IR 05000324/1988014

Insp Repts 50-324/88-14 & 50-325/88-14 on 880301-31. Violations Noted.Major Areas Inspected:Followup on Enforcement Matters,Maint Observation,Surveillance Observation & Operational Safety Verification

ML20153G718
Person / Time
Site:	Brunswick
Issue date:	05/05/1988
From:	Fredrickson P, Ruland W, Tobin W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20153G699	List: ML20153G696 ML20153G710 ML20153G718
References
50-324-88-14, 50-325-88-14, NUDOCS 8805110322
Download: ML20153G718 (14)

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' '4farg# UNITED STATES

'o NUCLEAR REGULATORY COMMISSION

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. TA REGION 11 hs

• i 'e 101 MARIETTA STRE ET, ATL ANT A, G EORGI A 30323

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Report No. 50-325/88-14 and 50-324/88-14 Licensee: Carolina Power and Light Company P. O. Box 1551 Raleigh, NC 27602 Docket No. 50-325 and 50-324 License No. DPR-71 and DPR-62 Facility Name: Brunswick 1 and 2 Inspection Conducted: March 1 - 31, 1988 Inspector: W ! f on s ,.* : / / )

W. H. Ruland g["-

k f/$88 Date' Signed Accompanying Personnel: S. M. Shaeffer R. B. Latta E. H. Girard Inspector:

(Paragraph 10 . .

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(h Date' Signed Approved By: J .

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P. E. Fredrickson, Section Chief e Mb Bate' Signed Reactor Projects Section 1A Division of Reactor Projects SUMMARY Scope: This routine safety inspection by the resident inspector and others involved the areas of followup on previous enforcement matters, maintenance observation, surveillance observation, operational safety verification, in-office Licensee Event Report (LER) review, followup on inspector identified and unresolved items, plant modifications, allegation followup, and unusual event Results: Three Unit i violations were identified: Failure to determine reactor vessel pressure and shell temperature every 30 minutes during plant heatup; failure to verify position of an unlocked manual CAD valve in its correct position every 31 days; and failure to perfonn a 10 CFR 50.59 evaluatio A potentially significant safety issue was identified by the licensee regarding operator inattention during shutdown condition A licensed operator unknowingly allowed reactor coolant system temperature increase 90 degree This item was identified as a licensee-identified violatio hDj ]ggCK22 88050s O 05000324 DCD

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REPORT DETAILS Persons Contacted Licensee Employees W. Biggs, Principal Engineer

• E. Bishop, Manager - Operations

• J. Brown, Res. Engineer - Engineering

• Callis, Onsite T. Cantebury, Licensing Mechanical Engineer -Supervisor Maintenance Licensing Unit (& Nuclear 1) Fuel G. Cheatham, Manager - Environmental & ladiation Control R. Creech, I&C/ Electrical Maintenance Supervisor (Unit 2)

• C, Dietz, General Manager - Brunswick Nuclear Project W. Dorman, Supervisor - QA

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• R. Eckstein, Manager - Technical Support

• K. Enzor, Director - Regulatory Compliance R. Groover, Manager - Project Construction W. Hatcher, Supervisor - Security A. Hegler, Superintendent - Operations R. Helme, Director - Onsite Nuclear Safety - BSEP J. Holder, Manager - Outages

• P. Howe, Vice President - Brunswick Nuclear Project

• L. Jones, Director - Quality Assurance (QA)/ Quality Control (QC)

• M. Kesmodel, Supervisor - Document Control R. Kitchen, Mechanical Maintenance Supervisor (Unit 2)

J. Moyer, Manager - Training G. Oliver, Manager - Site Planning and Control J. O'Sullivan, Manager - Maintenance B. Parks, Engineering Supervisor

• R. Poulk, Senior NRC Regulatory Specialist

• D. Queener, Principal Engineer - Environmental & Radiation Control J. Smith, Manager - Administrative Support

• J. Titrington, Principal Engineer - Operations V. Wagoner, Director - IPBS/Long Range Planning

• R. Warden, I&C/ Electrical Maintenance Supervisor (Unit 1)

