IR 05000334/1986004

Insp Rept 50-334/86-04 on 860201-0310.No Violation Noted. Major Areas Inspected:Previous Insp Findings,Plant Operations,Housekeeping,Fire Protection,Radiological Controls & Physical Security

ML20210E302
Person / Time
Site:	Beaver Valley
Issue date:	03/19/1986
From:	Lester Tripp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20210E287	List: IR 05000334/1986004 ML20210E283
References
50-334-86-04, 50-334-86-4, NUDOCS 8603270311
Download: ML20210E302 (22)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /86-04 Docket N '

Licensee: Duquesne Light Company

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One Oxford Center 301 Grant Street Pittsburgh, PA 15279

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Facility Name: Beaver Valley Power Station, Unit 1 Location: Shippingport,-Pennsylvania Dates: February 1 - March 10, 1986

Inspectors: W. M. Troskoski, Senior Resident Inspector A. A. As rs, Resident Inspector Approved by: . . I uhh9 /M8$

3 L. E. TrApp, Chief, Reactor Projects Section 3A - ' Date Summary: Inspection No. 50-334/86-04 on February 1 - March 10, 1986 Areas Inspected: Routine inspections.by the resident insp'ectors (141 hours0.00163 days <br />0.0392 hours <br />2.331349e-4 weeks <br />5.36505e-5 months <br />) of licensee actions on previous inspection findings, plant operations, housekeeping, fire protection, radiological controls, physical security, engineered safety fea-tures verification, maintenance activities relating to the river water system, surveillance testing, control room annuciator survey and followup on special re-port Results: No violations were identified. Significant items reviewed included a

.. full power trip due to a malfunction in the steam. generator level control system

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(Detail 4.b.2), a RCS valve packing leak inside containment (Detail 4.b.1), second-ary system safety-related valve steam leaks (Detail 4.b.4), inadvertent partial actuation of the cable tray Cardox system (Detail 4.e), and a potential unreviewed-

, safety question concerning permissive P-10 of the RPS (Detail 9.b).

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TABLE OF CONTENTS Page Persons Contacted..................................................... 1 Plant Status......... . ............................................. 1 Followup on Outstanding Items......................................... 1 Plant 0perations...................................................... 4 Genera 1.......................................................... 4 Operations....................................................... 5 Plant Security / Physical Protection............................... 8 Radiation Controls............................................... 8 Plant Housekeeping and Fire Protection........................... 8 Engineered Safety Features (ESF) Veri fication. . . . . . . . . . . . . . . . . . . . . . . . . 9 Maintenance Activities............................................... 10 Surveillance Testing................................................. 11 Control Room Annunciator Survey...................................... 12 Followup on Special Reports.......................................... 13 1 Exit Interview....................................................... 14 Attachment 1 - Annunciator Survey

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DETAILS Persons Contacted J. J. Carey, Vice President, Nuclear Group R. J. Druga, Manager, Technical. Services T. D. Jones, General Manager, Nuclear Operations W. S. Lacey, Plant Manager J. D. Sieber, General Manager, Nuclear Services N. R. Tonet, General Manager, Nuclear Engr. & Constr. Unit The inspector also contacted other licensee employees and contractors during this inspectio . Plant Status The. plant operated at full power throughout the' inspection period with the exception of one trip (discussed.in detail 4 b) on February 10, 1986, and a four day power reduction to about 65% on February 14 - 18, 1986, to allow re-pair of the 8 main feedwater pump (Detail 4.b).

3. Followup On Outstanding Items The NRC Outstanding Items (0I) List was reviewed with cognizant licensee per-sonnel. Items selected by the inspector were subsequently reviewed through discussions with licensee personnel, documentation reviews and field inspec-tion to determine whether licensee actions specified in the OI's had been satisfactorily completed. The overall status of previously identified in-spection findings were reviewed, and planned and completed licensee actions were discussed for those items reported below:

(Closed) IFI (85-15-03): Review performance and documentation of radiological surveys at appropriate intervals by qualified technicians. This item was opened after it was determir.ed that rad technicians in training were conduct-ing routine surveys unsupervised. The routine survey program, Appendix 4 to the BVPS Radcon Manual, has since been revised to include requirements that the surveys be conducted by personnel qualified to ANSI N18.4 - 1971, or by technicians undergoing on-the-job training who are directly supervised by a qualified individual. Surveys conducted as part of on-the-job training will be so documented on the survey sheet. The inspector reviewed several monthly, weekly, and daily surveys to verify that they are being performed and docu-mented adequately. This item is close (Closed) IFI (85-22-07): Review dosimetry issue and site specific worker radiation training for visiting NRC inspectors. ~This item was previously addressed in Inspection Reports 334/85-21, 85-22 and 86-01. The licensee has developed and implemented adequate guidelines to aid in the timely access of NRC inspectors to radiologically controlled areas. The inspector reviewed the guidelines and observed the performance of the brief site specific radi-

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ation worker training and found no inadequacies. The inspector also verified that modified exposure history forms for NRC inspectors are readily available in the dosimetry booth. This item is close (0 pen) IFI (86-01-03): Update on replacement of the deformed river water system expansion joint and determination of the cause of deformation. The effluent header expansion joint deformation was identified on January 2,198 The licensee developed a temporary operating procedure should it fail and made plans to install an emergency patch until the expansion joint could be re-placed during the upcoming refueling outage scheduled for May, 1986. The emergency patch has not yet been installed. The patch design required speci-fic materials for which the licensee is having difficulty obtaining the re-quired quality assurance documentation. The inspector has observed no appre-ciable changes in the appearance of the expansion joint since the problem was initially identified. The inspector noted that resolution of this problem has been slo This item remains ope (Closed) Unresolved Item (85-16-02): Determine how the reactor vessel head vent high pressure alarm knife switch had been incorrectly identified. The vessel head vent alarm had been lit because of. leakage past a solenoid operated valve in centainment. A second in-line solenoid operated valve blocks further leakage to either the PRT or containment. The licensee had changed the alarm background from red to green signifying a normally lit con-dition. The licensee stated that the alarm color was changed because when

