IR 05000334/1982026
| ML20028D718 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 12/21/1982 |
| From: | Lazarus W, Lester Tripp, Troskoski W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20028D701 | List: |
| References | |
| 50-334-82-26, IEB-79-14, IEB-79-21, NUDOCS 8301190374 | |
| Download: ML20028D718 (14) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No. 50-334/82-26 Docket No. 50-334 License No. DPR-66 Priority
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Category C
Licensee: Duquesne Light Company 435 Sixth Avenue Pittsburgh, Pennsylvania Facility Name: Beaver Valley Power Station, Unit 1 Inspection At: Shippingport, Pennsylvania Inspection Cond ted: October 26 - November 30, 1982 Inspectors:
. b. subl0 l 8I ?
t y. M. TrHkoski, Resident Inspector date
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W.gtazaru p Project Engineer date
/ 2 ?k Approved by:
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'L. E. THpp, Chief, Reactor Projects date Section No. 2A, Projects Branch 2 Inspection Summary:
Inspection on October 26 - November 30, 1982 (Inspection No. 50-334/82-26)
Areas Inspected: Routine inspections by the resident inspector (89.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />)
and a Region-based inspector (24.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />) of: licensee action on previous inspection findings, plant operations, followup on IE Bulletins, housekeeping, fire protection, radiological controls, physical security, radwaste opera-tions, surveillance testing, maintenance activities, in-office and onsite licensee event report review.
Results: No violations were identified.
8301190374 830105 gDRADOCK05000
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DETAILS 1.
Persons Contacted F. Bissert, Manager, Nuclear Support Services J. Carey, Vice President, Nuclear Division H. Caldwell, Assistant Station Superintendent M. Coppula, Superintendent of Technical Services K. Grada, Superintendent of Licensing and Compliance R. Hansen, Maintenance Supervisor J. Indovina, I&C Supervisor T. Jones, Manager, Nuclear Operations J. Kosmal, Radiological Operations Coordinator W. Lacey, Station Superintendent V. Linnenbom, Radiochemist J. Lukehart, Security Director L. Schad, Operations Supervisor E. Schnell, Radcon Supervisor J. Sieber, Manager, Nuclear Safety and Licensing R. Swiderski, Superintendent of Nuclear Construction N. Tonet, Manager, Nuclear Engineering J. Wenkhous, Reactor Control Chemist T. Zyra, Plant Performance and Testing Supervisr'
The inspector also contacted other licensee empl)yees and contractors during this inspection.
2.
Licensee Action on Previously Identified Inspection Findings The NRC Outstanding Items (OI) List was reviewed with cognizant licensee personnel.
Items selected by the inspectors were subsequently reviewed through discussions with licensee personnel, documentation review, and field inspection to determine whether licensee actions specified in the OIs had been satisfactorily completed. The overall status of previously identified inspection findings was reviewed, and planned anc completed licensee actions were discussed for those items reported below.
(Closed) Violation (79-13-01):
Failure to perform adequate design control for circuit modification on diesel genera' ors. This violation involved a teaporarily installed test circuit (used for troubleshooting)
that did not receive the detailed evaluation a circuit is required to undergo when interfacing with safety related equipment.
In the licens-ee's response of September 12, 1979, it was stated that this circuit was not a permanent change and as such, did not require an in-depth review.
Since then, QA Procedure OP-10, Maintenance and Modification Planning, Revision 4, has been established to assure that activities that may affect the function of safety-related components are performed in a manner at least equivalent to that specified in the applicable codes, standards, design requirements, and material specifications. The inspec-tor had no further questions on this ite _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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(Closed) Unresolved Item (79-24-03):
Review action taken for lighted and out-of-service annunciators. This issue was brought to the attention of DLC management during the last Systematic Assessment of Licensee Per-formance (SALP) meeting (No. 50-334/82-07) on March 25, 1982. The licensee's response to this item, contained in a letter to the NRC dated April 16, 1982, is currently being tracked as outstanding item 82-07-03.
(Closed) Unresolved Item (79-24-05):
Review implementation of fire barrier integrity program and revisions to associated surveillance proce-dures. This issue was last updated in NRC Inspection Report No.
