IR 05000324/1982024

IE Insp Repts 50-324/82-24 & 50-325/82-24 on 820515-0615.No Noncompliance Noted.Major Areas Inspected:Review of Lers, Maint Activities,Surveillance Activities,Operational Safety Verification & Independent Insp Effort

ML20058G045
Person / Time
Site:	Brunswick
Issue date:	07/09/1982
From:	Burger C, Garner L, Myers D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20058G044	List: IR 05000324/1982024 ML20058G043
References
50-324-82-24, 50-325-82-24, NUDOCS 8208030182
Download: ML20058G045 (10)

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[ ? UNITED STATES

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g ka NUCLEAR REGULATORY COMMISSION REGION li e # 101 MARIETTA ST N.W., SUITE 3100 b

g..... ATLANTA, GEORGIA 30303 Report Nos. 50-324/82-24 and 50-325/82-24 Licensee: Carolina Power and Light Company 411 Fayetteville Street Raleigh, N. C. 27602 Facility Name: Brunswick Docket Nos. 50-324 and 50-325 License Nos. DPR-62 and DPR-71 Inspection at Brunswick ite near Wilmington,, North Carolina Inspectors: ,

e M/4v . I9 A D. O. Myers / / // DpeS~rted 0 h . daun L. W. Garde //

i 7 9bx ned Approved by: , ,

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[9 2-C. W.'Burgerr Chief ivision of Project and D/te 'S/grfed Resident Program SUMMARY Inspection on May 15 - June 15, 1982 Areas Inspected Total inspection involved 280 inspector hours on site in the areas of review of licensee event-reports, maintenance activities, surveillance activities, followup of plant transients, operational safety verification, independent inspection, review of shutdown and refueling activitie Results Of eight areas inspecteJ, no violations were identifie PDR ADOCK 05000324 G PDR

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DETAILS Persons Contacted Licensee Employees A. Bishop, Technical and Administrative Manager J. Boone, Engineering Supervisor J. Cook, E & RC Foreman

• C. Dietz, General Manager, Brunswick E. Enzor, I&C/ Electrical Maintenance Supervisor

• M. Hill, Maintenance Manager

• R. Knobel, Manager of Operations

• R. Morgan, Plant Operations Manager D. Novotny, Regulatory Specialist G. Oliver, E&RC Manager L. Boyer, Administrative Manager

-* Poulk, Regulatory Specialist L. Tripp,'RC Supervisor

• W. Tucker, Manager of Operation V. Wagner, Director, Planning and Scheduling Other licensee employees contacted included technicians, operators and engineering staff personne * Attended exit intervie . Exit Interview The inspection scope and findings were summarized on June 18, 1982 with those persons indicated in Paragraph 1 above. Meetings were also held with senior facility management periodically during the course of this inspection to discuss the inspection scope and finding . Licensee Action on Previous Inspection Findings Not inspecte . Unresolved Items Unresolved items were not identified during this inspectio . Onsite Review Committees The inspectors attended Plant Nuclear Safety Committee (PNSC) Meeting conducted on 5-26-8 .J

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The inspector verified tha following items:

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Meetings were conducted in accordance with Technical Specification requirements regarding quorum membership, review process, frequency and personnel qualifications;

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Meeting minutes were reviewed to confirm that decisions / recommendations were reflected and follow-up of corrective actions were complete No violations were identifie . Operational Safety Verification The inspector verified conformance with regulatory requirements throughout the reporting period by direct observations of activities, tours of facilities, discussions with personnel, reviewing of records and independent verification of safety system status. The following determinations were made:

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Technical Specifications. Through log review and direct observation during tours, the inspector verified compliance with selected Technical Specifications Limiting Conditions for Operatio By observation during the inspection period, the inspector verified the control room manning requirements of 10 CFR 50.54 (k) and the Technical Specifications were being met. In addition, the inspector observed shift turnovers to verify that continuity of system status was maintained. The inspector periodically questioned shift personnel relative to their awareness of plant condition Control room annunciators. Selected lit annunciators were discussed with control room operators to verify that the reasons for them were understood and corrective action, if required, was being taken.