' the associated knife switch was pulled, the annunciator failed to clear as expecte At that time, it was believed that the switch was mislabeled. Re-view of the reactor vessel head vent system design revealed that the high pressure alarm circuit is designed to alarm when field contacts on the pres-sure instruments open. Therefore, opening the knife switch will not clear the alarm because this is interpreted as an open contact. To clear it, the knife switch must be opened and the annunciator side contact must be shorted with a jumper. The licensee verified that the knife switch was properly labeled and wired by demonstrating the process.to clear the alarm. This item is close (Closed) IFI (86-01-04): Observe data collection and review results in de-termination of motor driven auxiliary feedwater pump 3B operability. The in-spector observed performance of TOP 86-03,.FW-P-3B Operability Data Collec-tion, on February 12, 1986. The collection data was compared to the pump manufacturer's curve and the pump requirements of the FSAR. The data from FW-P-3B matched the manufacturer's curve and exceeded the FSAR miniiaum re-quirements of 350 gpm delivered to two steam generators against an extrapol-ated steam generator pressure of 1108 psi The inspector also noted that the licensee had obtained from Westinghouse a safety evaluation to assess the impact of reduced minimum auxiliary feedwater flow during the postulated loss of collant accident, loss of normal feedwater and feedwater pipe rupture ac-cidents. The LOCA was reanalyzed using reduced auxiliary feedwater flow from the motor driven pump of 310 gpm, and the feedwater accidents were reanalyzed using a flow of 325 gpm. These analyses determined that the conclusion of FSAR, Chapter 14, remain valid at the reduced flow rates. From the data i

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analysis of TOP 86-03, FW-P-3B was shown to be capable of delivering the re-quired 350 gpm flow to two steam generators against the required pressur This item is close (Closed) Unresolved Item (85-17-09): DLC to issue supplemental response to item 3.6 of IEB: 82-02. By letter dated February 13, 1986, the licensee provided a description of the types and compositions of fastener lubricants and injection sealant materials that have been used in the reactor coolant pressure boundary. This item is close (Closed) Violation (85-26-01): Failure to issue photo-identification badge to employee prior to granting access to the protected area. This item was discussed by representatives of the licensee with Region I management at an Enforcement Conference held on December 19, 1985. The licensee formally re-sponded to the violation in a letter dated February 27, 1986. The inspector reviewed the corrective actions taken for the card key control of the pro-tected area entry turnstiles and other actions taken to prevent reoccurrence as specified in the lette tory.and this. item is closed. Licensee actions were determined to be satisfac-(Closed) Unresolved Item (85-27-01): Update appropriate procedures to require that the containment airleck door is de-energized when not in~use. Station

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Administration Procedure No. 28, Reactor Containment Entries, was revised on l March 5, 1986, to include this requirement. This item is close '

(0 pen) Violation (85-25-04): Lack of emergency lighting for safe shutdown i

access and egress routes. An inspection of the licensee's compliance to Ap-pendix R requirements identified four plant areas in the various routes re-quired for safe shutdown that lacked a required 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> battery supply sourc The licensee's response to this violation dated February 11, 1986, indicated that a light would be installed in the Clean Shop Area and that an exemption request from the Section III.J requirements of 10 CFR 50, Appendix R, would be submitted requesting credit for the security perimeter lighting for the other area The inspector verified that the Clean Shop emergency light had been adde Though the use of the security diesel power light system appears acceptable in place of the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> battery power supply, this item remains open pending final NRR resolutio J (0 pen) TI (25-15-73): Inspection requirements for IEB: 85-03, MOV Common Mode Failures. This item has been opened to track future licensee actions in this are (Closed) Unresolved Item (84-30-02): Establish measures to control material and parts to prevent inadvertent use of uncontrolled items 'in safety related applications. This concern resulted from the identification of an untagged circuit breaker observed on the floor inside a safety related 125 Volt DC switchboard. After researching this concern, the licensee determined that the circuit breaker was included in the scope of work conducted under Design Change Package 55 This DCP replaced a spare 70 amp circuit breaker with a 25 amp breake The 70 amp breaker was left in the cabinet untagged after

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the work was complete. The' breaker has subsequently been placed in the storeroom for use on non-safety systems. To prevent recurrence, the licensee revised Procedure CDN 3.4, Section 17.0, Retirement and Reclamation of Mate--

rial, to assure that existing plant material maintains traceability ~after replacement. This item is close (Closed). Violation (85-22-04): Failure to issue 10 CFR 20.408 Termination Exposure Reports. The. inspector-reviewed the actions specified in DLC letter of December 16, 1985. The Secu'ity r Department provides a 45 day badge report compiled from the inactive and revoked badges printout once every two weeks to the Dosimetry Laboratory to allow a double check of personnel termination records issued by the first line supervision. In addition, the inspector re -

viewed radiation worker termination notifications issued since January,.1986, and verified compliance with Nuclear Group Directive No. 28. No discrepancies were identified and this item.is close . Plant Operations i . a, General Inspection tours of the plant areas listed below were conducted during both day and night shifts with respect to Technical Specification (TS)

compliance, housekeeping and cleanliness, fire protection, radiation control, physical security and plant protection, operational and main-tenance administrative control Control Roo Primary Auxiliary Building

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Turbine Building

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Service Building

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Main Intake Structure

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Purge Duct-Room

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East / West Cable Vaults

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Emergency Diesel Generator Rooms

