IR 05000267/1987034

Insp Rept 50-267/87-34 on 871122-1231.Violations Noted. Major Areas Inspected:Operational Safety Verification, Followup of Allegation 4-86-A-119,followup of NRC Compliance Bulletin 87-002 & Monthly Security Observation

ML20149D499
Person / Time
Site:	Fort Saint Vrain
Issue date:	02/02/1988
From:	Farrell R, Michaud P, Westerman T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20149D458	List: IR 05000267/1987034 ML20149D456 ML20149D474
References
50-267-87-34, IEB-87-002, IEB-87-2, NUDOCS 8802090517
Download: ML20149D499 (16)

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h/ APPENDIX A

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U. S. N0 CLEAR REGULATORY COMMISSION

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r .NRC Inspection. Report: 50-267/87-34 License: DPR-34 3 i

Docket: 50-267  ;

Licensee: Public Service Company of Colorado (PSC) l I

Facility Name: Fort St. Vrain Nuclear Generating Station l i

Inspection At: Fort St. Vrain-(FSV) Nuclear Generating Station, i Platteville. Colorado and PSC Offices, Denver, (

Colorado  :

i Inspection Conducted:. November 22 through December 31, 1987 [

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Inspectors: , 1,2 /~ M R. E. Parrel F, Senior Resident Inspector (SRI)~ Date i

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P. W. Michaud, Hesident Inspector (RI) Date !

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Approved: 7* - ~ 7"If T. F. Westerman, Chief Date l Reactor Projects Section B

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8802090517 080203 >

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- PDR ADOCK 05000267 PDR Q

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Inspection Summary Inspection Conducted November 22 through December 31, 1987 (Report 50-267/87-34)

Areas Inspected: Routine, unannounced inspection of followup of licensee action on previously identified inspection findings, operational safety verification,-followup of Allegation 4-86-A-119, followup of Bulletin 87-02, critical rod height predictions, design control, procedural adequacy and-compliance, engineered safety features walkdown, reserve shutdown system, region peaking factor surveillance, monthly surveillance observation, monthly maintenance observation, radiological protection, and monthly security observatio Results:. Within the 14 areas inspected, 1 violation was identified (inadequate procedural controls, paragraph 8).

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' Persons Contacted Principal Licensee Employees  :

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• Williams, Jr. , Vice President, Nuclear Operations i D. Alps, Supervisor, Security l F.' Borst, Manager, Training [

• L. Brey,. Manager, Nuclear Licensing and fuels _!

R. Craun, Manager, Nuclear Site Engineering  !

D. Evans, Superintendent, Operations '

• M. Ferris, Manager QA Operations -

• C. Fuller, Station Manager

"J.' Gramling, Supervisor, Nuclear . Licensing Operations M. Holmes, Manager, Nuclear Licensing <

M. Niehoff, Manager, Nuclear Design

• F. Novachek, Manager, Technical / Administrative Services  !

• J. Reesy, Staff Assistant

• D. Scott, Manager, QA Services P. Tomlinson, Manager, QA  :

R. Walker, Chairman of the Board and CEO .

• D. Warembourg,~ Manager, Nuclear Engineering -l The NRC inspectors also contacted other licensee and contractor personnel  ;

during the inspectio ;

• Denotes those attending the exit interview conducted January 12, 198 [ Followup of Licensee Action on Previously Identified Inspection Findings

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(CLOSED) Open Item (267/8413-01): Incorporate LCO 4.2.2 Into LCO 4.2.1 - [

A review of LCO 4.2.2 determined that it WWre of a definition for t operable circulators than a limiting condition for operation. The  ;

licensee submitted a proposed Technical Specification change to revise ,

these LCOs in letter P-85042, dated February 6,1985. Changes to these '

Technical Specifications are being accomplished under th4' (2chnical [

Specification Upgrade Program, and will-not ha done as a separate license l amendment for these LCO ;

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(CLOSED) Unresolved Item (267/8203-01): Control of Temporary Plant  !