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Monitoring instrumentation. The inspector verified that selected instruments were functional and demonstrated parameters within l

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Technical Specification limit Safeguard system maintenance and surveillance. The inspector verified by direct observation and review of records that selected maintenance i and surveillance activities on Safeguard Systems were conducted by l qualified personnel with approved procedures, acceptance criteria were met and redundant components were available for service as required by

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Technical Specification Major components. The inspector verified through visual inspection of selected major components that no general condition exists which might

prevent fulfillment of their functional requirements.

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Valve and breaker positions. The inspector verified that selected valves-and breakers were in the position or condition required by Technical Specifications for the applicable plant mode. This verification included control board indication and field observation (Safeguard Systems).

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Fluid leaks. No fluid leaks were observed which had not been identified by station personnel and for which corrective action had not been. initiated, as necessar Radioactive releases. The inspector verified that selected liquid and gaseous releases were made in conformance with 10 CFR 20 Appendix B and Technical Specification requirement Radiation Controls. The inspector verified by observation that control point procedures and posting requirements were being followed. The inspector identified no failure to properly post radiation and high radiation area Securi ty. During the course of these inspections, observations relative to protected and vital area security were made, including access controls, boundary integrity, search, escort, and badgin Of the areas and activities reviewed the following inspector concerns were noted, Housekeeping During the inspection period inspectors noted severci areas where housekeeping met only the minimum requirements of administrative Procedure AI 17. The areas of concern included the 3 foot level of the radwaste building, the south RHR room and HPCI room of the

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U-1 reactor building. The areas and inspector concerns were brought to the attention of licensee management who appeared concerned and initiated corrective actions. Inspectors will i

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continue to monitor these areas to insure poor housekeeping trends are not developing, Valve Indentification

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During plant tour on May 25, it was observed that the cooling water supply valves SW-V212 to diesel generator DG No. 3 were mislabeled. The supply valve from Unit 1 service water was designated as being from Unit 2 service water and vice versa. In addition both cooling water supply valves SW-V210 to diesel generator no.1 were labeled as from Unit 2. The correct valve numbers were obtained from drawing F.P. 9527-22351, 22369, 22349 and 22367. The licensee has corrected the problem on DG no.3 and is in the process of making a new tag for DG N i

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4 Local MCC Panel Indications During unit 1 and 2 elevation tour on June 10, eleven motor control center (MCC) breakers were noted to be on but had neither lamp lit due to burned out bulbs. These lamps indicate valve position or pump motor status. Five of these, E51-F045, E51-V8, E11-F028B, E11-f009,~SW-V117, are to be locally controlled from MCC El-29, where plant shutdown from outside control room is implemented. All eleven were verified to be properly aligned by observation of primary indication in the control room. Replace-ment of burned out bulbs is the responsibility of each shift's auxilary operators. The relatively large number of burned out bulbs was pointed out to plant management for corrective actic No violations were identified in this are . Review of Licensee Event Reports The below listed Licensee Event Reports (LERS) were reviewed to determine if the information provided met NRC reporting requirements. The determination included adequacy of event description and corrective action taken or planned, existence of potential generic problems and the relative. safety significance of each event. Additional in-plant reviews and discussions with plant personnel, as appropriate, were conducted for those reports indicated by an asteric UNIT 1 1-81-83 (3L) Suppression Chamber Water Level Recorder, 2-CAC-LR-2602, out of calibratio (3L) Failure of Primary Containment Atmospheric 0xygen Analyzer, 1-CAC-AT-1259- (3L) No indication on RTGB for E11-F048-A Valve due to magnetic and trip having tripped breaker with valve in open positio Supplement "A" loop of Suppression Pool Cooling inoperabl (3L) Post-accident Drywell Particulate Radiation Detection Instruments, 1-CAC-AQH-1261-1 and 1262-1, indicating downscal *1-82-19(3L) "B" Subsystem Pressure Switch, 1-SW-PS-1176, actuated intermittently in response to test signal and "B" RHRSW ,