D. Warren, Acting Engineering Supervisor B. Wilson, Engineering Supervisor

• T. Wyllie, Manager - Engineering and Construction Other licensee employees contacted included construction craf tsmen, engineers, technicians, operators, office personnel, and security force member * Attended the exit interview

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2 Exit Interview (30703)

The inspection scope and findings were summarized on March 31, 1988, with those persons indicated in paragraph 1. The inspectors described the areas inspected and discussed in detail the inspection findings listed belo Dissenting comments were not received from the license Proprietary information is not contained in this repor Item Number Description / Reference Paragraph 325/88-14-01 VIOLATION - Missed Surveillance of Reactor Vessel Temperature and Pressure During Inadvertent Heatup (paragraph 8.a). (A licensee identified violation was also reviewed.)

325/88-14-02 VIOLATION - Surveillance Procedure Failed to Check a CAD Valve Position (paragraph 8.c).

325/88-14-03 VIOLATION - Failure to Perform 10 CFR 50.59 Evaluation On Service Water System (para-graph 8.b).

325,324/88-14-04 IFl - Correction of CAD System Hardware Problems (paragraph 8.c).

Note: Acronyms and abbreviations used in this report are listed in paragraph 1 . Followup on Previous Enforcement Matters (92702)

(CLOSED) Violation 325/86-29-01; Improper Temporary Revision to OP-03 When Operating EPA Breakers, response dated January 2, 198 The inspectors reviewed OP-03, Revision 8 for Unit 1 and OP-03, Revision 19 for Unit 2 and detarmined that these procedures had been modified to provide for the opening of the EPA breakers prior to opening the MG set output breaker. These revisions adequately address the issue identified in the violation. This item is close (CLOSED) Violation 325/87-31-01 and 324/87-35-01; Inadequate DG Surveill-ance Procedure - TCV Not in Restoration Lineup, response dated November 25, 1987. The inspectors reviewed PT-12.2A, Revision 32, PT-12.2B, Revision 31, PT-12.2C, Revision 31, and PT-12.2D, Revision 33 and deter-mined that these procedures had been modified to include the subject temperature control valves in the system restoration valve lineu Additionally, these procedures have been revised to provide operational information regarding normal and abnormal valve operation and required corrective action. The revised pts adequately address the issue identi-fied in the violation. This item is close . _ _

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) MaintenanceObservation(62703)

The inspectors observed maintenance activities, interviewed personnel, and reviewed records to verify that work was conducted in accordance with approved procedures, Technical Specifications, and applicable industry codes and standards. The inspectors also verified that: redundant components were operable; administrative controls were followed; tagouts were adequate; personnel were qualified; correct replacement parts were used; radiological controls were proper; fire protection was adequate; quality control hold points were adequate and observed; adequate post-maintenance testing was performed; and independent verification require-ments were implemente The inspectors independently verified that selected equipment was properly returned to servic Outstanding work requests were reviewed to ensure that the licensee gave priority to safety-related maintenance. '4umerous maintenance items were reviewed throughout the reporting perio The inspectors observed /

reviewed, in detail, those portions of the following maintenance activi-ties:

86-BAFF1 1-CAC-V10 Outboard Drywell Purge Exhaust Valve 88-AIYB1 DG No. 3 Oscillations / Troubleshooting No significant safety matters, violations, or deviations were identifie . SurveillanceObservation(61726)

The inspectors observed surveillance testing required by Technical Speci-I fication Through observation, interviews, and record review, the l inspectors verified that: tests conformed to Technical Specification

! requirements; administrative controls were followed; personnel were qualified; instrumentation was calibrated; and data was accurate and complete. The inspectors independently verified selected test results and proper return of equipment to service.

