for the vacuum motor should decrease to less than 40 amps within approximately 40 seconds. When it did not do so after 45 seconds, the motor was deenergized. Investigation of the problem found wiring damaged on the pigtail between the ATC grapple head and the vacuum moto In addition, external wiring on the ATC was.found to be damaged. This wiring was repaired on February 9. ,As part of the troubleshooting to determine the cause of the wiring damage, the 400 Hz motor generator was tested separately and found to be defectiv A new 400 Hz generator was obtained, installed, and tested satisfactorily on February 1 On February 14, the vacuuming effort was partially successful, but the l probe stopped approximately 2 feet before it should have.' An

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investigation discovered the vacuum tool hopper / canister had a restriction which prevented movement of the probe for the last 2 feet. This restriction was relieved and the RSD material from Regions 5 and 22 l was removed on February 1 Regions 3 and 34 were successfully vacuumed j on February 1 On February 19, while vacuuming RSD material from Region 28, the 400 Hz-vacuum motor failed. The motor was disassembled and sent out for repair on February 22. Alternate means of removing RSD material had been under 1 consideration from the outset. An auger was developed which was fairly successful in shop tests, but which tended to grind the RSD balls together, creating dust and other concerns. An external blower scheme was also

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developed which was similar to the existing vacuum tool, but utilizing an external blower rather than a motor lowered into the reactor. This external blower appeared to be a viable option and was reviewed by the NRC

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l Controlled Work Procedure (CWP)89-050, "In-Core Removal of Reserve l

Shutdown Material with Roots Blower" was developed by the license The I i

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NRClinspectorsreviewedCWP89-050andtheassociated'designinformation and safety evaluation. Concerns with this proposed method were discussed with the licensee and resolved. However, the 400 Hz vacuum motor was rebuilt and returned to the site before this alternate meth3d was implemente The existing vacuum tool was ceassembied and all.RSD material was removed from the core on March A visual examination was made on the RSD material removed from each of the seven regions. A sample from one high and one low boron concentration material was sent offsite to be chemically analyzed for leachable boron content, in accordance with TS 4.1.9 D.4. A visual inspection of one RSD hopper was also performe Verification that'all RSD material had been removed from each of the seven regions was accomplished by weighing the removed material and comparing it to records of the weight of material which had been installed. A tolerance of plus or minus 1 pound of material was developed and documented in licensee Memo PPS-89-0433, dated February 3, 198 The 1 pound tolerance was based on consideration of: The accuracy and calibration of the scale used to weigh the material Uncertainty over the accuracy of the original weights from records Verification that sufficient material was inserted to fill the RSD channel above the height of the active core The amount of material which could be left in the core with no significant effects on reactivity Any effects on axial or radial power distribution that any material left in the core would have Assurance that a subsequent discharge of RSD material into the same region would not overfill the RSD channel The NRC inspectors reviewed the licensee's calculations and evaluation of the 1 pound tolerance and found them acceptabl The RSD material removal from six of the seven regions met this acceptance criteri However, the. material removed from Region 25 weighed 5 pounds l more than that recorded as being loaded in 1985. Nonconformance Report

(NCR)89-050 was written to' address this discrepancy. The licensee's evaluation and disposition of this NCR concluded that an error must have been made during the loading in 1985 since there is no possible pathway for RSD material to migrate from one region to another. The NRC I-inspectors reviewed the licensee's actions in response to this NCR and l found them acceptabl l l

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_ h No violations or. deviations were identified in the review of this program area.

t 5. Operational Safety Verification (71707)

The NRC. inspectors made daily tours of the control room during normal working hours and at least once per week during backshift hours. Control room staffing was verified to be at the proper level for the plant conditions at all times. Control room operators were observed to be

) attentive and aware of plant status and reasons why annunciators were lit.

The.NRC inspectors observed the operators using and adhering to approved

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procedures in the performance of their duties. A sampling of these procedures by the NRC inspectors verified current revisions and legible copies. During control room tours, the NRC inspectors verified that the required number of nuclear instrumentation and plant protective system channels.were operable. The operability of emergency AC and DC electrical power, meteorological, and fire protection systems was also verified by the NRC inspectors. The reactor operators and shift supervisor logs were reviewed daily, along with the TS compliance log, clearance. log, operations deviation report (0DR) log, temporary configuration report (TCR) log, end operations order book. Shift turnovers were observed at least once per week by the NRC inspectors. Information flow was consistently good, with the shift supervisors soliciting comments or concerns from the reactor operators, equipment operators, auxiliary tenders,.and health physics' technicians. The licensee's station manager, operations manager, and superintendent of operations were observed to make j routine tours of the control roo l The NRC inspectors made tours of all accessible areas of the plant to j assess the overall conditions and verify the adequacy of plant equipment, radiciogical controls, and security. During these tours, particular 1 atter. tion was paid to the licensee's fire protection program, including j fire extinguishers, fire fighting equipment, fire barriers, control of I flammable materials, and other fire hazard l l