50-334/79-27, and left open pending the incorporation of a specific list of fire barrier doors into logs and surveillance procedures for periodic verification of integrity. OST 1.33.9, CO Fire Protection System
Inspection Test, has since been issued and requires an operability check of all fire doors for both safety related and non-safety related areas on a weekly basis to comply with technical specification requirements.
Additionally, Station Logs contained in the BVPS OM 1.54.3, were revised with a similar list that checks these doors each shift.
Various security procedures also include this list as part of the routine security pa-trols.
(Closed) Violation (79-29-01):
Failure to measure and record hydraulic fluid temperature as required by Test Procedure 1-75-159.
In a response dated January 17, 1980, the licensee stated that the fluid temperature data was not recorded during the test because it was believed that in doing so, the as found condition of the snubbers would be disturbed.
Data from OST's which recorded the ambient temperature in the Auxiliary Building were reviewed for the test period to verify that the tests were conducted in the acceptable temperature range of 72 to 85 F, assuming the hydraulic fluid was in thermal equilibrium with the testing environment.
Test procedure CMP-1-75-159, Operating and Maintenance for ITT Grinnell Snubber Tester, was subsequently revised to measure the fluid temperature by placing a calibrated thermometer in one of the bleed ports. This method does not alter the as found condition of the snubber and satis-factory resolves the issue.
(Closed) Unresolved Item (79-29-02):
Licenset to evaluate temperature compensation applicability to lock-up and bleed rate acceptance criteria in Test Procedure 1-75-159.
The evaluation was performed and a set of curves relating the bleedrate and lock-up acceptance criteria to tempera-ture were developed. CMP-1-75-159, Operating and Maintenance for ITT Grinnell Snubbers, was subsequently revised to include those curves.
(Closed) Unresolved Item (80-26-01):
Confirm proper reporting of clai-ding failure in 1A charging pump casing.
Because this event was caused by multiple personnel errors, there is no 10 CFR Part 21 reporting concern. This item is closed.
(Closed) Violation (80-27-02):
Failure to maintain drawings for remote shutdown panel pressurizer level indicator. The inspector reviewed the licensee's response of June 25, 1981, and verified that the appropriate
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Westinghouse Electric Corporation vendor drawings incorporated in MSP 6.42, L-460 Pressurizer Level Protection Channel II Calibration, were revised to include the remote shutdown panel indicator, LI-RC-460A.
In this response, the licensee stated that the MSPs which calibrate instru-ments on the shutdown panel were being revised to correct loop diagram inconsistencies, and that any changes deemed necessary would be incor-porated in the referenced drawings prior to use. The inspector discussed this with personnel from the procedures group and was informed that as of October, 1982, all loop diagrams were being deleted from the MSPs prior to their next issue. The procedures will now reference the print number so that technicians can refer to one of the controlled copies for refer-ence. This ensures that current as-built documents are used for safety related work. The inspector had no further questions.
(Closed) Unresolved Item (80-27-05): Review corrective and preventive actions for foreign material found in QS pump. The Nuclear Services Quality Control Department headed a special task force for this inves-tigation and preser.ted their findings to DLC management in a report dated November 5, 1980. The scope of the investigation involved a detailed appraisal of Construction Division Nuclear (CDN) activities concerning various Design Change Packages and individual Modification Work Packages performed during the outage, plus a review of Station Maintenance and Operations activities. The report identified the decontamination work on Quench Spray Tank-1 performed early in the outage, as the probable source of debris. The Low Head Safety Injection pump suction lines were veri-fied to be clean and various pump data were trended during surveillance tests to assure system operability.
The Task Forces evaluation could not determine exactly when the debris was introduced into the safety system. However, it did determine that prior to May 8, 1980, there was a repetitive problem experienced with CDN's housekeeping controls. This resulted in the Schneider QC group being assigned responsibility for enforcing housekeeping controls; including the maintenance of Clean Zones under QC control; keeping of detailed tool and Material Control Logs; and, final work cleanliness verification. Because of the effectiveness of this new program, the Task Force felt that the debris was probably introduced at some time before May 8, 1980. All safety systems were verified to be operable and no other similar problem has been observed since.
(Closed) Unresolved Item (80-27-08):
Review DLC action for misoriented check valves (LER 80-68). The licensee committed to review and revise all procedures relating to installation of non-welded CAT I check valves to require specific verification of valve orientation vs. system flow direction.