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The inspectors witnessed / reviewed portions of the following test activi-ties:

1MST-APRM12W APRM Channel B, D, and F Functional Test IMST-BATT11Q Batteries, 125 VDC Quarterly Operability Test IMST-HPCl22M Steamline Low Pressure Channel Calibration IMST-HPCI26M HPCI Suppression Pool High Level Instrument Channel Calibration 2MST-HPCl27M HPCI and RCIC CST Low Water Level Instrument Channel Calibration

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2MST-RPS23R RPS High Reactor Pressure Instrument Channel Calibra-tion 2MST-SEIS21R SMA-3 Seismic Monitoring System Channel Calibration No significant safety matters, violations,.or deviations were identifie6 Operational Safety Verification (71707)

The inspectors verified that Unit 1 and Unit 2 were operated in compliance with Technical Specifications and other regulatory requirements by direct observations of activities, facility tours, aiscussions with personnel, reviewing of records and independent verification of safety system statu The inspectors verified that control room manning requirements of 10 CFR 50.54 and the Technical Specifications were met. Control operator, shift supervisor, clearance, STA, daily and standing instructions, and jumper /

bypass logs were reviewed to obtain information concerning operating trends and out of service safety systems to ensure that there were no conflicts with Technical Specifications Limiting Conditions for Opera-tions. Direct observations were conducted of control room panels, instru-mentation, and recorder traces important to safety to verify operability and that operating parameters were within Technical Specification limit The inspectors observed shift turnovers to verify that continuity of system status was maintained. The inspectors verified the status of selected control room annunciator Operability of a selected Engineered Safety Feature division was verified weekly by ensuring that: each accessible valve in the flow path was in its correct position; each power supply and breaker was closed for components that must activate upon an initiation signal; the RHR subsystem cross-tie valve for each unit was closed with the power removed from the valve operator; there was no leakage of major components; there was proper lubrication and cooling water available; and a condition did not exist which might prevent fulfillment of the system's functional requirement Instrumentation essential to system actuation or performance was verified operable by observing on-scale indication and proper instrument valve lineup, if accessibl The inspectors verified that the licensee's health physics policies /

procedures were followed. This included observation of HP practices and a review of area surveys, radiation work permits, posting, and instrument i calibratio The inspectors verified that: the security organization was properly manned and security personnel were capable of performing their assigned functions; persons and packages were checked prior to entry into the protected area; vehicles were properly authorized, searched and escorted within the PA; persons within the PA displayed photo identification badges; personnel in vital areas were authorized; and effective compen-saton measures were employed when require . _

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The inspectors also observed plant housekeeping controls, verified position of certain containment isolation valves, checked a clearance, and verified the operability of onsite and offsite emergency power source No significant safety matters, violations, or deviations were identifie . In Office Licensee Event Report Review (90712)

The listed LER was reviewed to verify that the information provided met NRC reporting requirements. The verification included adequacy of event description and corrective action taken or planned, existance of poter,tial generic problems and the relative safety significance of the even (CLOSED) LER 2-88-03; Auto-Isolation of Reactor Water Cleanup System Inlet Outboard Isolation Valve 2-G31-F004 With Reactor Defueled. The inspector reviewed the event and had no questions. This iten is close No significant safety matters, violation,, or deviations were identifie . Followup on Inspector Identified and Unresolved Items (92701) (CLOSED) Unresolved Item 325/88-01-03; Inadvertent Heatup During Cold Shutdown. The inspector reviewed the OER which described the event and documented the licensee's root cause determination and '

corrective actions. The licensee found two causes for the event:

(1) operator inattentiveness; (2) the SR0 directed the control  ;

operator to throttle the heat exchanger outlet valve, 1-E11-F003B, to control temperatur This valve has a key lock switch on the main *

control board that only has open and closed positions. The operator had to place the switch in a "dead" position to throttle the valv This action, as well as throttling the F003B valve, was not authorized by the system operating procedur <

i The actions initiated by the licensee to prevent recurrence included disciplinary action for the operators involved, on-shift training,  ;

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evaluation of switch replacement for the E11-F003B valve, development  !

of a routine pre-outage training package, proceduralization of key l parameter monitoring frequency, and evaluation of the need for l appropriate alarms during shutdown conditions.

The inspector reviewed real time training package 88-1-8 prepared for j the event and found it acceptable. The licensee identified the failure to follow the RHR operating procedure and met the require- .

ments of 10 CFR 2 Appendix C, for licensee identified violation '

Therefore, no violation will be issued for the failure to follow procedur However, during the heatup, the licensee failed to determine that the reactor vessel pressure and shell temperature were

. within limits every 30 minutes per TS 4.4.6.1.1. Since the licensee I failed to identify the missed surveillance, a Violation is being l Issued on this particular item: Missed Surveillance of Reactor Vessel Temperature and Pressure During Inadvertent Heatup

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(325/88-14-01).