Configurations;(CLOSED)OpenItem(S67/8708-04): High Temporary Change i Backlog - Temporary configuration reports (TCR) were numerous and in some  :

cases very old. Many required permanent design change requests and  !

reviews. An enforcement conference on temporary configurations was held  !

in the NRC Region IV office on September 10, 1987. The licensee committed l to and has performed short term corrective actions as discussed in  ;

paragraph 7 of this report. The licensee's approach to future handling of 6 TCRs and action to correct and manage the design change process will be i tracked under NRC Violation 267/8717-0 These items are closed, j f

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(OPEN) Open Item (267/8632-02): No Procedure for Technical Specification Surveillance Requirements SR 4.1.9.0.3 and SR 4.1.9.0.4. The licensee has issued SR 4.1.9,0.4-R, "Functional Test of Reserve Shutdown Assemblies,"

which implements the associated surveillance requirements. A procedure 4 has not been generated for surveillance requirement SR 4.1.9.D.3, which requires a visual examination of the pipe sections which require disassembly and reassembly within the refueling penetrations at each ,

refueling outage. The licensee has committed to have these procedures in place by the fourth refueling outage. This item will remain open until the second procedure is implemente No violations or deviations were identified in the review of this program are . Operational Safety Verification The NRC resident inspectors reviewed licensea activities to ascertain that the facility is being operated safely and in con'ormance with regulatory requirements and that the licensee's management control system is effectively discharging its responsibilities for continued safe operatio The NRC resident inspectors toured the control room on a daily basis [

during normal working hours and at least twice weekly during backshift hours. The reactor operator arid shif t supervisor logs and Technical Specification compliance logs were reviewed daily. The NRC resident inspectors observed proper control room staffing at all times and verified operators were attentive and adhered to approved procedures. Control room instrumentation was observed by the NRC inspectors and the operability of the plant protective system and nuclear instrumentation system were ,

verified by the NRC resident inspectors on each control room tou Operator awareness and understanding of abnormal or alarm conditions was verified. The NRC resident inspectors reviewed the operations order book, operations deviation report (ODR) log, clearance log, and temporary configuration report (TCR) log to note any out-of-service safety-related systems and to verify compliance with Technical Specification requirement The licensee's station manager and superintendent of operations were observed in the control room on a daily basis, with the superintendent of operations frequently in the control room during the day and during l startups or special test '

The NRC resident inspectors verified the operability of a safety-related system on a weekly basis. The helium circulator bearing water and buffer -

helium systems, steam / water dump system, emergency feedwater system, control room ventilation system, and 480 VAC essential power distribution system were verified operable by the NRC resident inspectors during this .

report period. During plant tours, components of these systems were (

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inspected to verify valve positions, power supplies, and instrumentation j

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housekeeping was acceptable, but certain areas required additional effort for housekeeping cleanliness. This was discussed with the licensee's management and corrective action has been initiate Shift turnovers were observed at least weekly by the NRC resident inspectors. The information flow appeared to be good, with the shift supervisors routinely soliciting comments or concerns from reactor operators, equipment operators, and auxiliary tender A complete plant walkdown was performed by the NRC resident inspectors accompanied by the licensee's fire protection engineer on December 22, 198 Attention wac paid ;o any potential fire hazards, fire extinguisher inspection records, fire aigade lockers and equipment, and hose station No major discrepancies were noted, although an oil film was found on the boiler feed pumps and brought to the licensee's attention for cleanup and identification of the sourc No violations or deviations were identified in the review of this program are . Followup of Allegation 4-86-A-119 This allegation raised two separate concerns. One concern was the independence of the independent design reviewer process within the licensee's special projects group. The second concern was that the licensee had apparently reclassified a nonsafety related diesel generator .

as safety-related. These concerns are addressed below, Independent Design Verification Within Special Projects Group The NRC SRI interviewed the manager, an engineer, and an engineering technician of the special projects group. The special projects group is quite small, consisting of a manager, three engineers, one engineering technician, four draftsmen, and two non-engineering support personnel. Only the engineers and the engineering technician perform design work. Consequently, only the engineers and the engineering technician perform design review The procedures utilized by the special prJjects group are the same procedures utilized by the rest of the licensee's nuclear engineering divisio The procedures which govern design and independent design verification are ENG-1, "Control of Modification and Documentation Changes," dated January 23, 1987, and E0-100, "CN Preparation on Document Control," dated January 23, 198 While personnel from other groups within the nuclear engineering division can be made available to perform independent design verifications, it is the licensee's practice within each of the three nuclear engineering division groups to limit this to special cases when no personnel with the required expertise are available within the responsible group to perform the independent verificatio ____