Internal licensee memorandum, BVPS:RLH:51 dated October 27, 1980, and BVPS:MC:3, dated January 15, 1981 document this review of the plant's equipment list and identification of the applicable check valve.
Corrective Maintenance Procedure 1-75-186, Repair of Mission Check Valves, has since been revised to require that both a mechanic and an operator verify correct valve orientation. This completes the licensee's corrective actio __.
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(Closed) Unresolved Item (80-27-11):
Review DLC preventive actions for Reactor Coolant System (RCS) dilution (LER 80-70). The two valve align-ments that resulted in the RCS dilution event were established by OST 1.11.10, Boron Injection Flow Path Power Operated Valve Exercise.
Each alignment (CH-135 or MOV-CH-350 open) establishes a flow path whereby a portion of the boric acid flow is diverted from the blender. When this occurs and the blender controls are being operated in "AUT0", the flow difference is made up with primary water, initiating RCS dilution.
Accordingly, a precautioning note and a step requiring the posting of a caution tag on the blender controls were added to the OST. Whenever CH-135 or MOV-CH-350 is opened per OST 1.11.10, the blender control is placed in manual. This item is closed.
(Closed) Unresolved Item 80-27-12): DLC to issue followup report con-taining the engineering evaluation for Velan C58 check valve binding.
(LER 80-74). The inspector reviewed the update of LER 80-74 issued on September 8, 1982.
Interference between the valve disc and swing arm prevented the valves from properly seating. This interference was caused by a vendor design change to the valve disc that would also require a minor modification to the swing arm for the valve applications at BVPS.
Instructions to this efi'ect were not supplied by the vendor when replace-ment discs were ordered by the Maintenance Department. The licensee has since obtained updated drawings from the vendor and reworked all of the affected valves per a revised Corrective Maintenance Procedure..Addi-tionally, all other C58 check valves have had the old discs replaced with the new discs per Design Change Requests 422 and 423 performed during the second refueling outage in 1982.
(Closed) Unresolved Item (80-27-14):
Review DLC actions for Rod Position Indication System anomalies. The inability to calibrate the Analog Rod Position Indicator channels to the 12 step accuracy required by Techni-cal Specifications 3.1.3.1 and 3.1.3.2 was extensively discussed in NRC Inspection Reports 50-334/80-27, 80-30, and 81-04. Technical Specifica-
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tions (Amendment 35) were subsequently developed and approved which recognized many of the RPI characteristics that were studied during the Cycle 2 startup testing period (discussed in the above inspection re-ports). A permanent long term fix was to be developed later.
In the interim, it was recognized that a large amount of operator attention was being directed toward the RPI system, detracting from other overall duties. Consequently, DLC contracted with Westinghouse Electric Corpo-ration to perform a reanalysis of the rod misalignment situation. This was done using a 116 step misalignment as compared to the 12 step misalignment used in the Standard Technical Specifications and was submitted as TS Amendment 42 which was approved for use in Cycle 2 only.
As a fix for Cycle 3, extensive technical discussions were conducted between DLC, Westing b use Electric Corporation, and the NRC, including the Core Performance Granch and the Operating Reactor Assessment Branch of NRR. Amendment 51 was subsequently issued which modified the rod misalignment, core pecking factor limits and associated parameters while Amendment 57 changed the RPI requirements for modes 3,4 and 5.
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mance with these revised Technical Specification requirements is reviewed by the inspector as part of the routine program.
(Closed) Inspector Follow Item (80-27-24): Auxiliary Feedwater (AFW)
system procedures not consistent with Primary Demineralized Water Storage Tank (PDWST) alarm configuration. The licensee completed Design Change Package No. 304 which installed a low-low level alarm on the PDWST. This alarm configuration is now consistent with Emergency Operating Proce-dures, E-1, E-2 and E-3, Revision 10, which require operators to transfer the AFW system to an alternate source of water on a PDWST low-low level alarm.
(Closed) Unresolved Item (80-27-30): Acceptability of DLC IEB:79-14 Actions per Immediate Action' Letter 80-43.