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b. (OPEN) Unresolved Item 325/88-05-01 and 324/88-05-01; Service Water System Operating Mode Concern Further inspector icview revealed that no 10 CFR 50.59 review was performed as a result of the problem with the V106 valve prior to Unit I start-up on February 20, 198 FSAR table 9.2.1-1, Service Water Flow Distribution - One Reactor Plant, lists Reactor Building Ct'J Heat Exchanger flow rate from the nuclear service water header during the first 10 minutes following a Loss of Coolant Accident as zero gallons per minute. Based on the information that the licensee had received regarding how the service water system could operate during the above time, it was possible with V106 failed open, that flow through the RB CCW HX would be greater than zero gallons per minute. Although no mechanical or electrical components were modified, this information did constitute a change to the Service Water System in that failure of the V106 valve would cause the system to not perform as described in the FSA This change necessitated a 10 CFR 50.59 safety evaluation, prior to restart of the unit, to determine whether an unreviewed safety question existed. The licensee first became aware of this change in the facility ir> February 198 A 10 CFR 50.59 evaluation was not performed until March 22, 1988, after being identified by the inspec-tion in February 198 Failure to perform a written 10 CFR 50.59 safety evaluation is a Violation: Failure to Perform 10 CFR 50.59 Evaluation On Service Water System (325/88-14-03).

The licensee had reviewed the issue at PNSC meeting 88-018 held on February 12, 1988, and concluded that an operability concern did not exist. Actior4 items were established to resolve the issue perman- l entl After questions by the inspector, the licensee completed a JC0 with a 10 CFR 50.59 evaluation (EER-88-167) for the SW V106 issue on March 23, 1988. The EER took credit for the licensee's interim action of limiting RB CCW SW flow to 5,000 GPM. Based on the SW pump curve, with SW flow assumed at 8,600 GPM, the total dynamic head of the pump would be about 108 feet, equivalent to 48 PSIG. This is above the 40 PSIG low header pressure autostart signal for another NSW pump. This portion of the item remains unresolved pending review by the licensee and inspector of the SW design data and a permanent fix, if require c. (CLOSED) Unresolved Item 325/88-05-05 and 324/88-05-05; CAD System Discrepancies. The associated issues were identified during a CAD system walkdown, and are addressed belo The inspector found valve CAC-V168 unlocked in the open (required)

position. All unlocked valves in the system's flow path shall be ,

demonstrated to be operable at least once per 31 days by verifying valve position. Failure to observe the valve position is a violation of TS 4.6.6.2.a.2. The missed surveillance was due to an inadequate procedure, in that valve CAC-V168 does not appear on PT-16.1, Rev.12, Section 7.1.3, System Operability Verification Checklis This valve is a flow path valve in that closure of the valve could prevent the system from injecting nitrogen into the drywell or torus after a single valve failur The inspector reviewed the history of

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p all pericdic test procedures associated with the CAD system for accuracy and timeliness. No other surveillance problems were note This is a Violation: Surveillance Procedure r ailed to Check a CAD  ;

VelvePosition,(325/88-14-02).

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On-February 24, 1988, the inspector observed that the loop A and 0 CAD vessel pressure indicators, PI-2703 and PI-2704, exceeded the maximum allowable working pressure of the CAD vessel which is 100  ;

PS! The pressure indicators were above their maximum gauge  ;

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graduations-(greater than 100 PSIG). The process requirement for the CAD system is between 80 and 90 PSI The CAD vessel had exceeded i these values due to the evaporation of liquid nitrogen within the vesse The design of the system incorporates three separate pressure relief mechanisms: (1) two pressure control valve (PCV-2706 and 2707) with build-up coils designed to maintain 85 psig in the CAS tank; (2) PCV-2705 relieves pressure to at-osphere at 90 psig; and (3) pressure safety valve PS-V3 relieves to atmosphere at 100 psig. There was no evidence that any #f these functions were operating within their setpoints. The licensee's engineering group

