IR 05000334/1985024

Insp Rept 50-334/85-24 on 851101-30.No Violation Noted.Major Areas Inspected:Plant Operations,Fire Protection, Radiological Controls,Physical Security,Esfs Verification, Surveillance Activities & Review of LERs

ML20138Q785
Person / Time
Site:	Beaver Valley
Issue date:	12/17/1985
From:	Asars A, Lester Tripp, Troskoski W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20138Q784	List: IR 05000334/1985024 ML20138Q782
References
50-334-85-24, IEB-85-001, IEB-85-002, IEB-85-1, IEB-85-2, NUDOCS 8512300002
Download: ML20138Q785 (13)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /85-24 Docket N ; Licensee: Duquesne Light Company One Oxford Center 301 Grant Street-Pittsburgh, PA 15279 Facility Name: Beaver Valley Power Station, Unit 1 Location: Shippingport, Pennsylvania Dates: N ember 30, 1985 7 nspector: W . .

M /k7!8f Date Avvk. M. Tr dkbski, Senior Resident Inspector h)4.A.AsaM,ResidentInspector

. t. s izHer Date Approved by . L. E. Trfbb,- Chief, Reactor Projects Section 3A Ik/7[8[

' Da'te Inspection Summary: Inspection No. 50-334/85-24 on Novembec 1-30, 1985 Areas Inspected: Routine inspections by the resident inspectors (117 hours0.00135 days <br />0.0325 hours <br />1.934524e-4 weeks <br />4.45185e-5 months <br />) of i

licensee actions on previous inspection findings, plant operations, housekeeping, fire protection, radiological controls, physical security, engineered safety features verification, surveillance activities, cleanup of Unit 2 contaminated areas, IE Bulletin followup, and review of licensee event report Results: No violations were identified. Potentially significant items identified

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were main feedwater valve reliability problems that resulted in the introduction of loose parts to the main feedwater lines of the A and C steam generators (dis-cussed in Details 3.b.3 and 4), and the one-time failure of the C charging pump discharge check valve to close when required (Detail 5).

I 8312300002 851218 ' ' ' *- '

PDR ADOCK 05000334  !

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J TABLE OF CONTENTS Page

' Persons Contacted.................................................... 1 Followup on Outstanding Items........................................ 1 Plant Operations..................................................... 2 t' Genera 1.................~........................................ 2 Operations...................................................... 3 Plant Security / Physical Protection.............................. 4 Radiation Controls.............................................. 5 Plant Housekeeping and Fire Protection....................'.'.,.... 5 Engineered Safety Features.(ESF) Verification.....................'... 5 Surveillance Activities.............................................. 6

' Cleanup of Unit 2 Contaminated Areas................................. 7

. IE Bulletins......................................................... 8

~ .Inoffice Review of Licensee Event-Reports (LERs)..................... 10

/ Exit Interview.............~................................

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DETAILS Persons Contacted J. J. Carey, Vice President, Nuclear Group R. J. Druga, Manager, Technical Services T. D. Jones, General Manager, Nuclear Operations

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W. S. Lacey, Plant Manager J. D. Sieber, General Manager, Nuclear Services N. R. Tonet, General Manager, Nuclear Engr. & Constr. Unit The inspector also contactod other licensee employees and contractors during this inspectio . Followup on Outstanding Items The NRC Outstanding Items (OI) List was reviewed with cognizant licensee per-sonnel. Items selected by the inspector were subsequently reviewed through discussions with licensee personnel, documentation reviews and field inspec-tion to determine whether licensee actions specified in the OIs had been satisfactorily completed. The overall status of previously identified in-spection findings were reviewed, and planned and completed licensee actions were discussed for those items reported below:

(Closed) IFI (85-22-03): Cleanup activities of. contaminated areas of Unit 2 which resulted from the October 31, 1985, spill of radioactive liquid is ciscussed in detail 6 of this report.