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Containment Building

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Penetration Areas

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Safeguards Areas

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Various Switchgear Rooms / Cable Spreading Room

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Protected Areas

' Acceptance criteria for the above areas included the following:

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BVPS FSAR

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Technical Specifications (TS)

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BVPS Operating Manual (0M), Chapter 48, Conduct of Operations

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OM 1.48.5, Section D, Jumpers and Lifted Leads J, 1 --

OM 1.48.6, Clearance Procedures l --- OM 1.48.8, Records jl

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0M 1.48.9, Rules of Practice 1 l

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-0M Chapter 55A, Periodic Checks, Operating Surveillance Tests

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BVPS Maintenance Manual (MM), Chapter 1, Conduct of Maintenance

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BVPS Radcon Manual (RCM)

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10CFR50.54(k), Control Room Manning Requirements-

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BVPS Site / Station Administrative Procedures (SAP)

-- BVPS Physical Security Plan (PSP)

-- Inspector Judgement b. Operations

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j The inspector toured the Control Room regularly to verify compliance with NRC requirements and facility technical specifications-(TS). Direct ob-servations of instrumentation, recorder traces and control panels were made for items important to safety. Included in the reviews were the rod position indicators, nuclear instrumentation systems, radiation monitors, containment pressure and temperature parameters, onsite/offsite emergency power sources, availability of reactor protection systems and proper alignment of engineered safety feature systems. Where an abnormal condition existed (such as out of-service equipment), adherence to ap-propriate TS action statements was independently verified. Also, various operation logs and records, including com equipment clearance permits -in progress, pleted status surveillance board tests, maintenance and temporary operating procedures were reviewed on a sampling basis for compliance with technical specifications and those administrative con-

. trols listed in Paragraph 4 During the course of the inspection, discussions were conducted with operators concerning reasons for selected annunciators and knowledge of

I recent changes to procedures, facility configuration and plant condition '

The inspector verified adherence to approved procedure ~s for ongoing ac-tivities observed. Shift turnovers were witnessed and staffing require-ments confirme Except where noted below, inspector comments or ques-tions resulting from these daily reviews were acceptably resolved by licensee personne (1) A small primary reactor coolant system leak developed inside con-tainment on February 5, 1986. At about 4:30 a.m. , containment gaseous and particulate rad monitors (RM-215A and B) started to in-crease from a 400 cpm normal background level to about 4000 cp About an hour prior to that, control room personnel noted an in-creased makeup rate to the VCT. The licensee initiated OST 1.6.2,

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RCS Inventory Calculations, and noted that the containment sump pumpout rate had also increased from about 7 to 40 gallons per hou The initial RCS inventory calculation identified the magnitude of the leak as being approximately .9 gpm. A containment entry was made and operators visually identifieL the source as being one of

, the 3/4" high pressure isolation instrument valves for the loop bypass flow instrument, located next to the RTD manifold in the j C pump cubicle.

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Photographs taken during the containment entry identified the source of the leak as being from the valve packing. Discussions with lic-ensee personnel indicated that it was believed that the manual valve was off of its backseat and that manually repositioning the valve would terminate the leak. Initial radiation surveys of the area indicated that background dose rate was approximately 20 rad per hou Consequently, the licensee preplanned the next containment entry to use a long handle tool to turn the valve, thus minimizing personnel exposur A second containment entry was made and the leak terminated. The VCT makeup rate immediately returned back to normal. Inspector discussions with radcon personnel and operators who made the con-tainment entry indicated that the maximum exposure was about 385 mR, indicating that the ALARA considerations effectively minimized -

exposur Subsequent RCS inventory calculations indicated that the leak had been reduced to less than 0.1 gpm and the containment radiation monitors returned to normal background readings several hours'after the leak stopped. The inspector had no further concern (2) A steam generator (SG) high level turbine trip reactor trip oc-curred from full power at 7:27 p.m. on February 10, 1986. The high level (75%) in the A SG was caused by a malfunction in the main feedwater control system due to a blown 500 amp main power fuse in the No. 3 vital bus inverter. When the No. 3 vital bus power was lost, all three SG level control channels failed down scale and each feedwater reg valve control signal called for full open in re- !

sponse to the erroneous deman Control room operators immediately '

switched the level control systems from automatic to manual contro Review of the computer data indicated that'the operators were suc-cessful in turning the B and C SG 1evels around; however, the A SG feedwater reg valve failed to respond to the manual demand signal Feedwater flow-increased to about 140% of normal and quickly filled the SG to its high level trip setpoint. About eleven minutes into the transient, the vital bus was powered back up by its auxiliary power suppl )

ibis was the third trip related to the No. 3 vital bus since January, 1985. The first trip was an apparent breaker malfunction that oc-curred during performance of a maintenance surveillance test. The second trip occurred on October 4, 1985, and was due to the failure o-f the same 500 amp main power fuse (see Detail 3.b.1 of Inspection Report 334/85-22). Subsequent discussions with the I&C Supervisor indicated that licensee personnel found that the apparent cause of these fuse failures was due to a malfunction of the thyristor gating board due to overheating. Specifically,.a " thermal dot" indicated that the board had been subjected to a localized temperature of.150 F. The licensee replaced it with a spare, but maintained the N vital bus power supply from the auxiliary source.

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i lhe licensee subsequently brought a vendor representative on site to inspect the inverter (Cyberex). It was determined that several other circuit board cards also showed signs of slow,' prolonged heating. These cards were replaced with spares. Discussions with the I&C Supervisor indicated that all cards would be reinspected during the next refueling outage for indications of overheatin Determination of the cause of this condition is Unresolved Item (86-04-01).