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has raceemer.ded that the relief setpoints be calibrated, and if not within tolerance, the valves should be rebui't or exchange The licensee showed the inspector the vendor's hydrostatic test results. The test was performed at 176 PSIG. Thus, no safety concern exists regarding the CAD +ank integrity. Since the licensee had work requests written on the above problems and the safety significanca is small, no violation occurred. The CAD system, -

(lthough addressed in Technical Specifications, is not necessary to  :

prevent any explcsive mixture in a pest-accident containment environ-This was shown by generic analysis done by BWROGs for the

men licensee's hydrogen recombiner rule response. The CAD system would ,

only be needed post-LOCA if the containment '.ad not been inerte This occurs only during reactor startups and shutdown l Minor support problems were identified on the liquid nitrogen lines surrounding the CAD vessel loop 8 end. Apparently the icis.g of the nitrogen lines has caused two wood block supports to deteriorate &nd i become ncn-functional. The inspector observed tha' another support was left disconnected near valve BYCV-2716. The licensee has issued work requests to correct the deficiencies. The licensee has also

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been informed of varimus housekeeping items such as ground cables draped over nitrogen lines and debris in the control cabinets.

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Due to ., axtent >' the deficiences, the inspector will followup on  ;

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the -ns e's cr' '"e actions. This is an Inspector Followup Itr l e- H on ,

' System Hardware Problems (325/88-14-04 and

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, (CLO5ED) Inspector Followup Item 325/86-24-04 and 324/86-25-04;

_ OP-19 Revisions for HPCI CST Recirculation Mode. The licensee .

performed SP-66-080, HPCI System Operability Test, on October 9, 1986. The test checked the operation of the HPCI turbine control system during three conditions: operating in CST full flow test mode; only on the minimum flow valve; and a dual path mode. Problems with the governor ccntrol were noted during certain step changes in the flow demand signal made by the operator. The turbine RPM was hunting with 1,000 RPM swings with a reduced flow demand signal of 3,000 GPM.

A portion of the governor, designated EGM, was retalibrated and SP-86-080 re-performed with no anomalies noted. On November 8, 1986,

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the licensee replaced, under DR-86-0140, the EGM with a newer design, further improving the control syste The inspector reviewed SP-86-080, interviewed plant personnel, and reviewed the current revision (13) of 1-0P-19, HPCI System Operations Procedure. The licensee has added sections to-describe the reactor pressure control mode of the HPCI system. Also, while switching modes, the procedure requires the operator to take the HPCI flow controller to manual. Based on the inspector's review of the l licensee's actions, this item is close [ (CLOSED) Inspector Followup Item 325/87-42-08 and 324/87-43-08; ,

Submission of TS Amendment Request for RWCU Isolation Response Tim The licensee submitted the request for TS amendment on February 29, 1988. This item is close (CLOSED) Inspector Followup Item 325/88-05-03 and 324/88-05-03; Sand Introduced Into Tcrus/ Vacuum Breakers From Failed Inerting Line. The inspectors reviewed the circumstances of the line break and the

adequacy of the corrective actions taken by the licensee. The review

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was conducted through visual observation of the vacuum breakers and t

the replacement inerting line, discussions with the Mechanical  :

Discipline Group Project Engineer, and review of the following Work Request and LLRT records:

Unit 1 Work Requests WR/JO 88-AELT1 WR/JO 88-AEIW1 WR/JO 88-AEIU1 jnit 2 Work Requests g i

WR/JO 88-AENW1 WR/JO 88-AFAC1  !

WR/JO 88-AAUH1 WR/JO 88-AAVII

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Unit 1 LLRTs PT-20.3.71 performed 1/10/88 on V17 PT-20.?.70 performed 1/10/88 on V16 PT-20.3.71 performec' :','10/88 on V17 PT-20.3.70 performed 2/08/88 on V16 Unit 2 LLRTs PT-20.3.71 performed 2/11/88 on V17 PT-20.3.70 performed 2/11/88 on V16 Relevant information and findings obtained by the inspectors:

(1) The inerting line failure was apparently the result of excessive cooling and resultant stress in the line. The excessive cooling was due to the introduction of insufficiently heated nitrogen frem a malfunctioning vaporize (2) S.nd