1 (Closed) IFI (85-17-03): Special report due per T3 6.9.2, regarding inoper-able rad monitor RM-MS-100 The licensee issued Revision 1 to this special report on November 12, 198 The cause of the inoperable main steam monitor was determined to be a loose triaxial connector which was replaced. Other similar rad monitors were also checked and no further problems were note This item is close (Closed) Unresolved Item (85-22-02): Determine the amount of activity re-leased during the Unit 1 - Unit 2 spill of October 31, 1985. The licensee obtained a sample of the liquid discharged from the boron hold tank at the uncapped line outside the Unit 2 condensate polishing building during the event. Analysis identified Cesium 134, 137, Colbalt 58 and 60 as the primary radionuclides present. An estimated 300 gallons, initially diluted with about 700 gallons of residual sump water, was pumped to the storm sewer and eventu-ally to the rive This resulted in a potential body dose of 0.47 mrem at the site boundary, which represents about 33% of the quarterly limit defined in Technical Specification 3.11.1.2. The concentration as compared with the

' ' maximum permissible concentration (MPC, defined in 10 CFR 20.106) of the 1000 gallons, undiluted by storm sewer flow, was calculatec to be about 45, ex-cluding tritiu Total activity excluding tritium was 2.24 E-3 C1, and when averaged over one hour (10 CFR 20.106a) including dilution represents about 0.77 MPC. A special report was issued per TS 3.11.1.1 on November 29, 1985, detailing the releas It noted that the total concentration including tritium

.for the five minute discharge (with dilution) to the Ohio River exceeded the l

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maximum permissible concentration by a factor of 3 This is a violation of TS 3.11.1.1 that will~not be cited because it is the result of the un-r authorized. operation of a' Unit 1/2 system boundary isolation valve that re-sulted in violation'(85-22-01). The inspectors had no further concerns and

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.this item is close '

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'% ~ Plant Operations

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$ ' ' General Inspection to.urs'of the plant areas listed below were conducted during both day.an'd' night shifts.with respect to Technical Specification (TS)

, -compliance, housekeeping and cleanliness, fire protection, radiation 1

- control,-phyt.ical security and plant protection, operational and main-l tenance administrative control Control Room

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~ Primary Auxiliary Building

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Turbine Building

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-Service Building-

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Main Intake Structure

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Main Steam Valve Room

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. Purge Duct Room

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East / West Cable Vaults-

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Emergency Diesel Generator Rooms

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Containment Building E -- - Penetration Areas

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Safeguards Areas

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Various'Switchgear Rooms / Cable Spreading Room

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Protected Areas Acceptance criteria for the above areas included-the following:

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.BVPS FSAR

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Technical Specifications (TS)-

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BVPS Operating Manual. (OM), Chapter 48, Conduct of Operations

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OM 1.48.5, Section D, Jumpers and Lifted Leads'

p, '-- ,0M 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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0M Chapter SSA, Periodic Checks, Operat1ng Surveillance Tests

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

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BVPS'Radcon Manual-(RCM)

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

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BVPS Site / Station. Administrative Procedures (SAP)

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.BVPS Physical Security Plan (PSP)