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I The NRC SRI reviewed a sample of design change. packages and a log of completed change notices and concluded that the independent design verification process as performed by the special projects group was in compliance with NRC regulations and licensee commitment This item is close Classification of Nonsafety-Related Diesel Generator as Safety Related The license has, on site, a diesel generator known as the alternate cooling method (ACM) diesel. This diesel ge_nerator is used to power equipment for shutting down the plant in the event a fire renders other power sources', including the emergency diesel generators, inoperabl The licensee had classified this diesel generator as~nonsafety-relate In response to NRC Violation 267/8430-01, the licensee committed in letter P-85133, Gana to Johnson,'to reclassify this diesel generator as safety-related and treat it as such-in the futur The licensee completed the reclassification with Change Notice (CN) 1962, which changed the ACM diesel safety classification from nonsafety-related to safety-related. The NRC SRI reviewed CN 1962 and noted that it had a safety evaluation and had been approved by the plant operations review committe This item is close No violations or deviations were identified in the review of this program are . Followup of Bulletin 87-02 The NRC resident inspectors participated in the licensee'.s selection of fasteners to be tested in accordance with the requirements of Bulletin 87-0 Ten safety-related bolts, capscrews, and studs, consisting of material specifications A193 Gr.B7, A193 Gr.88, A193 Gr.88M, and A354 Gr.BD were selected. Ten safety-related nuts consisting of material specifications A194 Gr.2H, A194 Gr.8M, and A563 Gr.B were selecte The licensee's nonsafety-related stock is small and therefore a limited selection was available. In most cases, no material traceability exists for nonsafety-related fasteners. Ten bolts and capscrews and ten nuts were selected, which included samples of A194 Gr. 2H, 316 SST, and Grade 8 alloys, as well as some unknown material The NRC inspectors confirmed the information on these fasteners was properly transferred and recorded by the licensee on the appropriate data -

sheets and the samples were properly tagge . .__ The initial selection occurred on November 19, 1987, but some confusion over what to do about the lack of traceability on some of the nonsafety-related samples caused a delay in sending the samples out for testing until December 9, 1987. Additional samples of six items were required by the testing lab to complete the analysis. The NRC resident inspectors participated in the selection of these additional items on December 31, 1987, and verified they were the same as the original sample No violations or deviations were identified in the review of this program are . Critical Rod Height Predictions The NRC resident inspectors followed the licensee's actions to correct discrepancies between critical rod height predictions generated from the GAUGE computer code and predictions calculated manually in accordance with the licensee's Procedure 50P-12-02. This discrepancy was discussed in NRC Inspection Report 50-267/87-22, paragraph 3.b. New base reactivity and rod worth curves for 50P-12-02 were generated using the 7-Group GAUGE code (previously generated using a 4-Group GAUGE code) in order to improve the agreement between the two methods of calculation. These new curves were sent to the plant operations review committee (PORC) in accordance with Technical Specification 7.1.2 and to the nuclear facility safety committee (NFSC) in accordance with Technical Specifications 7.1.3 and 4.1.8 on December 11, 1987, for their respective approvals. Once approved, these new curves will replace the curves presently in Procedure 50P-12-02, which currently provides less accurate critical rod height prediction The predicted critical rod height for the December 11, 1987, startup was calculated using the 7-Group GAUGE code, the current SOP-12-02 curves, and the proposed new S0P-12-02 curves. The GAUGE prediction and the prediction from the proposed new S0P-12-02 curves gave values within the allowable deviation of plus or minus 0.3 percent reactivit The 50P-12-02 prediction using the current curves was less accurate. The NRC inspectors were satisfied that the new SOP-12-02 curves, once approved, will sufficiently address the critical rod height prediction discrepanc No violations or deviations were identified in the review of this program are . Design Control The NRC resident inspectors continued to follow licensee actions with regard to TCR committed to by the licensee in licensee letter P-87329, dated September 23, 1987. This letter was in response to an enforcement conference conducted September 10, 1987, in the NRC's Region IV offices in Arlington, Texas, and NRC Inspection Report 50-267/87-17. The NRC resident inspectors reviewed:

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. Station Service Request (SSR) 87510558, which reinstalled, calibrated, and functionally tested Thermocouple TE-1103 associated with TCR-85-12-25