DLC committed to submit a final report on the status of the equipment nozzle load evaluations by August 14, 1981. This report stated that all IEB:79-14 nozzle eval-uations had been completed and required modifications completed except for two items. The first item, control room air conditioning evaporator coils, could not have the load carrying capacity of the nozzle con-nections quantified by the vendor.
Because the licensee determined that the calculated loads and seismic contributor are small, it was not categorized as a safety problem and no modifications were performed.
For the redundant river water control room air conditioning coils, piping modifications were installed per Design Change Package 305. The Safety Evaluation of DCP 305 was reviewed and approved by the Onsite Safety Committee during meeting BV-OSC-64-82.
(Closed) Unresolved Item (81-08-08): Procedure for monitoring incore thermocouples does not include provisions for operators to acknowledge the computer output.
Revision 34 to Log L5-17, " Operation With Reactor Coolant System (RCS) Partially Drained," requires that operators log, in the Reactor Operators' log, core thermocouple readings every ten minutes if Residual Heal Removal (RHR) flow is lost, or hourly anytime RCS level, RHR flow, or RHR pump electrical current readings are unstab'ie. The inspector had no further questions in this area.
(Closed) Unresolved Item (81-25-05):
Implement program for Technical Specification (TS) implementation verification. The licensee issued Licensing and Compliance Proceduie LCP-9 to review and confirm implemen-tation of TS amendments. The procedure includes controls to notify affected departments of the proposed TS change request, to verify that applicable procedures are reviewed end revised to reflect the changes set forth in the TS amendment, and to at, sign the Senior Licensing Engineer responsibility of supervising the review and approval of TS changes, including the adequacy of those procedures written to comply with the TS amendment.
(Closed) Unresolved Item (81-28-10):
DLC to submit followup to LER 81-91 identifying the cause of the fire main rupture. The licensee issued a supplemental report on October 11, 1982, that contained the results of an engineering analysis performed by Industrial Testing Laboratory. The
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analysis identified external finger corrosion (initiated by the pipe bedding material chemistry) coupled with a slight bending moment as the cause of failure. Due to the extent of corrosion, the bending moment applied was much less than normally expected to cause a fracture.
Because this appeared in an isolated section of the fire main that was built to ANSI 21.6 and 21.8, the licensee determined that only the failed section warranted replacement. The inspector had no further questions at this time.
(Closed) Unresolved Item (81-02-06): On June 28, 1982 the licensee submitted an update to LER 81-06 documenting the actions taken to prevent recurrence of equipment inoperability due to loss of heat tracing.
It was noted that no heat trace related loss of capability occurred in the Winter of 1981-82. The inspector had no further questions in this area.
(Closed) Unresolved Item (81-18-09): The inspector verified that Station Administrative Procedure. Chapter 10, Onsite Committee, Rev. 2, had been revised to clarify identification of OSC members and alternate with requirements for documentation of activities. The inspector had no further questions in this area.
(Closed) Noncompliance (81-?0-01):
Failure to notify NRC of automatic
. actor trip in accordance with 10 CFR 50.72.
In response to this iol nion (DLC letter dated 11/4/81), the licensee stated that tr.aining in the reporting requirements of 10 CFR 50.72 would be included in the Licensee Operator Retraining Program. The inspector verified that this training had been completed by November 25, 1981.
(Closed) Followup Item (82-06-02): The inspector reviewed PMP 1-60CR-27-1ME, " Movable Platform and Hoists Inspection and Test," Rev. 4, to verify that changes had been made to clarify that MP or ultrasonic testing of crane books is required only if tram measurements (hook distortion) differed from original measurements.
(Closed) Followup Item (81-18-05): The cause of the battery charger output breaker trips was determined to be a faulty circuit breaker, not inadequate ventilation and cooling. The circuit breaker was replaced with no further evidence of the problem of tripping on thermal overload.
The inspecter has no further questions in this area.
3.
Licensee Action on IE Bulletins, Circulars and Information Notices a.
IE Bulletin 79-14: Seismic Analyses for As-Built Safety-Related Piping Systems Licensee actions for this bulletin were reviewed in NRC Inspection Reports 50-334/80-20, 80-27, 81-10 and 81-18.
Because the remaining commitments are being tracked by Unresolved Item 82-08-03 (as updated by IR 82-10), this bulletin is close _
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b.