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Inspector' Judgement-

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l b. Operations The inspector toured the Control Room regularly to verify compliance with NRC requirements and facility technical specifications (TS). Direct ob-servations of instrumentation, recorder traces and control panels were made for items important to safet Included in the reviews are the rod position indicators, nuclear instrumentation systems, radiation monitors, containment pressure and temperature parameters, onsite/offsite emergency power sources, availability of reactor protection systems and proper alignment of engineered safety feature systems. Where an abnormal con-dition existed (such as out-of-service equipment), adherence to appro-priate 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 temporary operating procedures were reviewed on a sampling basis for compliance with technical specifications and those administrative con-trols listed in Paragraph 3 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 condition The inspector verified adherence to approved procedures for ongoing ac-tivities observed. Shift turnovers were witnessed and staffing require-ments confirme Except where noted below, inspector comments or ques-tions resulting from these daily reviews were acceptably resolved by licensee personne (1) The outer containment airlock door failed its six month leak rate test on October 31, 1985. The licensee initiated a hold on con-tainment entries per TS 3.6.1.3. After replacing the 0-rings and adjusting the breech, BVT 1.3-1.47.8, 'ersonnel Airlock Type B Leak Rate Test, was successfully completed n November 1, 1985. The inspector reviewed the test data calcu ktions which indicated that the leak rate was 196.4 scfd. This is within the acceptance cri-teria of 324.5 scf (2) A load reduction to about 30% r,awer was initiated on November 1, 1985, to allow maintenance on the A and C main feed-reg valve During the load reduction, a containment entry was made after the Type B test to repair steam generator low point drain valve FW-60 This valve had been responsible for an increase in the containment sump pumpout rate. After repair, the pumpout rate dropped from 14 gph to a nominal value of 6 gp The reactor was placed in Mode 3 (hot standby) at 11:45 p.m. on November 2, 1985, after excessive leakage of hot feedwater through the clearance points prevented mechanics from disassembling the feed-reg valves while on bypass flow. With the area temperature limited to a workable range for the mechanics, the reactor was re-

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turned to Mode 2 (less than 5% thermal power) at 6:25 a.m. on November 3, 1985.

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, Inspection of the A and C-feed-reg valves revealed that the valve

plugs anti-rotation devices.were missing. The anti-rotation device

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remained intact for the B valve. These loose parts consist of a 2" x 1/2" bolt and two nuts, which are assumed to be in either

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(1) the:feedwater line elbow' located at the vertical rise just in-side containment,.(2) the feedwater ring, or (3) inside the steam

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. generators. .The cause of the failure was attributed to fatigue of the individual tabs of the one-sixteenth inch thick lock washers.

U -The licensee redesigned the lock washers using one-eighth inch thick  ;

t stainless steel' material and replaced all thr,ee anti-rotation de-

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. A review of the loose parts safety analysis prepared by Westing-

! house', the NSSS vendor,-indicated that the missing material did not

present an immediate problem. The analysis did recommend retrieval

! of the loose parts at the earliest opportunity, not to exceed startup from the next. refueling outage scheduled for May, 198 '

Review of; licensee actions to retrieve those parts and identifica-

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tion of any damage to the pipe elbow, feedwater ring, or steam

generator tubes is Unresolved Item (85-24-01).

(3) Control room operators rapidly. reduced reactor power to about 25%

on November 9, 1985, due to B steam generator feedwater control ,

l problems. . Investigation revealed that the feed-reg valve stem

. threads that attach.to the valve plug had worn smooth enough to

~ disengage. The valve stem was replaced. Maintenance engineers i- informed the inspectors that one'of the other two feed-reg valve i stems had been replaced with a new assembly on November 3, 198 .

!- An inspection of the other stem identified no problems. When the

! B valve was disassembled the first time to check its anti-rotation i device, no inspection of.its stem was conducted. The licensee has been in contact with Copes-Vulcan, the valve vendor, to determine what actions can be-taken to improve the reliability of these 8 inch

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feedwater' valves. Further review is. Inspector Follow Item'(85-24-

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I Implementation of the Physical Security Plan was observed in the areas listed in paragraph 3a above with regard to the following:

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Protected area barriers were not degraded;

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' Persons and packages were checked prior to allowing entry into the LProtected Area;.