After the trip, control room personnel noted that source range monitor NI-31 failed to' respond. Investigation revealed that the detector located in an instrument well adjacent to the reactor ves-sel, had failed. To replace this detector, intermediate range

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monitor NI-35 which is located in the lower half of the same in-strument well, had to be. removed. To allow the I&C technicians to do this, control room personnel removed the power and instrument fuses without bypassing the trip. Since the IRM reactor trip is a one out of two logic fail safe trip, the reactor protection system was inadvertently activated and the shutdown banks inserted. The licensee made the appropriate ENS notification. This work was per-formed u~nder the maintenance work request system without benefit of a corrective maintenance procedure as it was considered within the skills of the trade. Review of corrective actions to preclude recurrence is Unresolved Item (86-04-02).

(3) The reactor was restarted at about 2:40 p.m. on February 12, 1986, and full power was achieved without incident. Operations personnel noted that earlier vibration anomalies on the IB main feedwater pump had increased by about 0.9 mils on February 14, 1986. It was de-cided to remove the pump for maintenance and power was reduced to 65L The inspector observed the maneuver and identified no concern During maintenance activities on the 18 feedwater pump, the licensee found that a 5" x 1" metal stud was responsible for the vibratio This is the second time such a metal stud has been found in the IB pump. The. licensee is currently investigating the source of this foreign material and developing appropriate corrective actions to

'revent future introduction. Licensee action to limit the intro-duction of foreign material into the secondary system is Inspector Follow Item (86-04-03).

(4) During a tour of the plant on March 5, 1986, the inspector noted steam admission from the atmospheric exhaust of the turbine driven auxiliary feedwater pump. This indicated that one of the two parallel steam line isolation trip valves (TV-MS-105A, B) to the Terry Turbine was~ leaking b Discussions with operations and maintenance personnel indicated that the licensec intended to repair these valves (lap the seat) during the next refueling outage sched-uled for May, 1986. In the interim, condensate buildup in the tur-bine is being drained once per shift as necessary.

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During the same tour, the inspector also noted that several steam generator safety valve exhausts were emitting slight' whiffs of steam indicating leak-by and possible steam cutting of the safety' valve seats. This was also brought to the Maintenance Supervisor s at-tention. Followup to determine what corrective maintenance will be necessary during the next refueling outage is Inspector Follow Item (86-04-04). Plant Security / Physical Protection Implementation of the Physical Security Plan was observed in the areas listed in Paragraph 4a' above with regard to the following:

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Protected area barriers were not degrade'd;

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Isolation zones were clear; ,

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Persons and packages were checked prior to allowing entry into the Protected Area;

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Vehicles were properly searched and vehicle access to the Protected Area was in accordance with approved procedures;

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Security access controls to Vital Areas were being maintained and that persons in Vital Areas were properly authorized;

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Security posts were adequately staffed and equipped, security per-sonnel were alert and knowledgeable regarding position requirements, and that written procedures were available; and

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Adequate lighting was maintained.

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No discrepancies were observe Radiation Controls Radiation controls, including posting of radiation areas, the conditions of step-off pads, disposal of protective clothing, completion of Radi-ation Work Permits, compliance with the conditions of the Radiation Work Permits, personnel monitoring devices being worn, cleanliness of work areas, radiation control job coverage, area monitor operability (portable and permanent), area monitor calibration and personnel frisking proce-dures were observed on a sampling basi No deficiencies were observe e. . Plant Housekeeping and Fire Protection Plant housekeeping conditions including general cleanliness conditions and control of material to prevent fire hazards were observed in areas listed'in Paragraph'4a. Maintenance of. fire barriers, fire barrier pene-

trations, and verification of posted fire watchesLin these areas were also observed. 'The inspector identified no concerns. . , , . . . . _ - . - - . . _ _ - _ _

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A partial test for the cable tray mezzanine fire dampers was conducted per OST 1.33.13, Fire Protection System Detection Instrumentation Test, on March 6, 1986. It was conducted as a post-modification test for four fire dampers to ensure closure upon activation of the cardox blow out feature (See Unresolved Item 85-11-01). One of the four modified dampers (VS-D-85B) and one unmodified damper (VS-D-908) failed to clos Appro-priate maintenance work requests were initiated to repair the hardware and the one hour fire patrols continued. The unresolved item remains open pending damper repair and testin During the test, a procedure deficiency was identified. The test is conducted by manual isolation of the C02 line from its 10-ton source and pressing the local discharge button. This starts a six minute timer that controls the discharge sequence, which allows time for personnel to evacuate the local area and limit the amount of C02 discharged. After the puff test, the local alarm panel was reset; however, there is no reset for the timer, which was still calling for system actuation. When the local control panel was returned to normal, the discharge starte The panel was immediately de energized by the operator, limiting the discharge duration to only several seconds. There were no personnel injuries as access to the area was restricted by security personnel per the OS Discussions with the station personnel indicated that OST 1.33.13 would be revised to allow the timer to sequence out prior to resetting the system. Additionally, other fire protection tests would be reviewed and modified similarly. The inspector had no further concern . Enaineered Safety Features (ESF) Verification The operability of the Hydrogen Recombiners were verified during the week of March 3, 1986, by performing walkdowns of accessible portions that included the following as appropriate:

(a) System lineup procedures matched plant drawings and the as-built con-figuratio (b) Equipment conditions were observed for items which might degrade per-formanc Hangers and supports were operabl (c) The interior of breakers, electrical and instrumentation cabinets were inspected for debris, loose material, jumpers, et (d) Instrumentation was properly valved in and functioning; and had current calibration date (e) Valves were verified to be in the proper position with power availabl Valve locking mechanisms were checked, where require No deficiencies were identifie . .

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6. Maintenance Activities The inspectors observed portions of selected maintenance activities'on safety-related systems and-components to verify that those activities were being conducted in accordance with approved procedures, technical specifications and appropriate industrial codes and standards. The inspectors conducted record reviews and direct observations to determine that:

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Those activities did not violate a limiting condition for operatio Redundant components were operabl Required administrative approvals and tagouts had been-obtained prior to initiating wor Approved procedures were used or the activity was within the " skills of the trade."