A walkdown of the purge vacuum system, reactor building ventilation system, reactor building area radiation monitoring system, control room ventilation system, 480 VAC essential power distribution system, and portions of the firewater system was performed by the NRC inspector I These systems were selected because of their relation to work performed j during various portions of the reserve shutdown material retrieval and '

l moisture removal efforts. Valve and breaker positions were verified, where possible. When affected by a clearance, the valves or breakers were

, verified to be positioned in accordance with the clearance requirement Power supplies for components in these systems were verified, but were also subject to clearances in some cases. During these system walkdowns, the NRC inspectors verified the operability of standby or backup equipment when components or portions of systems were inoperable due to clearance The NRC' inspectors reviewed seseral TCRs which were used to install equipment in support of the outage recovery efforts. Proper reviews and i

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approvals were verified for each TCR. Three of then TCRs were

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independently verified by the NRC inspectors:

TCR 89-01-01 installed a manometer on the refueling f;oor to read reactor pressure under refueling conditions (subatmospheric),

t TCR 89-01-04 supplied power to the reserve shutdown vacuum motor l generator from a spare breaker on Reactor Motor Control Center l

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TCR 89-03-02 deenergized. rod position. indications so that oxidation j of electrical connections for rods which were removed from the core

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would be minimize No discrepancies were noted during these walkdown The licensee estimated that approximately 250 gallons of water' entered the reactor coolant system through a leaking core support floor cooling water tube. This tube was known to have a leak and was inadvertently left unisolated during the reserve shutdown removal effort. The licensee determined it would take approximately 2 months to remove this amount of water through the purification system. In an effort to shorten the time to remove the water, and the associated dates for criticality and unrestricted operation, the licensee devised a setup to evacuate the PCR This evacuation to as-close-to-full-vacuum-as-achievable, was calculated to remove the water in approximately 1 to 2 weeks rather than 2 months using the normal purification system lineu The system for evacuating the PCRV was designed and approved under Change Notice (CN) 2916. The NRC inspectors reviewed this CN and the associated safety evaluation in detail. The PCRV was designed for a vacuum pressure of -12 psig as documented in FSAR Appendix E, Design Criteria DC-11- Evacuation of the PCRV was previously accomplished ~in 1976 and 198 CN-2916 included temporary.and permanent modifications to evacuate the PCRV through a refueling penetration, a cold trap to remove moisture, and the installed purge vacuum pumps The discharge of the purge vacuum pumps

'was directed to the reactor building ventilation system which flows through the reactor plant exhaust filters before exhausting through the plant stack. An additional flowpath from another refueling penetration to the reactor building ventilation system was used to initially draw a vacuum on the PCRV, via a commercial blower, at a faster rate than possible with the purge vacuum pumps. In addition, a separate cold trap and vacuum pump was used to evacuate the core support floor (CSF) in order to minimize the differential pressure between the CSF internals and the PCR The NRC inspectors examined the licensee's design and analysis of the PCRV evacuation process, paying particular attention to radiological concerns, structural integrity concerns, instrumentation and monitoring issues, and the potential effects of moisture saturated coolant on core component .;

The radiological concerns are addressed in paragraph 6 of this repor _ _ _ - _ _ _ _ _ _

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The ~ structural effects of the' evacuation on the core support floor p , concrete,. steel, vent system, and cooling water system were analyzed and

" documented. The effects on the'PCRV steel liner, primary closure, helium qcirculator shutdown seals, PCRV rupture discs, and region isolation valve '

seals were also documented in the design analysi . .The design analysis for CN-2916 also considered'the effects of moisture on--

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control' rod drive-(CRD) and RSD system operability. With the PCRV at a

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, vacuum, purge flow-to the.CRDs had to be secured. The analysis considered lthe' effects of: saturated vapor conditions on.these components for the 7 L days during which the PCRV was evacuate c The:11censee reviewed plant drawings and operating procedures, then walked down systems which could have possibly been damaged.from evacuation of the, PCRV to establish boundaries to isolate all affected instrumentation. All piping subject to a vacuum was double isolated, and all critical