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. Vehicles were properlyisearched and vehicle access to the Protected L Area was.in accordance with approved procedures;

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Inspection of the A and C feed-reg valves revealed that the valve plugs anti-rotation devices were missing. The anti-rotation device remained intact for the B valve. These loose parts consist of a 2" x 1/2" bolt and two nuts, which are assumed to be in either (1).the feedwater line elbow located at the vertical rise just in-side containment, (2) the feedwater ring, or (3) inside the steam generator The cause of the failure was attributed to fatigue of the individual tabs of the one-sixteenth inch thick lock washer The licensee redesigned the lock washers using one-eighth inch thick stainless steel material and replaced all thr,ee anti-rotation de-vice A review of the loose parts safety analysis prepared by Westing-house, the NSSS vendor, indicated that the missing material did not present an immediate problem. The analysis did recommend retrieval of the loose parts at the earliest opportunity, not to exceed startup from the next refueling outage scheduled for May, 198 Review of licensee actions to retrieve those parts and identifica-tion of any damage.to the pipe elbow, feedwater ring, or steam generator tubes is Unresolved Item (85-24-01).

(3) Control room operators rapidly reduced reactor power to about 25%

on November 9, 1985, due to B steam generator.feedwater control problems. . Investigation revealed that the feed-reg valve stem threads that attach to the valve plug had worn smooth enough to disengage. The valve stem was replaced. ; Maintenance engineers informed the inspectors that one of;the other two feed reg valve stems had been replaced with a new assembly-on November 3, 198 An inspection of the other stem identified no problems. When the B valve was disassembled the first- time to check its anti-rotation device, no inspection of its stem was conducted. .The licensee has p been.in contact with Copes-Vulcan, the valve vendor, to determine what_ actions can be taken.to improve the reliability of these 8 inch feedwater valves. Further review is Inspector Follow Item (85-24- '

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c. Plant Security / Physical 1 Protection Implementation of the Physical Security Plan was-observed in the areas listed in paragraph 3a above with regard to the following:

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Protected area barriers were not degraded;

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Isolation zones were clear;

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Persons and' packages were checked prior to allowing entry into the

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

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

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

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Adequate lighting was maintaine No discrepancies were observe A Regulatory Effectiveness Review of the plant physical security was performed in April, 1985. The results, which identified several safe-guards inadequacies, program concerns and general observations were for-warded to DLC by letter dated August 29, 1985. The licensee responded to these items in a letter dated November 13, 1985, and committed to complete corrective actions by July, 1986. Verification that those com-mitments are implemented, is Inspector Follow item (85-24-03). Radiation Controls Radiation controls, including posting of radiation areas, the conditions of step-off pads, disposal of protective clothing, completion of Radi-ation Work Permits, compliance with the conditions of the Radiation Work Permits, personnel monitoring devices being worn, cleanliness of work areas, radiation control job coverage, area monitor operability (portable and permanent), area monitor calibration and personnel frisking procedures were observed on a sampling basi The licensee informed the inspector that an error had been discovered 3 in the method used to calculate the gaseous waste dose contained in the semi-annual report. The corrected values are less than 1% of the tech-nical specification limit The licensee's representative stated that

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a calculation package would be prepared to formally document the errors

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and corrections. Review of this package and verification that the pre-viously reported values are updated in the next semi-annual report, is Inspector Follow Item (85-24-04). Plant Housekeeping and Fire Protection Plant housekeeping conditions including general cleanliness conditions and control of material to prevent fire hazards were observed in areas listed in Paragraph 3a. Maintenance of fire barriers, and fire barrier penetrations and verification of posted fire watches in these areas were also observe . Engineered Safety Features (ESF) Verification The operability of the Low Head Safety Injection System was verified on Novem-ber 14, 1985, by performing a walkdown of accessible portions that included the following as appropriate:

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(1) System lineup procedures matched plant drawings and the as-built con-figuratio (2) Equipment conditions were observed for-items which might degrade perfor-manc Hangers and supports were operabl (3) The interior of breakers, electrical and instrumentation cabinets were inspected for debris, loose material, jumpers, et (4) Instrumentation was properly valved in and functioning; and had current calibration date (5) Valves were verified to be in the proper position with power availabl Valve locking mechanisms were checked, where require No deficiencies were identifie . Surveillance Activities To ascertain that surveillance of safety-related systems or components is being conducted in accordance with license requirements, the inspector ob-served portions of selected tests to verify that: The surveillance test procedure conforms to technical specification requirements,