- did not reach tra Unit 2 vacuum breakers, but rather was found on the inboard side of closed butterfly valves (V16 and V17) located inboard of the vacuum breaker Also, the sand probably did not contribute to the LLRT failures. The butterfly valves are normally closed and sand could not easily get onto the seals. The project engineer stated that the butterfly valves had a past history of excessive leakage and that they had actually leaked less than in past tests. As of March 4, mainten-ance on one valve had resulted in acceptable leakage and the other showed leakage just in excess of the specified limit. The identical Unit 1 valves passed the LLRTs and required no repai (3) The inerting line failure did not occur on January 4, as initially though In order fer the sand to have reached Unit 2 valves V16 and V17, the failura would have had to occur on or before the Unit 2 shutdown (about January 1,1988).

(4) An attempt to repair the uderground inerting line proved unsuccessfu When an identical break area was repaired the

! line continued to exhibit leakage; indicating an additional j break or breaks. Most of the underground line has now been l

bypassed. A temporary replacement line installed above ground is now being use (This is a non-safety-related line.)

(5) The sand that entered the Containment Atmosphere Control System through the break in the inerting line was found at the following locations:

Almost 2 quarts in the Unit i drywell near the 18-inch inerting line opening (underground line was 8-inch size).

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• 810wn into torus of Unit 1;(from 20-inch line).

None found in drywell or torus of Unit ~' - Approximately 2 cups of-sand / water slurry was found in the

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Unit 2 pipe between V16 and V1 *

Approximately 2 gallons of slurry was found between Unit i valves V16 and V1 *

About a handful of dry sand near V5 (downstream side) on Unit The inspectors had no further questions. This' item is close (OPEN) Inspector Followup Item 325/88-05-04 and 324/88-05-04, Hydrogen Leak in Turbine Building Pipe Tunnel. The inspectors "

reviewed Operating Experience Report No.88-024 and conducted an inspection of the rad-waste pipe tunnel and the electrical tunnel located above the pipe tunnel. The inspectors determined that the pre-installation pressure test requirements had been modified to preclude pressure testing of valves on their back seat, that temporary hydrogen monitors will be installeo above the subject hydrogen valves per SP-88-01 (not yet accomplished as of March 29),

and that routine system checks for hydrogen leaks were being conducte The potential for hydrogen intrusion into safety-related equipment I areas does not represent a hazard in the pipe tunnel as currently configured. This item remains open pending the inspector's review of other possible hydrogen interactions with safety-related equipmen One potentially significant safety matter regarding operator inattention during plant shutdown conditions was identified. Three violations, and no deviations were identifie . Plant Modifications - Unit 2(37700)

The inspector reviewed PM-87-128, Weld Overlay Repair, Field Revisions 30 and 31, Repair of CRD Line Leaks, to verify compliance with ENP-03, Re , Plant Modification Procedure. The inspector observed the pipe replace-ment and repair, which used cryogenic couplings to connect the new pip The inspector also reviewed purchase ordars and specifications related to the cryogenic fittings as well as ALARA ;we-job reviews conducted for the job. Personnel interviews of selected eng'neering ALARA personnel were also conducte . . - _ - - - - - - - - - - - - - - -

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. t The licensee replaced the CRD lines after leaks were found on insert and withdraw lines inside containment during the reactor coolint system hydrostatic tes Small weeping leaks were found on 5 lines on March 19, 1988. Additional CRD line leaks were found on subsequent inspections; all leaks were repaired. A-brown discoloration on the lines was found, by i CP&L's laboratory, to contain chlorides and sulfu Two regional inspectors reviewed the licensee's metallurgical investigation at the lab

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(see report 325,324/88-17). The inspector will continue to follow the licensee's resolution as cart of the LER closecu .

No violations or deviations were identifie . Allegation No. RII-88-A-8011 This allegation was received by the Senior Resident Inspector on February 24, 1988, and later, on February 29, was repeated by the alleger to the Region II Allegation Coordinator in Atlanta, Georgie. On March 23, an inspector conducted an announced inspection of the allegation which .

involved a Shift Operations Supervisor allegedly not being fit for dut The inspector reviewed several pertinent documents to include the Quality ,

, Check Report No. B-88-01-10/QCR No.14221, and met with the following individual C. Dietz - Plant General Manager A. Bishop - Plant Operations Manager S. Choplin - Director of Personnel Relations

E. Enzor - Director of Regulatory Compliance W. Ruland - Senior USNRC Resident Inspector There is no USNRC regulation governing Fitness for Duty. However, the licensae's policy, Drug and Alcohol Abuse Statement of Practice, states...