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instrumentation had process lines removed or vented in case of failure of

.the double valve isolation. The isolation' valves were positioned and restored on an eight part clearance, which provided a means to return essential equipment'to_ service more easily should,an emergency have

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occurred. The NRC~ inspectors reviewed l portions of these clearances and found no' discrepancie The NRC inspectors also reviewed CWP 89-38, which provided instructions for.the installation and removal of equipment to perform the PCRV

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evacuation. Temporary tie-ins for process flows, cooling water, and -

drains ^were described for the installed. purge. vacuum pumps, the core support floor vacuum pump, and the commercial blower. The CWP was found to contain sufficient steps,-adequate detail, and provisions for QC

' involvement, where required; On March 12, 1989, the licensee entered a loss of forced circulation, with

1. a calculated allowable time of,193.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> before forced circulation of primary coolant.was- required. This provided a sufficient window in which to perform the PCRV evacuation. The NRC. inspectors witnessed the initial phase'of the evacuation, and subsequently' monitored its status. The PCRV

, evacuation proceeded smoothly, with approximately 180. gallons of wate removed from the PCRV at the end of this inspection period. The NRC s .< - ' inspectors will monitor the licensee's activities in completing the evacuation'and restoring from it.

The NRC inspectors. randomly verified that the number of armed security officers required by the security plan were present. A lead security y '

officer was on duty to direct security activities on each shif The NRC inspectors verifled that search equipment, including an x-ray machine,

. explosive detector, and metal detector,'was operational or a 100 percent hands-on search was conducte The protected area barrier was surveyed by the NRC inspectors to ensure it

.was not compromised by erosion or other objects. The NRC inspectors

, t observed that vital area barriers were well maintained and not '

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. 8-12-compromise The NRC inspectors also observed that persons granted access to the site were badged and visitors were properly escorte . Radiological Controls (71707)

The NRC resident inspectors observed health physics technicians performing surveys and checking air samplers and area radiation monitor Contamination levels and exposure rates were posted at entrances to radiologically controlled areas and in other appropriate areas and were verified to be up-to-date by the NRC inspectors. Health physics technicians were present to provide assistance when workers were required to enter radiologically controlled area The NRC inspectors observed workers following the instructions on radiation work permits concerning protective clothing and dosimetry, and observed workers using proper procedures for contamination control including proper removal of protective clothing and whole body frisking upon exiting a radiologically controlled are During this report period, the licensee was involved in a number of evolutions with radiological concerns. This provided the NRC inspectnrs with an opportunity to observe the licensee's health physics department dealing with moderately high contamination and radiation levels, which is not a usual occurrence at Fort St. Vrai Work performed in support of the reserve shutdown material removal and the evacuation of the PCRV involved direct communication with reactor internals, with associated high radiation levels and handling of contaminated equipmen The NRC inspectors observed thorough involvement and preparations by the health physics department in all phases of these evolutions. All personnel' involved were observed to adhere to requirements of radiation work permits (RWPs). ' The NRC inspectors reviewed several RWPs to assess their adequacy in relation to the area involved and the work to be performed. These included:

RWP 10571 - Reserve Shutdown Work in the New Fuel Loading Port

.RWP 10576 - Maintenance in the Auxiliary Transfer Cask RWP 10577 - Control Rod Drive Work in the Hot Service Facility RWP 10578 - Decontaminate / Modify Refueling Sleeve The overall radiological controls in support of evolutions during this report period were good, though one instance of noncompliance was found by the NRC resident inspectors. During the reserve shutdown removal effort, the refueling floor (Level 11) was made a radiologically controlled area (RCA), in order to control access. The entry point was established on Level 10 at the south stairway to Level 1 Because of relatively high background radiation levels and the licensee's desire to monitor for ;

potential contamination at the refueling floor, friskers were located at the top of the stairs on Level 1 Personnel leaving the refueling floor i