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The work was performed by qualified personne The procedures used were adequate to control the activit Replacement parts and materials were properly certifie Radiological controls were properly implemented when necessar Ignition / fire prevention controls were appropriate for the activit OC hold points were established where required and observe Equipment was properly tasted before being returned to servic An independent verification was conducted to verify that the equipment was properly returned to servic The following activities were reviewed: The "B" River Water Subsystem was declared inoperable per Technical Specification 3.7.4.1 due to an inoperable river water pump discharge valve coincident with extensive maintenance being performed on the "B" river water pump. On February 6,1986, during routine performance of OST 1.30.6, Reactor Plant River Water Pump 1C Test, the motor operated discharge valve on pump 1C to the "B" header (MOV-RW-102C1) failed to automatically close as. required when the pump was secured. The valve electrically closed after the handwheel was turned slightly in the closed .

direction. The inspector observed maintenance performed per Corrective Maintenance Procedure 1.75.79, Limitorque Motor Operator Repair Mainten-ance. The valve open limit was discovered set 1too close to the manual stop. Therefore, when the valve opened, it jammed - turning the. hand-wheel had freed the valve and allowed it to close by the operation of

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the motor operator. The licensee reset the open limit to one-half turn more towards the closed direction to preclude any further jamming for this reason. The surveillance test was performed successfully and the river water' subsystem was declared operabl On March.5, 1986, M0V-RW-102Cl failed to stroke test from the control room. Subsequent' investigation determined that the line' starter coil was hanging up on a small plastic clip from the linestarter housin After removal, the valve was successfully stroked several times and returned to service the same da . On February 21 and 22, 1986, the licensee observed a relatively large increase in pressure differential across the A and B Reactor Plant Com-ponent Cooling Heat Exchanger tubes which contain river water. The lic-ensee attributed this pressure increase to silt accumulation in the heat exchanger tubes as a result of running Auxiliary River Water Pump 9A for a monthly. surveillance test on February 21, 1986. During recent high river water levels, silt, mud and debris had collected on the traveling screen in the Auxiliary Intake Structure. Both heat exchangers were cleaned and returned to servic The licensee discovered that the B charging pump speed increaser was leaking water out of the inboard and outboard shaft seals o~n February 24, 1986. Further investigation revealed that the speed increaser oil reservoir contained an oil - water mixture. River water to the speed -

increaser heat exchanger was isolated.and the leak stopped. The B charging pump was taken~out of service and the C pump replaced it as the standby charging pump. The licens'ee is currently replacing the speed increaser and its associated heat exchanger with spares. Investigation and determination of the cause of the lube oil heat exchanger failure is Inspector Follow Item (86-04-05). Two pinhole leaks were discovered in the river water discharge pipe be-tween the C component cooling water (CCR) heat exchanger and manual isolation valve RW-199. The licensee placed a temporary rubber patch clamped over the holes as a short-term fi The inspector noted that the side of the common RW effluent header that the C CCR heat exchanger effluent line discharges to visibly vibrates more than the other. sid Discussions with station personnel indicates that the licensee believes the cause of the pinholes to be due to a localized eddy caused by a throttled butterfly valve used 'for pressure control. Review of long term corrective actions is Inspector Follow Item (86-04-06). Surveillance Testing To ascertain that surveillance of safety-related systems or components is being conducted in accordance with license requirements, the inspector ob-served portions of selected tests to verify that: The surveillance test procedure conforms to technical specification re-quire'. ant . .

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12 Required administrative approvals and tagouts are obtained before initi-ating the tes ~ Testing is being accomplished by qualified personnel in accordance with an approved test procedure.

, Required test instrumentation-is calibrated.

LCOs are met.

3 The test data are accurat'e and complete. Selected test resu' lt- data was independently reviewed to verify accurac The test provides for independent verification of system restoration, j Test results meet technical specification requirements and test discre-pancies are rectified.

, The surveillance test was completed at the required frequenc The following in progress tests were witnessed by the: inspector:

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MSP 1.04, Reactor Trip Breaker Test, on February 24, 198 OST 1.30.3, Reactor Plant River Water Pump 1B Test, on March 7, 198 Revision 30 of MSP 1.04 was run for the first time when observed by the in-spector. The I&C technicians noted that there were several steps that ap-peared to be out.of sequence as a result of the' revision, After discussing the procedural errors with the I&C Engineer, it was decided to restore the system to normal alignment and correct the procedure. A field revision was

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subsequently issued and the MSP completed. Though the original p~rocedure re-view was inadequate, the I&C technicians were experienced,. recognized the deficiency and acted to correct it. The inspector had no further questions at this time.

I Control Room ~ Annunciator Survey The inspector performed a snap-shot survey of control room annunciators on February 18, 1986. Noted were safety and non-safety-related annunciators which were alarmed red, alarmed green and yellow-tagged out-of-service or disabled (See Attachment 1, Annunciator-Survey). Plant conditions during the j survey were as follows: 99% power, diluting to compensate for Xenon concen-

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tration increases, and power range nuclear instrumentation NI-N43 calibratio The inspector identified 30 annunciators alarmed red, 29 alarmed green, and 15 out of service. Annunciators which are alarmed green are normally lit during plant operation and have had~their background color changed to.facili-

- tate operator identification of abnormal alarms. . Operator response.to changes in green annunciator status' is identical to that for red annunciators. The inspector discussed wit.h licensee personnel the status of several alarms which