IE Bulletin 79-21: Temperature Effects on Level Measurements NRC Inspection 50-334/80-25 reviewed the licensees action in regard to IEB:79-21. The inspector had examined the steam generator level loop protection channel calibration and channel test procedures (MSP 24.01 - 24.09 and 24.17 - 24.25) on September 10, 1980, and de-termined that they had not been revised to change the low-low level setpoints from 10% to 12%. The most current completed tests were reviewed on October 29, 1982 to verify that the set point changes have now been implemented. This completes the licensee required action and this Bulletin is closed.
4.
Plant Operations a.
General Inspection tours of the plant areas listed below were conducted during both day and night shifts with respect to Technical Speci-fication (TS) compliance, housekeeping and cleanliness, fire pro-tection, radiation control, physical security and plant protection, operational and maintenance administrative controls.
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Control Room
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Primary Auxiliary Building
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Turbine Building
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Service Building Main Intake Structure
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Main Steam Valve Room
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Purge Duct Room
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East / West Cable Vaults Emergency Diesel Generator Rooms
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Containment Building 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 include the following:
BVPS FSAR Appendix A, 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 OM 1.48.6, Clearance Procedures
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OM 1.48.8, Records
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OM 1.48.9, Rules of Practice
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OM Chapter 55A, Periodic Checks - Operating Surveillance Tests
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BVPS Maintenance Manual (MM), Chapter 1, Conduct of Maintenance
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10CFR50.54(k), Control Room Manning Requirements BVPS Site / Station Administrative Procedures (SAP)
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BVPS Physical Security Plan (PSP)
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Inspector Judgement
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Operations The inspector toured the Control Room regularly to verify compliance with NRC requirements and facility technical specifications (TS).
Direct observations of instrumentation, recorder traces and control panels were made for items important to safety.
Included in the reviews cre the rod position indicators, nuclear instrumentation systems, radiation monitors, containment pressure and temperature parameters, onsite/offsite emergency power sources, availability of reactor protection system and proper alignment of engineered safety feature systems. Where an abnormal condition existed (such as out-of-service equipment), adherence to appropriate TS action statements were independently verified. Also, various operation logs and records, including completed surveillance tests, equipment clearance permits in progress, status board maintenance and tempo-rary operating procedures were reviewed on a sampling bases for compliance with technical specifications and those administrative controls listed in paragraph 4a.
During the course of the inspection, discussions were conducted with operators concerning reasons for selected annunciators and knowledge of recent changes to procedures, facility configuration and plant conditions. The inspector verified adherence to approved procedures for ongoing activities observed.
Shift turnovers were witnessed and staffing requirements confirmed.
Except as noted below, inspector comments or questions resulting from these daily reviews were acceptably resolved by licensee personnel.
(1) Prior to the weekend of October 30-31, 1982, intermittent flow oscillations were observed in the "B" feedwater loop. The licensee opted to reduce reactor power to less than 30% and use the bypass control valve to maintain steam generator water level while the main feed regulator valve, FW-488, was isolated and out-of-service for repairs. The valve trim (plug and cage), I/P converter, air regulator and positions were replaced on FW-488. This appears to have corrected the flow oscilla-tions.
Before returning the valve to service on October 30, 1982, OST 1.1.10, Cold Shutdown Valve Exercise Test, was employed as guidance for stroking FW-488. An initial condition for com-pleting all portions of this OST requires that the station be cooled down to less than 200 F.
Though there is no safety significance associated with exercising valve FW-488 under those existing operating conditions, the inspector noted that this appeared to be another example where trouble shooting or other maintenance activities were conducted using portions of an OSC approved procedure that specified initial conditions not applicable to the task at hand.
This was discussed with the Manager of Nuclear Operations and the Plant Superintendent in conjunction with corrective actions being considered by DLC in
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response to a violation cited in NRC Inspection Report 50-334/82-22. Because the licensee's response to the IR:82-22 violation will address this area of concern, the inspector had no further questions at this time.
(2) While passing through the east cable mezannine, on October 28, 1982, an individual grabbed a cable tray for support and observed temporary arcing from one of the cables at the point where it left the horizontal tray for a vertical rise. This ground caused the number 1 vital bus to momentarily spike low, resulting in numerous control room alarms.