Subsystem declared inoperabl *1-82-22 (3L) 1A Reactor Recirculation Pump tripped due to MG set undervoltag ,

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1-82-28 (3L) Suppression Chamber Water Level Check revealed that RTGB Instruments, 1-CAC-LR-2602 and 1-CAC-LI-2601-3, out of calibratio (3L-) Failure of Primary Containment Atmospheric 0xygen Analyzer, 1-CAC-AT-1263- *1-82-30(3L) HPIC -Steam Line Area High Differential Temperature Switch, 1-E51-DTS-N604D, would not respond to test signal due to electrical failure of switch modul ,

1-82-31(3L) RTGB Suppression Chamber Water Level Instruments inoperabl (3L) Post-accident Monitoring Control Rocm Recorder / Indicator, 1-CAC-AR-1263, observed exhibiting unvarying indication of Drywell Oxygen Concentratio *1-82-43 (3L) Exhaust Ventilation Radiation High Instrument, 1-D12-RM-N010A, inadvertently actuated, resulting in a Group 6 PCIS signa (3L) Suppression Chamber Water Level Check revealed RTGB Instru-ment,1-CAC-LR-2602, out of calibratio *1-82-46 (3L) Specific Gravity for Pilot Cell of Division II Battery', .1B-2, was less than the specified requiremen UNIT 2 2-80-73 (3L) Failure of Containment Atmospheric 0xygen Analyzer, 2-CAC-ATH-1259- (3L) Failure of Containment Atmospheric Monitor, 2-CAC-AT-125 (3L) Failure of Primary Containment Oxygen Analyzer, 2-CAC-AQH-1263- (3L) Failure of Primary Containment Atmospheric Monitor Oxygen Analyzer, 2-CAC-ATH-1259- (3L) Primary Containment Atmospheric 0xygen Analyzer, 2-CAC-ATH-1259-2, giving false reading cue to loose pipe couplin (3L) Suppression Chamber Water Level Indicator, 2-CAC-Li-2601-3, out of calibratio *2-81-138 (3L) Following a Reactor Scram from 15% power, Reactor Coolant activity exceeded Technical Specification limi ;

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• 2-81-140 (3L) A Reactor Coolant l i Dose Equivalent Analysis revealed dose equivalentexceededNchnicalSpecificationsandnine subsequent analyses had not been performe (3L) Failure of Primary Containment Atmospheric 0xygen Analyzer, 2-CAC-AT-1263- (3L) Failure of Primary Containment Atmospheric 0xygen Analyzer, 2-CAC-AT-1259- (3L) RCIC System automatically isolated due to receipt of a spurious steam leak detection differential temperature "A" Channel Isolation signa (3L) Failure of Primary Containment Atmospheric 0xygen Analyzer, 2-CAC-AT-1259- *2-82-43(3L) The " Tripped /Not Available" and " Abnormal Condition" Annunciators for No. 3 Diesel Generator received at the RTG (3L) Post-accident Monitoring Control Room Recorder / Indicator, 2-CAC-AR-1263, observed exhibiting downscale indications of Drywell Oxygen Concentratio *2-82-60 (3L) Specific Gravity for Pilot Cell of Division II No. 34 and 35 of 125VDC Division II Batteries 28-1 and 28-2, were higher than indicated maximum level mar (3L) Failure of Primary Containment Atmospheric 0xygen Analyzer, 2-CAC-AT-1263- (3L) To determine cause of spurious RCIC Steam Leak Detection Isolation signal, HPCI System automatically isolated and declared inoperabl (3L) CRD Accumulator High Water Level Annunciator received for-Hydraulic Control Unit (HCU) 30-43 and declared inoperabl (3L) Suppression Chamber Water Temperature Recorder 2-CAC-TR-778, showed normal indications of temperature but was not recording the (3L) Failure of Drywell Hydrogen Analyzer, 2-CAC-AT-1263- (3L) While performing plant modification to relocate Hydraulic Snubber, 2-E11-10755573, it had been removed without notifying operation (3L) Suppression Chamber Water Level Instrumentation Check revealed Remote Shutdown Panel Instrument, 2-CAC-LI-3342, out of calibratio .