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were unnecessarily lit. The licensee is currently investigating alarmed and out of-service annunciator status and plans to make the necessary changes and repairs during the upcoming refueling cutage (May, 1986). This issue _is also being examined and tracked by NRR in the control room, design review as a human factors concer . Followup on Special Reports NRC Vendor Branch Inspection Report 99901033/85-01, issued December 27, 1985, concerned an inspection conducted at Power Inspection, Inc. This inspection report noted significant vendor QA program failures related to services provided to various utilities including Duquesne Light Com-pany. This specifically concerned certification of personnel for non-destructive examination, and calibration of testing equipment used on ET of control room air conditioning condensers, component cooling water heat exchangers in the reactor and turbine plant, and other heat ex-changers (diesel generator, recirculation spray, and blowdown). The inspector brought this to the attention of licensee personnel and re-quested that they evaluate the significance of these deficiencies upon the services provided to the statio A licensee representative stated that PII had been placed on the QA Qualified Suppliers Hold List for Beaver Valley, Unit 1. Additionally, a review would be conducted to sort those eddy current examinations into two groups: (1) those performed in support of maintenance, and (2) those performed for ISI baseline data. No further action is planned for the first group, but a sample reinspection will be performed for the secon Review of licensee action to validate any ISI baseline data obtained by PII is Unresolved Item (86-04-07). Westinghouse notified the NRC of a potential unreviewed safety question on February 26, 1986, applicable to all reactor protection logic systems designed by them. The potential malfunction involved the P-10 permissive which is used to enable (3 out of 4 logic below 10%) and block (2 out of 4 logic above 10%) the low power flux trips'provided by the source range, intermediate range.and power range monitors. If one of the four power range monitors becomes inoperable while at power, technical speci-fications require tripping all bistables, including the P-10 permissiv Should power subsequently be reduced below 10%, and assuming a single failure of any one of the three remaining P-10 bistables, the low power flux trip would not be automatically reinstituted. This is contrary to the assumptions used in the boron dilution and uncontrolled bank with-drawal from subcritical analysis of the FSAR and reduces the margin of safety as defined in the basis of the technical specificatio The Plant Manager informed the inspector that Westinghouse had informed the station of this concern. The shift supervisor turnover logs added a caution with regard to any power reduction below 10% as a short term

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I action. Long term resolution will be tracked as Unresolved Item (86-04-08).

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10. Exit Interview Meetings were-held with senior facility management periodically during the-course of this inspection to discuss the inspection scope and findings. A summary of inspection findings was further discussed with the licensee at the conclusion of the report perio i l

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ATTACHMENT 1 ANNUNCIATOR SURVEY

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PANEL ANNUNCIATOR ST5TUS DISCUSSION ,

ANN.A1/12 VITAL BUS III A/R 2-10-86 BLOWN CONTROL FUSE ON-TROUBLE BUS INVERTER, CURRENTLY ALIGNED TO ALTERNATE FOR INVESTIGATION

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OF FAILUR , ANN.1A/91 RIVER WATER PUMP A/R~ RW PUMP 1B MAINTENANCE 18 SEAL WATER PRESS. LOW

ANN.A2/38 STEAM GENERATOR A/R HOLDING TANK FOR-LIQUID DRAIN TANK LEVEL LOW WASTE - NORMALLY EMPT ANN.A2/51 LIQUID WASTE A/R LW EVAPORATOR ON CLEARANCE EVAPORATOR LEVEL HIGH i ANN.A2/53 LIQUID WASTE EVA A/R LW EVAPORATOR ON CLEARANCE BOTTOMS PUMP THERMAL OVERLOA ANN.A2/54 LIQUID WASTE LOW

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A/R NORMAL, TANK IS CURRENTLY LEVEL DRAIN TANK EMPT A LEVEL LO ANN.A2/56 LIQUID WASTE HIGH A/R LW EVAP. ON CLEARANCE j LEVEL DRAIN TANK 2A LEVEL HIGH.

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- ANN.A2/57 LIQUID WASTE EVAP' A/R NORMAL, TANK IS CURRENTLY i TEST TANK SA LEVEL EMPTY.

LOW.

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ANN.A2/61 CONTAMINATED SHOWER A/R NORMAL, TANK IS CURRENTLY-AND LAUNDRY DRAIN EMPT TANK 6A LOW LEVEL.

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ANN.A2/62 LIQUID WASTE LOW A/R NORMAL, TANK IS CURRENTLY

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LEVEL DRAIN TANK EMPT LOW LEVEL.

ANN.A2/64 LIQUID WASTE HIGH A/R NORMAL, TANK IS CURRENTLY F

LEVEL DRAIN TANK FULL BECAUSE LW EVAPORATOR j 2B' LEVEL HIG ON CLEARANCE.

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Attachment 1 2 PANEL ANNUNCIATOR STATUS DISCUSSION ANN.A2/81 AUX. B0ILER LOCAL A/R_ NORMAL, B0ILER CURRENTLY IN WET PANEL TROUBL LAYUP, ALARM HAD NOT YET BEEN RESET.

4 ' ANN.A2/125 BORON EVAP.-TEST A/R NORMAL, TANK IS CURRENTLY TANK 2B LEVEL LO EMPTY.

ANN.A3/4 BORON EVAP. BOTTOMS A/R NORMAL, TANK'IS CURRENTLY HOLD TANK LEVEL LOW EMPTY.

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ANN.A3/11 BORON EVA A/R B EVAPORATOR NOT IN SERVICE, DISTILLATE ACCUMULATOR ACCUMULATOR EMPTY.

j 1B LEVEL LO ,

ANN.A3/16 BORON EVAP. BOTTOMS A/R NORMAL, FOR PREVENTION OF j COOLER DISCHARGE BORON SOLIDIFICATION TEMP. HIG ANN.A4/49 LOOP OVERPOWER A/R- BISTABLES ARE CURRENTLY

, DELTA T TRIPPED FOR MSP ON NI-N43

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ANN.A4/50 LOOP OVERPOWER A/R BISTABLES CURRENTLY TRIPPED-

, DELTA T AUTO TURBINE FOR NI-N43 MS , RUNBACK BLOCKED R0D 1 WITHDRAWAL.