Investigation by the licensee identified a hole in the insulation of the cable cable apparently caused by contact with a bare metal edge of the cable tray.
Review of the licensee's corrective actions for repairing the vital bus cable is an inspector follow item (82-26-01).
c.
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:
Protected area barriers were not degraded;
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Isolation zones were clear; Persons and packages were checked prior to allowing entry into
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the Protected Area; Vehicles were properly searched and vehicle access to the
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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 manned, equipped, and security personnel were alert and knowledgeable regarding position requirements, and that written procedures were available; and Adequate lighting maintained.
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No inadequacies were observed.
d.
Radiation Controls Radiation controls, including posting of radiation areas, the conditions of step-off pads, disposal of protective clothing, completion of Radiation Work Permits, compliance with Radiation Work Permits, personnel monitoring devices being worn, cleanliness of work areas, radiation control job coverage, area monitor operability
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(portable and permanent), area monitor calibration, and personnel frisking procedures were observed on a sampling basis.
No violations were identified, e.
Plant Housekeeping and Fire Protection Plant housekeeping conditions including general cleanliness con-ditions and control of material to prevent fire hazards were ob-served in areas listed in paragraph 4a. Maintenance of fire barri-ers, fire barrier penetrations, and verification of posted fire watches in these areas was also observed.
No inadequacies were observed.
On November 30, 1982, the inspector attended an NRC/NRR meeting with Duquesne Light Company regarding issues related to fire protection for BVPS Unit 1 (10CFR 50, Appendix R).
Clarifications were ob-tained from DLC on all concerns.
Issues that needed additional technical justification or information from DLC were:
(1) absence of a source range monitor at a remote location; (2) use of incore thermocouples in place of T-Hot for one instrument train, and the use of steam generator pressure for T-cold indications in the other-train; and, (3) verify breaker protection (especially for diesel generators) for a voltage spike event originating on the secondary side of the current transformer.
Each of the above items are remaining fire protection licensing issues to be resolved by NRR/NRC and DLC.
5.
Engineered Safety Features (ESF) Verification a.
The operability of various ESF systems was verified by performing a complete walkdown of all accessible porgions that included the following as appropriate:
(1) System lineap procedures match plant drawings and the as-built configuration.
(2) Equipment conditions were observed for items which might degrade performance. Hangers and supports are operable.
(3) The interior of electrical and instrumentation cabinets were inspected for debris, loose material, jumpers, etc.
(4) Instrumentation was properly valved in and functioning; and had current calibration dates.
(5) Valves were verified to be in the proper position with power available.
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(6) Technical Specification required surveillance testing was current.
The Containment Depressurization System, including the Quench Spray, Chemical Addition (NaOH), and Inside and Outside Recirculation Spray subsystems were inspected on November 2-3, 1982. The Auxiliary Feedwater System was inspected on November 17, 1982.
b.
Other selected ESF trains were inspected to verify operability of (
major flow paths and components.
ESF trains so inspected were:
(1) Diesel Generator No. I and 2 - October 29, 1982 (2) Chemical and Volume Control System - November 23, 1982 6.
Surveillance Activities Portions of various surveillance tests were observed to verify that:
(1)
technical specification test frequencies were met (2) the procedure was followed, (3) testing was performed by qualified personnel, (4) LCOs were being met, and (5) system restoration was correctly accomplished follow-ing the tests.
The following activities were witnessed by the inspector:
OST 1.13.1,1A Quench Pump - Flow Test, performed on November 2,
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OST 1.7.1, Boric Acid Transfer Pump (1CH-P-2A) Operational Test, performed on November 2, 1982.
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OST 1.5A.1. Delta Flux Alarm Program Operability Check, performed November 16, 1982.
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OST 1.30.10, Silt Check-Main Intake Structure, performed November 19, 1982.
OST 1.1.1, Control Rod Assembly Partial Movement Test, performed on
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November 23, 1982.
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MSP 2.04, NI-42 Power Range Monitor Quarterly Calibration, performed November 22, 1982.
No violations were identified.
7.
In Office Review of Licensee Event Reports (LERs)
The inspector reviewed LERs submitted to the NRC:RI office to verify that the details of the event were clearly reported, including the accuracy of the description of cause and adequacy of corrective action. The inspec-tor determined whether further information was required from the licens-ee, whether generic implications were indicated, and whether the event warranted onsite followup.