' Required administrative approvals and tagouts are obtained before initi-ating the tes . Testing is being accomplished by qualified personnel in accordance with an approved test procedur Required test instrumentation is calibrate LCOs are met, The test data are accurate and complete. Selected test result data was independently reviewed to verify accurac Independently verify the system was properly returned to servic Test results meet technical specification. requirements and test discre-pancies are rectifie . The surveillance test was completed at the required frequenc The following in progress tests were witnessed by the inspector:

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OST~1.36.1, Diesel Generator No. 1 Monthly Test, November 6, 198 OST 1.11.2, Safety Injection Pump Test (SI-P-1B), November 6, 198 i l

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OST 1.7.6, Charging Pump 1C Test, November 14, 198 OST 1.11.1, Safety Injection Pump Test (SI-P-1A), November 20, 198 MSP 6.22 & 6.40, Monthly Test and 18 Month Calibration of Channel III Overtemperature and Delta Temperature, November 20 - 21, 198 After an extended outage for modification and alignment, the C charging pump was returned to service on November 14, 1985, and left in continual service until the next charging pumps monthly surveillance test came due on November 19, 1985. When CH-P-1C was shutdown after startup of the second pump, RCP seal injection was los Control room personnel m diately restarted CH-P-1C to restore seal injection. The operator at the pump indicated that no re-verse rotation was observed during the several seconds it was off; the pump was still coasting down. It was determined that the discharge check valve, CH-24, failed to properly seat. The pump was secured a second time without further loss of seal injection flow, indicating proper seatin The discharge valve is a 3 inch Velan swing check valve that had been modified after a failure resulted in an overpressurization of the charging pump suction line in 1981 (See NRC Inspection Report 50-334/81-08, Detail 7, and Immediate Action Letter 81-16, dated March 27, 1981, for additional information). It represents a boundary between the high pressure piping (2500 psig) and the lev pressure suction line (275 psig). The operations supervisor informed the inspector that the valve would be disassembled and inspected during the fifth refueling outage provided no further problems occurred. Review of the results of this inspection is Unresolved Item (85-24-05).

6. Cleanup of Unit 2 Contaminated Areas On November 1, 1985, the licensee began removing the radioactive water from the Condensate Polishing Building (CPB) sump which had collected there as a result of the release from the Unit 1 Boric Acid Hold Tank (BR-TK-7) on Octo-ber 31, 1985 (see Inspection Report 85-22, Detail 3.b.5). The waste was pumped, per TOP 85-03: Transfer of Unit 2 Spill to Unit 1, and RCH Chapter 1, Procedure F, Use of Temporary Hoses / Piping, to the Unit 1 Fuel Building (FB) sump via a temporary hos From there, it was transferred to low level waste drain tanks (LW-TK-3A, 3B), as liquid wact The transfer of water continued until the morning of November 2,1985, at which time LW-TK-3A and 3B became full. All sump pumps were secured and the contents of LW-TK-3A and 3B were pumped to steam generator drain tank 7B (LW-TK-78). Later that afternoon a high level alarm was received for the tunnel sump and the sump pump was started to transfer the water to LW-TK-3A and 3 Check valve DA-207 on the fuel building sump pump 2A failed, creating a flow-path from the tunnel sump to the FB sump due to the temporary alignment de-scribed above. The FB sump filled and overflowed about 130 gallons of water ont the floo It went unnoticed by the operators due to a failure of the sump hign level alarm, until discovery at approximately 4:00 p.m. Operators investigated and subsequently started DA-P-2A and 2B at 4:10 p.m. and pumped

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i I the excess water into LW-TK-3A and 3 The FB sump high level alarm was

! satisfactorily tested and declared operable but as an added precaution, an operator was posted at the FB sump to monitor level during all pumping acti-

vities involving the sump.