] ANN.A4/53 LOOP OVERTEM A/R BISTABLES CURRENTLY TRIPPED DELTA T

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FOR NI-N43 MSP.

j ANN.A4/54 LOOP OVERTEM A/R BISTABLES CURRENTLY TRIPPED DELTA T AUTO TURBINE FOR NI-N43 MS RUNBACK BLOCK R0D WITHDRAWAL.

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-ANN.A6/6 P0OL PURIFICATION A/R PUMP SHUT 00WN FOR FUEL

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-PUMP 4A DISCHARGE P00L WORK ACTIVITIE PRESSURE LO ANN.A6/39 CHEMICAL ADDITION ~ A/R CHANNEL 00S 2-11-86 MWR

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TANK TEMP. LOW 860212, FAILED LO CHANNEL I.

T l ANN.A6/66 COOLING TOWER A/R LOW LEVEL IN GLYCOL-TANK -

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DE-ICING SYSTEM INDICATOR TROUBL CURRENTLY t

TROUBL UNDER INVESTIGATIO ANN.A6/93 SCREENWASH PUMP A/R PUMP OOS, RETIRED IN PLACE.

i TROUBLE.

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Attachment 1 3 i PANEL ANNUNCIATOR STATUS DISCUSSION ANN.A6/102 INTAKE STRUCTURE A/R RIVER LEVEL CURRENTLY HIG RIVER WATER LEVEL TEMP. ABNORMA ANN.A7/71 STEAM GENERATOR A/R SYSTEM NOT INSERVICE - NEW BLOWDOWN FLASH TANK BLOWDOWN SYSTEM INSTALLE #1 LEVEL LO ANN.A9/128 FAULT RECORDER A/R CURRENTLY 005 UNABLE.T0 LOCATE TROUBL PROBLE ANN.A11/104 METER 0ROGICAL A/R UNDER INVESTIGATION - ALARMS BUILDING TROUBLE WHEN LIGHTS ARE TURNED O ANN.A11/125 HYDR 0 PNEUMATIC A/R FIRE PROTECTION SYSTEM COMPRESSOR TANK AIR COMPRESSOR -

OVERHAUL AFTER FAILUR THERMAL OVERLOA ANN.A11/127 FIRE PROTECTION A/R FAILED POWER SUPPLY - CURRENTLY SYSTEM TROUBLE - UNDER REPAI WAREHOUS ANN.A1/29 SI PUMP 1B A/R 1-15-86 MWR'860093 SEAL ACCUMU-SEAL WATER YELLOW TAGGED LATOR LEVEL SWITCH LOOSE WIRE LEVEL LO ANN.A1/42 CONTAINMENT INS DISABLED 5-16-85 MWR 830942 RECEIVED PIT LEVEL HIG GROUND ON ANNUNCIATO ANN.A1/35 CONTAINMENT AIR DISABLED 9-25-84 DCP 612 PARTIAL PRESS HIGH-LOW CHANNEL ANN.A1/36 CONTAINMENT AIR DISABLED 3-16-84 DCP 612 PARTIAL PRESS HIGH-HIGH CHANNEL ANN.A2/4 EVAPORATOR CIR DISABLED 8-16-84 MWR 841941 - LOCAL FLOW WATER PUMP SEAL GAUGE READS ZER0 AND LOW FLOW WATER FLOW LO ALARM IN, ADEQUATE SEAL FLOW PRESENT (ENSING LINE MAY BE PLUGGE ANN.A2/122 B0RON EVAP. CIR DISABLED 6-24-85 MWR 821467 0FF SCALE PUMP 68 DISCHARGE LOW, NO TRANSMITTER, PRESS. HIGH-LO _

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Attachment 1 4 PANEL ANNUNCIATOR STATUS DISCUSSION ANN.A3/31 BORIC ACID DISABLED 2-18-81 MWR 826531 BORONMETER CONCENTRATION HIGH- WILL BE RETIRED IN PLAC LO ,

ANN.A3/118 RCP 1C SEAL DISABLED 5-19-85 MWR 837701 INSTRUMENT INJECTION BYPASS TO CALIBRATION, NORMALLY ALARMED /

VCT FLOW LO GREE ANN.A4/14 PZR PORV LOW PRESS DISABLED 1-5-85 MWR 851286 GROUND IN RELIEF PROTECTION SEQUENCE OF EVENTS RECORDER INOPERABL BAY ANN.A4/76 COMPUTER ALARM R00 DISABLED 11-19-80 NOTHING CAN BE DONE DEVIATION / SEQUENCE SHORT OF REDESIGNING RPI AND NIS POWER RANGE TILT REPROGRAMMING P250 COMPUTE ALARM MAY BE RETIRE ANN.A6/49 STEAM GENERATOR DISABLED 1-8-86 CLEARANCE #49410 BLOWDOWN DRAIN STRAINER DIFFERENTIAL PRESSURE HIG ANN.A6/96 AUX. CHLORINATION DISABLED 6-30-85 OPEN KNIFE SWITCH ON CHEMICAL FEED LOCAL ANNUNCIATOR - NUISANCE ALAR PANEL TROUBL ANN.A7/95 TURNING GEAR SHAFT A/R 10-9-85 SPEED INDICATOR SIGNAL AT ZER0 SPEED YELLOW ORIGINATED IN TURBINE SUPER-TAGGED VISORY, WHICH IS ON CLEARANCE, AWAITING DCP 615 MODIFICATIO ANN.A7/112 TURBINE SUPERVISORY A/R 1-28-85 NEW VIBRATION MONITORING INSTRUMENT POWER ' YELLOW SYSTEM IN SERVICE - OLD SYSTEM 0F TAGGED REMOVE ANNUNCIATOR WILL BE MODIFIE ANN.A11/52 CONTROL ROOM DISABLED 10-18-85 CLEARANCE #514985 EMERGENCY AIR INSTALLING NEW COMPRESSO COMPRESSOR TROUBL ANN.A1/69 BIT TEMP. HIGH A/G NORMAL - PREVIOUS HIGH BORON CONCENTRATION REQUIRED HIGH TEMPERATURE ANN.A1/100 OUTSIDE RECIR A/G NORMA SPRAY PUMP 21 SEAL-WATER LEVEL HIG . - - . --