. SSR 87510559, which reinstalled, calibrated, and functionally tested Thermocouple TE-1104 associated with TCR-85-12-25

. SSR 87510560, which functionally tested Thermocouple TE-1102 associated with TCR-85-12-25 ,

. SSR 87510561, which functionally tested Thermocouple TE-1101 associated with TCR-85-12-25

. Methodology for determining which drawings were affected by TCRs and required marking tc indicate pertinent TCR The NRC resident inspectors witnessed portions of the work performed on the above listed SSRs. The NRC resident inspectors also reviewed drawings in the control room and shift supervisor's office to verify that marking of drawings to indicate drawings affected by TCRs was accomplished at these location The NRC SRI concluded that the licensee had completed those actions that the licensee, in letter P-87329, committed to completing prior to operation above 35 percent powe No violations or deviations were identified in the review of this program area.-

8. Procedural Adequacy and Compliance The NRC resident inspectors reviewed three instances during the inspection period which demonstrated either inadequate procedures or inadequate adherence to procedures. They are as follows: Loss of Offsite Power On Lecember 7, 1987, the NRC resident inspectors witnessed the licensee's performance of a functional test of the reserve auxiliary transformer (RAT) deluge system after Auxiliary Relay CR-4505 had been repaired following the October 30, 1987, event reported in licensee event report (LER) 87-25. Procedure SSR 87509817 provided instructions to perform post-maintenance testing which included i isolating the transformer deluge system to prevent spraying firewater on the transformer, and lifting of a lead to disable the tripping of the RAT breakers when the fire signal was initiated, since house power was being supplied through the RAT from offsite source Technicians lifted the lead for Terminal 2 to 2C from relay 86RTD as instructed by the engineer in charge of the test. When the test was initiated, the RAT breakers tripped causing a loss of offsite power

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and the subsequent starting and loading of the emergency diesel generators. The lead which should have been lifted to prevent tripping the RAT breakers was for terminal 2 to 2C from relay 86R The licensee does not require independent review of post-maintenance-testing procedure Incorrect Wiring of Plant Protection System Relay Clearance 17895 was placed to electrically isolate plant protective system equipment to allow a cable repair. Four conductors were lifted. When the four conductors were reterminated, they were

' connected to terminal points 46, 47, 48, and 49 instead of points 46, 47, 48, and 53 as specified. This supplied 110 volt AC power to a

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7 volt relay resulting in damage to several relays and causing a spurious reactor scram signal on December 7,1987. At the time of the spurious scram, the reactor was shutdown with all rods inserte Failure of Emergency Lighting Batteries On December 21, 1987, the licensee energized the new emergency lighting system (not fully installed) for testing. This caused the new emergency lighting batteries to be damaged because a manufacturer installed grounding connection, which should have been removed prior to inplant installation of the equipment, was not identified for removal in the licensee's Control Work Procedures (CWP) 87-47 through 56, 182 and 183. The batteries short circuited to ground and were damaged beyond repai These examples indicate a lack of procedural adequacy, insufficient review of procedures, and a lack of compliance with procedures. This is an apparent violation of NRC requirements (267/8734-01).

9. Engineered Safety Features (ESF) Walkdown The NRC resident inspectors performed a complete walkdown of all accessible portions of the steam / water dump system. Section 6.4 of the FSAR, Technical Specifications 4.3.3 and 5.3.1, and Reference Design Manual Chapter 22 were examined by the NRC inspectors to review the system and requirements. The status and operability of the steam / water dump system was verified by:

. Comparing the as-found configurations with drawings PI-22-1, PI-22-6, j

and PI-22-5 to check their agreement.

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! . Checking instrumentation properly valved in and functionin . Inspecting equipment conditions and general housekeeping items which could degrade performance.