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2-82-75(3L) Drywell Floor Drain (DWFD) Flow Integrator, 2-G16-FQ-K601, continuously indicated DWFD Sump Flow with no DWFD Pumps running and declared inoperabl No violations were identified in this area.

8. Followup of Plant Transients and Safety System Challenges During the period of this report, a followup on plant transients and safety system challenges was conducted to determine; the cause, that safety systems and components functioned as required that corrective actions were adequate; and the plant was maintained in a safe conditio On June 1, 1982 at 1014 hours0.0117 days <br />0.282 hours <br />0.00168 weeks <br />3.85827e-4 months <br />, Unit I reactor experienced a turbine stop valve (TSV) fast closure trip from 80% of full power. No engineered safeguard features (ESF) were required. Vessel level and pressure were centrolled by use of the feedwater and n'ain conJenser systems. Investigation revealed that PS-105A, low condenser vacuum turbine trip switch, had failed and shorted to ground. The resulting false low vacuum signal actuated the turbine master trip system and closed the TSV's. The defective switch was replaced and criticality was achieved at 0206 hours0.00238 days <br />0.0572 hours <br />3.406085e-4 weeks <br />7.8383e-5 months <br /> on June On June 2, 1982 at 1032 hours0.0119 days <br />0.287 hours <br />0.00171 weeks <br />3.92676e-4 months <br />, Unit 1 reactor experienced a high pressure trip from 24% of full power when all four turbine bypass valves closed from full open. No ESF were required. Vessel Level and'

pressure were controlled by use of the feedwater and main condenser systems. Reactor pressure spiked to approximately 1060 psig. Prior to the scram, trouble had been experienced in placing the turbine in service due to electro-hydraulic control (EHC) problems. Reactor power had been reduced to 24% with all four bypass valves full open so that instrument and control, technicians could troubleshoot the EHC syste Investigation revealed that during the troubleshooting a circuit card involving the control valves logic was unpluged. The I&C personnel were unaware that the card also involved the bypass valve low vacuum logic. Removal of the card, de-energized the low vacuum logic thereby causing the bypass valves to close. When the unit tripped, the card was immediately inserted and the bypass valves re-opened. Reactor startup commenced the same day at 1256 hour0.0145 days <br />0.349 hours <br />0.00208 weeks <br />4.77908e-4 months <br /> On June 2 at 2054 hours0.0238 days <br />0.571 hours <br />0.0034 weeks <br />7.81547e-4 months <br />, Unit 1 reactor experienced an intermediate range monitor neutron flux high trip from 5.5% of full power. No ESF were required. Vessel level and pressure were controlled by use of the feedwater and main condenser system .