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Attachment 1 5

PANE ANNUNCIATOR STATUS DISCUSSION I

ANN.A1/124 0UTSIDE RECIR A/G NORMA SPRAY PUMP 2B SEAL WATER LEVEL HIGH.

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ANN.A1/12 RESIDUAL HEAT A/G NORMAL.

, REMOVAL SYSTEM DISCHARGE FLOW LOW.

ANN.A2/18 H2 RECOMBINER A/G NORMAL - DE-ENERGIZED DURING LOCAL PANEL A TROUBL NORMAL OPERATIO ANN.A2/19 H2 ANALYZER LOCAL A/G NORMAL - DE-ENERGIZED.DURING PANEL 1 TROUBL NORMAL OPERATIO ANN.A2/22 H2 REC 0MBINER A/G NORMAL - DE-ENERGIZED DURING

'rF c PANEL B TROUBL NORMAL OPERATIO ANN.A2/23 H2 ANALYZER LOCAL A/G NORMAL - DE-ENERGIZED DURING

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PANEL B TROUBL NORMAL OPERATIO . ANN.A2/31 N2 SUPPLY HEADER A/G NORMAL - N2 NOT NORMALLY ALIGNED

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PRESS. LO TO HEADER UNLESS IN US ANN.A2/115 BORON EVAP. CIR A/G NORMAL ~- EXCESSIVE SEALWATER PUMP 6A SEAL. WATER DILUTES EVAPORATOR.

FLOW LO ANN.A2/123 BORON EVAP. CIR A/G NORMAL - EXCESSIVE SEAL WATER PUMP 6B SEAL WATER DILUTES EVAPORATO i FLOW LOW.

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ANN.A3/60 LOOP FILL HEADER A/G NORMAL - SETP0 INT 2300 PSIG, PRESS. HIG CHARGIMG PUMP DISCHARGE PRES PSIG, FILL HEADER NOT NORMALLY IN USE.

J ANN.A3/102 RCP 1A SEALING A/G NORMAL - RETIRED IN PLAC . BYPASS TO VCT FLOW LO ANN.A3/110 RCP 1B SEAL IN A/G NORMAL - RETIRED IN PLAC BfPASS TO VCT FLOW LO ANN.A4/7 PZR PORV N2 SUPPLY A/G NORMAL - ADMIN. PROCEDURE '

PRESS. LO REQUIRES LOW PRESSURE, PORV'S iiAVE N2 ACCUMULATOR I

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PRESSURE HIGH.

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Attachment 1 6 PANEL ANNUNCIATOR STATUSI DISCUSSION ANN.A4/33 PZR BACKUP HEATER A/G NORMAL, CURRENTLY DILUTING -

GROUP ON MANUA PROVIDES FOR MIXING IN PZ ANN.A4/82 NIS SOURCE RANGE A/G NORMAL - SOURCE RANGE BLOCKED LHI SHUTDOWN FLUX WHEN AT POWE ALARM BLOCKE ANN.A4/85 NIS SOURCE RANGE A/G NORMAL - DETECTOR DE-ENERGIZED LOSS OF CHANflEL I WHEN AT POWE DETECTOR VOLTAG ANN.A4/87 NIS 50bdCE RANGE A/G NORMAL - DETECTOR DE-ENERGIZED LOSS OF CHANNEL II WHEN AT POWE i DETECTOR VOLTAG ANN.A6/10 FUEL POOL PUMP 1B A/G -NORMAL - USUALLY RUN 1A PUMP DISCHARGE PRESS. LOW ONL ANN.A6/14 POOL PURIFICATION A/G NORMAL - USUALLY RUN 4A PUMP PUMP 4B DISCHARGE ONL PRESS. LO ANN.A6/16 REFUELING CAVITY 'A/G NORMAL - DRAINED DURING POWER LEVEL LO OPERATIO ANN.A6/38 CCR HX 8" DISCHARGE A/G NORMAL - LOW LGiO.0N CCR HEADE LINE LOW FLO ANN.A6/87 H2 BLANKETING A/G SYSTEM USUALLY NOT IN SERVICE SYSTEM PRESS. LOW AT POWER OPERATIO ANN.A8/40 BULK H2 STORAGE A/G NORMAL - H2 ISOLATED TO MAIN TANKS PRESS. LOW GENERATO ANN.A9/100 125V DC BATTERY A/G NORMAL - ALARM SETP0 INT HIGHER CHARGER 1 FAILUR THAN CURRENT CHARGER OUTPU ANN.A9.108 125V DC BATTERY A/G NORMAL -' ALARM SETPOINT HIGHER CHARGER 3 FAILUR THAN CURRENT CHARGER OUTPU ANN.A9/112 125V DC BATTERY A/G NORMAL - ALARM SETPOINT HIGHER CHARGER 4 FAILUR THAN CURRENT CHARGER OUTPU KEY: ,

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A/R - ANNUNCIATOR ALARMED RED A/G - ANNUNCIATOR ALARMED GREEN

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