The following LERs were reviewed:
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LER 82-41/03L Control Room Chlorine Detector Inoperable.
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LER 82-42/03L One Supplementary Leak Collection and Release
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System Train Inoperable.
LER 82-43/03L*
Failure to Log Delta Flux With Axial Flux
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Difference Monitor Alarm Inoperable.
LER 82-44/03L*
Failure to Establish Fire watch for a
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Non-functional Fire Penetration.
LER 82-45/03L Containment Vacuum Pump 1A Inoperable
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LER 82-46/03L Indicated Control Rod Misalignment LER 82-47/03L Pressurizer Level Channel I Indicator out of 4%
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Tolerance Band.
LER 82-50/03L Tritium Activity Downstream of BVPS above ETS
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Reporting Requirements.
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LER 82-48/03L*
loss of one Offsite AC Power Source.
- denotes those reports selected for onsite followup.
No unacceptable conditions were identified.
8.
Onsite LER Followup The inspector reviewed the licensees actions for the followi.ng LERs:
LER:
82-43: At about 10:00 a.m. on October 13, 1982, the plant computer (P-250) began to initiate spurious alarm and print out nonsensical data on the alarm types. After extensive trouble shooting, the Computer Engineer notified the Nuclear Shift Supervisor at 5:00 p.m. that all analog monitoring functions were lost. The Axial Flux Difference (AFD)
Monitor Alarm was one of those lost functions. Technical Specification 4.2.1.1.6 requires that when the AFD Monitor Alarm is inoperable, the axial flux difference for each operable excore channel must be monitored and logged at least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This was not realized by the licensee until the shift turnover at 11:30 p.m. at which time the log was started. During this period, an unrelated power tran-sient had forced a power reduction, requiring frequent visual observa-tions of the delta flux indicators to control delta flux within the target band.
Corrective actions stated in the LER included discussions with the individual involved and the computer engineer to assure aware-ness of the TS surveillance requirements associated with a computer failure. Other operators were aware of these requirements and had noted the failure to log values at shift turnover. The Plant Superintendent informed the inspector that current operational requirements would be strictly enforced.
Failure to log the axial flux difference while the
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alarm was inoperable from 5:00 p.m. to 11:30 p.m. on October 13, 1982 is a licensee identified violation of Technical Specification 4.2.1.1.b.
LER 82-44: This LER provides additional reportable information that was inadvertently omitted from LER 82-04, RHR Pump Inoperable Due to 4kV Bus Fault.
It specifically refers to a fire penetration left unsealed after running a jumper cab!e between MCC-E9 and MCC-E10 to re-establish a temporary power source for the A RHR pump. NRC Inspection Report No.
50-334/82-01 provides further discussion of this event. Because the Offsite Review Committees review of LER 82-04 determined that additional information was required to satisfy TS 6.9.1.9.6, this is consider (d a licensee identified violation and will not be cited.
LER 82-48: One of the two independent offsite AC power sources was lost due to an apparent overcurrent condition detected by the IB Station Service Transformer primary side overcurrent relay (type ITE 51I solid state, manufactured by ITE Imperial Corporation) at 0845 hours0.00978 days <br />0.235 hours <br />0.0014 weeks <br />3.215225e-4 months <br /> on October 18, 1982.
Emergency bus loads were picked up by the No. 2 Diesel Genera-tor until offsite AC power could be restored to the emergency buses at 0946 hours0.0109 days <br />0.263 hours <br />0.00156 weeks <br />3.59953e-4 months <br />.
The ITE 51I relay was replaced and tested. Though the licensee could not determine the cause of its actuation, a malfunction is suspected because the other overcurrent relays providing transformer secondary side and 4kV protection were not actuated during the event.
The inspector had no further questions on this item.
9.
Unresolved Items Unresolved items are matters about which more information is required to determine whether they are acceptable, items of noncompliance or de-viations. No new unresolved items were identified during this report period, followup on several previous unresolved items are discussed in Section 2.
10.
Exit Interview Meetings were held with senior facility management periodically during the course of this inspection to discuss the inspection scope and find-ings. A summary of inspection findings was also provided to the licensee at the conclusion of the report period.
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