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The overall condition of the system-was good and no discrepancies were found by the NRC inspector No violations or deviations were identified in the review'of this program are . Reserve Shutdown-(RSD) System The NRC resident inspectors reviewed surveillance procedures, operating procedures, and records of tests and procedures performed and inspected the RSD system to determine that the licensee conformed to Technical Specifications, the FSAR, and approved procedure The surveillance procedures were reviewed to determine whether the procedures:

. Require application of test pressure to the RSD hoppers in accordance with the Technical Specifications and the FSAR Identify appropriate initial conditions

. Include acceptance. criteria relating to performance.of the actuating valves, integrity of the rupture disc, and performance of.the. orifice {

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. Identify proper valve lineup to each hopper for returning the system to normal and require verification of the valve lineup

. Require periodic calibration of the pressure indicators

. Require testing two hoppers and rupture discs, one hopper containing 20 wt% and cne containing 40 wt% boronated material, in accordance with-requirements in Technical Specifications and the FSAR

. Specify appropriate handling requirements, test conditions, and acceptance criteria

. Require inspection of the hopper after the rupture disc ruptures to determine whether all the balls were released from the' hopper

. Identify appropriate specifications for the replacement rupture disc and requirements for inspection of the replacement rupture disc prior to and following instaliation

. Specify requirements for inspection of the balls before returning them to the hopper or for the characteristics of replacement ball The following currently issued surveillances were reviewed for conformance l to the Technical Specification requirements: l l

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. SR 4.1.8.A/B-W / 4.1.9.A/B-W, "Reserve shutdown Helium and Nitrogen Bottle Pressure Check," Issue 2, dated March 7, 1986

. SR 4.1.8.C.1/2/3-A, "Reserve Shutdown Hopper, ACM Disconnect, and Low Pressure Alarm Test," Issue 2, dated December 6, 1985

. SR 4.1.8.D-A / 4.1.9.C-A, Channel Calibration of the Reserve Shutdown Gas Pressure Instrumentation," Issue 1, dated July 12, 1985

. SR 4.1.9.D.1-R, "Reserve Shutdown Valve Operability Test," Issue 1, dated July 12, 1985

. SR 4.1.9.D.2-R, "Channel Calibration of Reserve Shutdown Hopper Pressure Switch," Issue 1, dated July 12, 190 . SR 4.1.9.D.4-RX, "Rese ve Shutdown Hopper Functional Test," Issue 1, dated July 10, 1987 The NRC resident inspectors inspected documentation of surveillance tests performed during the last year for conformance of the tests and test teruits to requirements in the Technical Specifications, FSAR, and approved procedure No replacement rupture discs or absorber ball.5 were procured during the last year, and no modifications to the reserve shutdown system were made in the last yea The NRC resident inspectors performed a physical inspection of all accessible portians of the reserve shutdown system to determine whether:

. The volume of ;he individual hopper helium bottles conforms to the requirements in the FSAR

. The pressure on each individual hopper helium bottle exceeds the lower limit set forth in the Technical Specifications

. The valve lineup is correct in accordance with the S0P valve lineup checklist and in agreement with the P&I drawing No violations or deviations were identified in the review of this program area.

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11. Region Peaking Factor Surveillance The NRC resident inspectors reviewed Surveillance Procedures SR 5.1.7.a-X,

"Calculated Region Peaking Factors," and SR 5.1.7.b-X, "Region Peaking Factor Discrepancies," which were performed on December 31, 1987. The purpose of this review was to ensure that region peaking factors (RPFs)

are being maintained in conformance with regulatory requirements and that a satisfactory management system exists for control of RPFs. Technical Specifications 4.1.7 and 5.1.7 and FSAR Section 3.6.6 were reviewed by the

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NRC resident inspectors to verify the requirements for periodic review of RPFs and calculation of individual RPFs and percent discrepancie The input data for both the calculated and measured RPFs was reviewed by the NRC resident inspectors to verify the accuracy of the calculated RPFs and percent discrepancies. A check of the licensee's calculations for all percent RPF discrepancies and correction factors by the NRC resident '

inspectors produced no difference in the resultant values. Corrected RPFs were generated for four regions (21, 22, 29, and 31) and the NRC resident inspectors verified entry of these corrected values into the data logger compute The licensee performed these surveillance procedures and controlling RPFs in accordance with the requirements of the Technical Specifications and FSA No violations or deviations were identified in the review of this program are . Monthly Surveillance Observation The NRC resident inspectors witnessed performance of a number of post-maintenance functional tests and special tests prior to plant