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Just preceding the scram, a circuit failure in the startup level control valve, FW-V177, circuitry caused FW-V177 to come full ope FW-V177 rejects feedwater from the reactor feed pump discharge header to the condenser hotwell. When reactor vessel level began decreasing, the control operator took manual control of the feedwater system to prevent a low water scram. However, too much cold water was added too rapidly. The cold water caused a neutron flux spike and resulting high flux tri On June 17, 1982 at 2125 hours0.0246 days <br />0.59 hours <br />0.00351 weeks <br />8.085625e-4 months <br /> unit 1 experienced a scram from 79% full power when all the main steam isolation valves closed from a Group 1 isolation signal. The reactor core isolation cooling (RCIC) system was called upon to start when vessel level decreased to low level No 2 but failed to inject water into the vessel. The high pressure coolant injection system responded as designed and supplied water for vessel level control. System pressure control was achieved by reactor operators using vessel relief valves to the suppression chamber in accordance with reactor scram procedure EI-31. Maximum system pressure was 1065 psig. The Group I isolation signal was subsequently reset and the main condenser was made available as a heat sink for normal cool-dow Licensee investigation revealed the Group I isolation occurred due to an undetected blown fuse in isolation channel "B" for DC control circuitry of the MSIV's coincident with a main steam line high radia-tion signal on isolation channel "A". The high radiation signal was false and was determined to be inducted by circuit noise resulting from ongoing neutron monitoring system testin The failure of the RCIC system to inject water was determined to be caused by a DC ground in the governor valve control circuitry. The control circuitry received the initiation signal, the steam supply valve (F045) opened and the turbine governor valve began to close to limit turbine acceleration. A ramp generator in the circuit provides the controlled acceleration and is designed to allow the system water flow signal to regulate governor valve position hence turbine speed, once the system has attained rated flow. However, the DC ground affected the control system ramp generator in a manner which simulated a demand for a governor valve position of closed, effectively shutting off the turbine steam supply. Technicians replaced the defective component and subsequently were able to attain satisfactory system performance as evidenced by successful performance of P.T 10.1.1 and 10.1.3. The unit was restarted on June 8, 198 No violations or deviations were identified in this are . Surveillance Testing The surveillance tests detailed below were analyzed and/or witnessed by the inspector to ascertain procedural and performance adequacy.

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The completed test procedures examined were analyzed for embodiment of the necessary test prerequisites, preparations, instructions, acceptance criteria and sufficiency of technical conten .

The selected tests witnessed were examined to ascertain that current, written approved procedures were available and in use, that test equipment in use was calibrated, that test prerequisites were met, system restoration was completed and test results were adequat The selected procedures perused attested conformance with applicable Technical Specifications, they had received the required administrative review and were performed within the surveillance frequency prescribe Procedure Ti tle Date PT Reactor-low water level #1 channel 6/4/82 calibration and functional test PT Emergency core cooling system 3/9/82 PT 1 Plant Battery 6/7/82 The inspector employed one or more of the following acceptance criteria for evaluating the above items:

10 CFR ANSI N1 Technical Specifications Of the areas inspected no violations were identifie . Maintenance Observations Maintenance activities were observed and reviewed throughout the inspection period to verify that activities were accomplished using approved procedures, that the activity was within the skill of the trade performing the work and that the work was donc by qualified personnel. Where appropriate, limiting conditions for operation were examined to ensure that while equipment was removed from service the technical specification requirements were satisfied. Also, work activities, procedures, and trouble tickets were reviewed to ensure adequate fire, cleanliness and radiation protection precautions were observed and that the equipment was tested and properly returned to service. Throughout the period the inspector selectively reviewed outstanding trouble tickets to ensure the licensee is giving priority to safety-related maintenance and not allowing a backlog of work items to permit a degradation of the systems functio Of the areas inspected no violations were identifie __

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1 Refueling Activities

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The inspectors continued direct inspection of the current refueling activities on Unit 2. The core reload was completed on June 13 with

. location and orientation of all 560 fuel assemblies-verified satisfactorily per procedure FH-11 by the licensee. Inspectors witnessed core verification in progress by the licensee personnel and independent verification was performe Refueling floor observations included fuel bundle sipping. The i licensee reports that all fuel reloaded into the core was subjected to clad integrity examination using fuel sipping. The process lead to the discovery of 2 leaking 8x8 fuel bundles. Subsequent detailed video examination of the

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leaking assemblies indicated small cladding failures suspected to be manufacturing flaws. General Electric, the fuel supplier is performing an evaluation.

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refueling area housekeeping observations

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verification of adequate staffing

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verification of source range instrumention available and operable per T.S. 3. verification of containment integrity during fuel handling by review of procedures I

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health physics practices in the refueling are verification that the spent fuel pool water level was maintained and that system performance was adequate.

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! Of the areas inspected no violations was identified.

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