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. -13-l frisked clean on Level'11, then walked down to Level 10', where they signed out and exited the RCA. This arrangement was acceptable'but for the fact that the NFLP access was located in the area between the access point on Level 10 and the frisking station on Level 11. The NFLP was utilized for emptying the reserva shutdown vacuum tool after removal of material from the reactor. As such, the NFLP was a high contamination / airborne activity area and was controlled under a separate RWP. Due to the high background radiation levels at the access.to the NTLP, a'frisker could not be ;ocated ther Thus, personnel exiting the NFLP. removed their protective clothing at a step-off pad located at the NFLP access, then climbed the stairs to Level 11 to use the friskers located there. The result was that potentially contaminated individuals were traversing the same area which i personnel who had frisked themselves clean were'using to exit the RC This was brought to the licensee's attention and was corrected by locating i the RCA access point at the north stairway between Levels 10 and 11. The licensee was informed that the failure to' adequately establish and control access to radiological contrcl areas and contaminated areas is an apparent violation of NRC regulations (267/8903-01). > Monthly Surveillance Observation (61726J The NRC resident inspectors observed portions of surveillance testing in support of the reserve shutdown material removal effor The selected surveillance procedures were reviewed for conformance with TS ,

requirements, in terms of LCOs and acceptability of result ;

Administrative approvals and clearances when required were verified by the NRC inspectors prior to test. initiation. Test equipment was erified to be within its calibration cycle. Testing was performed by qualified personnel in all cases. Portions of the following surveillance procedures were observed by the NRC inspectors:

SR-MA-11-RX, Issue 1, " Reserve Shutdown Material Sample and Hopper Inspection" SR-RE-48-X, Issue 6, " Refueling - CRD Penetration Leakage Test" SR-4.1.3.C-X, Issue 2, " Control Rod Drive and Orifice Operability" SR-4.1.6.C/D-X, Issue 2, " Shutdown Margin Evaluation for In-Core Maintenance" i

SR-4.1.9.D.3-RX, Issue 1, " Refueling Penetrations Piping Examination" SR-4.1.9.D.4-RX, Issue 2, " Reserve Shutdown Hopper Functional Test" SR-5.2.15-A, Issue 16, "PCRV Penetration Interspace Pressure Calibration" SR-5.2.28-62-R, Issue 6, "Holddown Plate Bolting Examination" i

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No violations or devii ..os-were identified in the review of this j-program are j l

l Monthly Maintenance Observation (62703) 1 l

The NRC inspectors observed portions of numerous safety related maintenance activities during this inspection period. Most of these activities were related to the RSD material removal effort. During observation of maintenance activities, the NRC inspectors verified the licensee was at all times in compliance with TS LCOs and that redundant components were operable as required. Activities were accomplished by qualified personnel. utilizing approved procedure ,

The activities observed and reviewert by the NRC inspectors included:

Operation of the reactor building crane in' accordance with Procedure M0P-1001, Issue 5. <This crane was used' extensively in support of the RSD removal effort. The procedure provides guidelines for reoperation inspections in accordance wii.n ANSI B30.2 and NUREG 0612 as well as for operations involving heavy loads on the refueling floo Operation of the auxiliary transfer cask'(ATC) in at cordance with Procedure M0P-1006, Issue 1. The ATC provides shieldirg for personnel and a radiological boundary to handle controi rod drive and orifice assemblies and the reserve shutdown' vacuum tool. The procedure provides precautions and instructions for handling of various components, including expected weights of eac Use and inspection of rigging' equipment in accordance with Procedure M0P-1007, Issue Removal and replacement of equipment storage well covers in accordance with Procedure MPF-1100, Issue 1.

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Removal and installation of control rod drive and orificing assemblies (CRD0As) in accordance with Procedure MPF-1056, Issue This procedure provides instructions for moving a CRD0A with the ATC between the PCRV, equipment storage wells, ar.d the hot service facilit Installation and renoval of shielding adapters in accordance with Procedure MPF-1065. Issue Shielding adapters provide personnel shielding and a platform to mount the ATC on an equipment storage well, hot service facility port, or the new fuel loading por Removal and insta)lation of reactor isolation valves (RIVs) in accordance with Procedure MPF-1067, Issue 1, cnd operation of RIVs in accordance with Procedure M0P-1009, Issue 1. RIVs provide personnel shielding and a platform for mounting the ATC, fuel handling machine, and primary seal cleaning equipment on the PCRV. These procedures

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-15- i provide prerequisites, precautions, and instructions for handling, installation, removal, and operation of RIV Removal and installation of PCRV top head holddown plates in accordance with Procedure MPF-1091, Issue Removal and installation of' reactor penetration covers (secondary i closures) in accordance with Procedure MPF-1094, Issue Removal and installation of helium purification train covers in accordance with Procedure MPF-1095, Issue These activities were monitored by the NRC inspectors on a random basis during this report period. Quality control and health physics involvement in all phases of'these activities were observed by the NRC inspectors. No discrepancies were noted during observations of the above activitie During routine annual preventive maintenance on the "B" Diesel Generator, the licensee discovered three cylinders on one engine (K-9206-X) with less