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startu These tests were performed after maintenance activities had been performed on equipment affected by the October 2, 1987, fir On November 23, 1987, the NRC resident inspectors witnessed a test of the control room ventilation system to verify that positive differential pressure was maintained betwcen the control room and the turbine building and also between the control room and Building 10. The test procedure, FT-2713, was reviewed by the NRC resident inspectors prior to performance of the test to verify all modes of the control room ventilation system would be tested and to verify the appropriate administrative controls had been adhered to in preparation, review, and approval of the test procedure. Ten different modes or combination of modes of operation of the control room ventilation system were tested, which included all possible configurations of the system. In each operational mode, differential pressure measurements were taken between the control room and turbine building and between the control room and Building 10 under steady state conditions. The control room to turbine building door was opened and closed four times for personnel access prior to each set of measurements being take Under steady state conditions, the differential pressure between the control room and both the turbine building and Building 10 was maintained positive under all modes of operation. The proposed value for the upgraded Technical Specifications (PSC Letter P-87133, dated May 15,1987) of +0.125 inches of water between the control coom and turbine building was exceeded for all modes of operation. The lowest observed steady state value was +0.15 inches of water, observed in the "economy" mode with the auxiliary bay supply fan inservice; the highest observed steady state value was +0.425 inches of water, observed in the "refrigeration" mode. During door openings for personnel access, the differential pressure was generally maintained slightly positive, but did momentarily reach 0.0 inches of water on one of the 40-door opening

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cycle Control: room pressure was never_ negative with respect to the turbine building or Building 10. .The test was completed satisfactorily

- and verified the control room ventilation system to be. operating as designe Due_to the fact that "C" helium circulator tripped on overspeed for no known cause during the fire event of October 2, 1987, a test was performed on "C" circulator to verify its operability at high speed Test Procedure T-370, "Circulator Operation on Steam at low PCRV Pressures,"

was performed on December 5, 1987, and witnessed by the NRC resident inspectors. Circulator "C" achieved a speed of 9500 rpm during this test-with no indication of ary abnormalities. The test was completed satisfactorily and "C" circulator was declared operationa Following the loss of offsite power occurrence on December 7, 1987, "B" helium circulator did not self-turbine when auxiliary systems were restored. Performance of a special test, T-373, was observed by the NRC resident inspectors on December 8, 1987. This test increased the bearing water cartridge differential pressure on "B" circulator to 1000 psid (normally 600-650 psid) then rolled the circulator on steam at 800 to s 1000 rpm for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> in order to flush the bearing water surfaces inside the circulator. The circulator was then run on emergency condensate and simulated firewater (condensate through booster pumps) to demonstrate operability for safe shutdown and Appendix R cooldown configurations. The-circulator wa; spun on the first attempt and the test was completed .

satisfactoril The NRC resident inspectors witnessed special test T-372, "Verification of Proper Operation of Valves Closed by Loop Shutdown That Were in the Fire Area," performed on December 8 and 9, 1987. This test integrated the operation of all automatic valves in the fire area on a loop shutdown signal to verify their operability following fire recovery maintenanc The test was completed satisfactorily on Loop 1 on December 8, 198 When Loop 2 was tested, one valve, FV-2206, "Loop 2 Feedwater Flow Control Valve," did not stroke as required. The licensee found a limit switch not in the fire area or affected by the fire which required adjustment to initiate a permissive for this valve to strok The test was satisfactorily repeated on loop 2 on December 9, 1987. The NRC resident inspectors reviewed the test procedure ar.d associated safety evaluation and observed the test performances. No discrepancies were note During testing of the plant protective system on December 24, 1987, "C" helium circulator tripped due to a test signal being applied on one channel while a noise spike caused a trip of another channel of the circulator overspeed signal instrumentation. Circulator "C" would not self-turbine following this trip, but did roll freely on emergency condensat On December 26, 1987, an operability test was performed using the applicable sections of SR 5.2.7, "Circulator Operation on Emergency Condensate and Simulated Firewater." This operability test, which verifies the circulator's ability to meet minimum flow requirements for safe shutdown configurations, was completed satisfactorily. The

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circulator _would still not~self-turbine, however. An' engineering evaluation determined some type of debris (rust or corrosion products)

must have lodged in the bearing surfaces. The circulator was taken to approximately 5600 rpm on steam and run for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> in an effort to flush any debris from the. bearing cartridge.' This effort was successful as the circulator self-turbined subsequently. The NRC resident inspectors will follow the licensee's effort lto determine a possible source of debris in the bearing water system and to prevent future instances of this typ In addition, the NRC resident inspectors verified the following surveillance tests were completed prior to reactor startup:

. SR 5.2.10.Al-M, Functional Test of Fire Water Pumps

. SR 5.3.2-W, Partial Stroke Test of Main Steam and Hot Reheat Steam Stop Check Valves

. SR 5.3.4.B2-A, Safe Shutdown Cooling Valves Functional Test

. -SR 5.3.6-M, Functional Test of Instrument Air System Pressure Indicators and Alarms

. SR 5.3.10c-X, Calibration and Functional Test of Reheat Steam Instrumentation

. SR 5.4.1.2.6b-M, Reheat Header Activity Monitors Functional Test No violations or deviations were identified in the review of this program are '

13. Monthly Maintenance Observation The NRC resident inspectors monitored maintenance activities included in the licensee's fire recovery effort. Electrical work that was observed on a daily basis included cable pulling and terminations, wiring verifications, cold testing, and functional test Mechanical work included hydraulic snubber replacement, valve repairs, structural steel replacement, and post-maintenance testin Some specific items monitored by the NRC resident inspectors included:

. Replacement of damaged snubbers with qualified spare The NRC inspectors also verified two surveillance requirements were included

'in the maintenance procedures for each snubber replaced. These surveillance requirements were a visual examination in accordance with SR-5.3.8A-X, which was performed under Maintenance Procedure MP-2554 and met or exceeded all. requirements of the surveillance, and a functional test in accordance with ,

SR-5.3.8D-1.SY, which was performed under Maintenance Procedure MP-2550 and was included in the work package for each

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. Repair and replacement of the hydraulic actuator on HV-2292, Loop.2'

startup bypass isolation valve, which was the source of oil which caused the October fire. The NRC inspectors. verified the orifice plates were installed upstream of the thermal relief valves on the actuator. The NRC resident inspectors observed installation and adjustment of position indicating limit switches and a functional test in accordance with Procedure T-371, which checked the position-indications and stroke time (5.1 sec.; acceptance criteria of less than 10 sec.).

. Functional test of HV-2293, Loop 1 startup bypass valve, in accordance with FT-2711-2 This test verified stroke-time (5.3 sec.) and position indications following replacement of cables 18252 and 18253, which were in the fire are . Megger test of Cable 16020 and subsequent stroke test of valve PCV-5201, auxiliary boilers to 150 pounds steam pressure control valve, in accordance with SSR 8750958 . Repair and adjustment of main steam safety valves in accordance with Procedure MP-1010. The NRC resident inspectors observed hydroset testing of Loop 1 and Loop 2 main steam safety valves performed in accordance with Section 6 of this procedure. Hydroset testing of hot reheat safety valves will be performed at approximately 70 percent powe No violations or deviations were identified in the review of this program are . Radiological Protection The NRC resident inspectors verified that required area surveys of exposure rates are made and posted at entrances to radiation areas and in .

other appropriate areas. The NRC resident inspectors observed health l physics professionals on duty on all shifts including the backshift. The NRC resident inspectors observed the health physics technicians checking area radiation monitors, air samplers, and doing area surveys for radioactive contaminatio The NRC resident inspectors observed that when workers are required to enter areas where radiation exposure is probable or contamination possible the health physics technicians are present and available to provide i assistanc No violations or deviations were identified in the review of this program are . Monthly Security Observation 1 The NRC resident inspectors verified that there was a lead security officer (L50) on duty authorized by the facility security plan to direct

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security activities onsite for each shif The LSO did not have duties that would interfere with the direction of security activitie The NRC resident inspectors verified, rardemly and on the backshift, 'that the minimum number of armed guards required by the facility's security plan were present. Search equipment, including the X-ray machine, metal

. detector, and explosive detector, were operational or a.100 percent hands on search was being utilize The protected area barrier was surveyed by the NRC resident inspectors.-

The barrier was properly maintained and was not compromised by erosion, openings in the fence fabric, or walls, or proximity of vehicles, crate or other objects.that could be used to scale the barrier. The NRC resident inspectors observed the vital area barriers were well maintained and not compromised by obvious breaches or weaknesse The NRC resident inspectors observed that persons granted access to the site are badged indicating whether they had unescorted or escorted access authorizatio No violations or deviations were identified in the review of this program are . Exit Meeting An exit meeting was_ conducted on January 12, 1987, attended by those identified in paragraph 1. At that time, the NRC resident inspectors reviewed the scope and findings of the inspection.