"The use of alcoholic beverages by an employee on or away from Company 1 property that may adversely affect the employee's job performance, or that may n eflect unfavorably upon public or governmental confidence in the manner in which the Company carries out its responsibilities may result in disciplinary action, including possible termination".

• ha inspector learned that a similiar allegation had been anonymously telephoned to the licensee's Quality Check Program during the evening 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> of January 25. Within a few hours after receipt, the Plant General i Manager instructed his operational staff to conduct an immediate investi-gation which included interviews of operation supervisors who had close contact with the Shift Operations Supervisor during the shift in questio The licensee concluded the allegation was without merit.

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Based upon interviews and a review of documentation the inspector was ( unable to substantiate the allegation and, given the absence of an NRC ,

l regulation governing Fitness for Duty, found the licensee to have been

responsive to this allegatio '

i No significant safety matters, violations, or deviations were identifie <

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11. Unusual Event / Fire (93702)

The licensee declared an Unusual Event for two minutes on March 9, 1988, at 2:40 p.m., after a fire could not be confirmed out within 10 minute The drywell chiller power cables (non-safety) on the Radwaste Building roof were reported on fire at 2:27 p.m. The cable insulation had chafed on the roof edge. The fire was announced at 2:28 p.m., and the fire brigade was dispatched. The cables were de-energized at 2:31 p.m., and the fire reported out at 2:36 p.m., by an I&C technician. Since the shift fire brigade commander did not report the fire out vithin 10 minutes, the licensee declared an Unusual Event per Emergency Frxecure PEP-2.1, Revision 21, Initial Actions, which requires the declaration of an Unusual Event with a fire within the protected area lasting more than ten minute The shif t fire coannander reported the fire out at 2:42 p.m. No safety-related equipment was affected by the fir Based on interviews with plant personnel, record and log review, the inspector concluded that the licensee complied with their emergency procedure No significant safety matters, violations, or deviations were identifie . List of Abbreviations for Unit 1 and 2 ALARA As Low As Reasonably Achievable A0 Auxiliary Operator APRM Average Power Range Monitor BWROGS Boiling Water Reactor Owners Group BSEP Brunswick Steam Electric Plant CAC Containment Atmospheric Control CAD Containment Atmospheric Dilution CCW Closed Cooling Water CRD Control Rod Drive CST Condensate Storage Tank DC Direct Current DG Diesel Generator DR Direct Replacement EER Engineering Evaluation Report ENP Engineering Procedure EPA Electrical Protection Assembly ESF Engineered Safety Feature F Degrees Fahrenheit FSAR Final Safety Analysis Report GPM Gallons Per Minute HP Health Physics HPCI High Pressure Coolant Injection HX Heat Exchanger I&C Instrumentation and Control IE NRC Office of Inspection and Enforcement IFI Inspector Followup Item JC0 Justification for Continued Operation LER Licensee Event Report

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LLRT Local Leak Rate Test {

LOCA Loss of Coolant Accident  ;

MG Motor Generator MST Maintenance Surveillance Test NRC Nuclear Regulatory Comission NSW Nuclear Service Water OER Operating Experience Report OP Operating Procedure PA Protected Area PI Pressure Indicator l PM Plant Modification  ;

PNSC Plant Nuclear Safety Committee i PSIG Pounds per Square Inch Gauge PT Procedure Test QA Quality Assurance  !

QC Quality Control  !

RB Reactor Building RCIC Reactor Core Isolation Cooling RHR Residual Heat Removal '

RPM Revolutions Per Minute l RPS Reactor Protection System RWCU Reactor Water Cleanup i SP Special Procedure l SRO Senior Reactor Operator

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STA Shif t Technical Advisor SW Service Water TC/ Thermostatic Control Valve TS Technical Specification URI Unresolved Item USNRC United States Nuclear Regulatory Comission V Volt VID Violation WR/JO Work Request / Job Order

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