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than~ nominal, th'ough acceptable, characteristic On'Febfuary 27, 1989, the licensee began maintenance on this engine under Station Service Request (SSR) 89500737. During disassembly, a metal piec'e,.later determined to be a coolant flow director, was found in the-engine coolant outlet plenum. An examination was made of the cylinder

. heads which had been removed for maintenance, and two coolant flow directors were found to be missing. These coolant flow directors channel flow to the center of a cylinder head to provide even heat removal characteristics. Two nonconformance reports (NCRs) were written as a result of these findings: NCRs 89-44 and 89-45. Because the type of heads involved could not be distinguished from other installed heads, the licensee decided to replace all 24 heads on all 4 diesel generator engines. New heads were obtained for the "B" Diesel Generator's engines and were installed and tested satisfactorily on March 13, 198 Replacement of the heads on tae "A" Diesel Generator's engines was in process at the end of this report period and will be covered in a future inspection repor On March 15, 1989, the NRC inspector met with the licensee's maintenance manager to review the licensee's preventive maintenant program. The preventive maintenance program is described in Proce 9 e MAP-1, "FSV Preventive Maintenance Program Description." The devc 3pment of specific preventiva maintenance requirements is described in Procedure SMAP-27,

" Preventive and Corrective Maintenance Equipment Review." The NRC j inspector reviewed the focus of these procedures with the licensee's :

maintenance manager. The development of maintenance procedures for larger activities and standardized station service requests with controlled work procedures for smaller activities was discussed in some detai Weaknesses in the maintenance procedures were attributed in many cases to the inconsistencies inherent in having different individuals developing varicus procedure The preventive maintenance program has resulted in an

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L overall: increase in the quality of maintenance procedures and consequently the work product. Improvements in the. licensee's ability to complete J

, preventive maintenance as scheduled suggest the program is having some success in shifting maintenance activities from a corrective focus to s preventive progra There are still some indications, however, that-the licensee's maintenance program is not yet where it should be. One example of this was the failure of Pressure Control Valve PCV-4256, which supplies a backup source of cooling water from the firewater system to the emergency diese'

generators' engine and room coolers, as well as other components. i'is valve had been leaking for over 2 years, causing the dow.1 stream relief valve (V-4599) to weep whenever the firewater header was pressurize On March 16, 1989, PCV-4256' finally failed during performance of SR-5.2.10.A.1-M, monthly firewater pump and instrumentation functional tes The failure of PCV-4256 caused relief valve V-4599 to cycle

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' rapidly, which caused repeated water hammer vibrations in the turb building firewater header. .These' vibrations resulted in the failure of the turbine lube oil reservoir room firevater deluge valve (HV-4507). The vibrations caused the disc of HV-4507 to fail, spraying firewater into the turbine lube oil reservoir room to a depth of approximately 6 inches before the header was isolated. The licensee subsequently repaired Deluge Valve HV-4507 and Pressure Control Valve PCV-425 No violations or deviations were identified in the review of this program are . Coastdown and Defueling Meeting (94702)

A meeting was held on March 7,1989 to discuss final coastdown operations and defueling of Fort St. Vrai This meeting was attended by the licensee and their consultants, NRR, and Region IV. Coastdown issues which were discussed included consideration of TS and FSAR limitations and criteria, reactivity effects, and power peaking issues. The licensee concluded operation during coastdown will be within the existing FSAR analyses and TS limitations, and will present no unreviewed safety questions. The licensee is scheduled to submit a coastdown safety j analysis report to the NRC by May. 31, 198 '

Various defueling plans and sequences were discussed, with emphasis on reactivity control and monitoring, accident and safety analysis, and computer modelin The licensee has decided to defuel by regions in the l reactor after considering the alternatives. This will include replacement !

of removed regions of fuel with boronated " dummy blocks" to maintain ]

structural integrity of the core during defuelin The defueling plan is l scheduled to be submitted to the NRC by May 31, 1989. This plan will include the defueling action plan, safety analysis report, and technical specifications.

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s.17 J A 10. ' Exit Meeting (30703)

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An exit meeting was conducted'on March 21, 1989, attended by those

. identified in paragraph 1. At this meeting, the NRC inspectors reviewed 2the scoperand findings of the inspectio . c. ,

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