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In both. instances, Unit 2 CPB and Unit 1 FB spills, there were no internal-uptakes of radioactivity. Only two Unit 2 personnel received minor contami-nation on shoes and pant cuffs during the Unit 2 CPB spill. During the Unit

! 2 CPB release, Radcon sampled the water coming from the 1.5 inch uncapped lin The gross activity of the water was 2E-3 micro.Ci/ml. Area air samples showed

{ a gross activity of IE-11 micro C1/ml, indicating that there was no airborne

, radioactivit Radiation levels in the CP8 (elevation 722') hallway ranged

.from 2 mr/hr to 0.1 mr/hr in the area between the Unit 2 end of the pipe

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trench and the furthest point affected by the water. Approximately 60 pounds of dirt was washed from the pipe trench into the hallway by the spill. Sur-j veys indicated that it was reading about 12 mr/hr on contact. The contami-

! nated dirt was transferred to solid waste for processing. Surface dirt out-side the CPB around the pipe trench was sampled and showed a gross activity

,,' of SE-3 micro Ci/ml.

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Unit 2 CPB decontamination efforts included removal of radioactive water,.CPB

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sump pumps, and assessment and cleaning of affected area Radcon personnel determined the affected areas to consist of portions of CPB (elevation 722')

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floor, all drains on and lower than elevation 722' which empty into the CPB sump, the CPB sump, the pipe trench between units and.the dirt and debris it contains, and the affected surface dirt outside the CPB. Radcon personnel

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mine if an area was fully decontaminated were (1) less than 450 pci/100 cm(2)

swipe or (2) less than 100 cpm above background.

r Radcon personnel determined the general area radiation levels of the FB (ele-vation 735') to be about 0.2 mr/hr; with the " hot spot" being a floor drain at 0.5 mr/hr (this drain goes to the tunnel sump). The FB has railroad tracks

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which-lead outside. These tracks and a pudule between them also filled with j water from the FB sump.but the water remained'within approximately 10 feet j

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of the FB. ~A grab. sample indicated a gross activity of 8.8E-5 micro Ci/m Decontamination of.the FB was completed satisfactorily using squeeges and rags.

L Cleanliness criteria were the same as'those;for Unit 2 CPB. .The inspectors closely followed the cleanup process, and had no additional concerns.

j' IE Bulletins

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j IE Bulletin 85-01: Steam Binding of' Auxiliary Feed Pumps

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i This bulletin addresses'the possible-inoperability'of.the Auxiliary j- Feedwater (AFW) pumps as a result of steam binding. Information Notice i 84-06 also dealt'with this problem (see Inspection Report 85-12, Detail '

{- 6.b). Steam binding of the AFW pumps is caused.by leakage'from the main

[ feedwater system past several check valves to the pump discharga and t

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possibly the pump suction. This is of particular concern if the AFW ,

pumps either discharge into common piping to feed the steam generators I or if the pumps take suction from a common header. At BVPS, Unit 1, this is not the case; each AFW pump has independent piping on its discharge and suction. Additionally, licensee personnel routinely check and log

.the temperatures of the AFW pipin The inspectors had no further concern IE Bulletin 85-02; Undervoltage Trip Attachments of Westinghouse D8-50 Type Reactor Trip Breaker This bulletin was issued to inform licensees of recent DB-50 reactor trip breaker (RTB) reliability problems. It also required those plants cur-rently operating with DB-50 RTBs that have not yet installed automatic shunt trip devices to perform an undervoltage trip attachment (UVTA)

force margin test. Beaver Valley, Unit 1, is currently scheduled to make those modifications during the next refueling outage, in May, 1986, and is one of three plants falling into this categor Required actions and station responses were as follows:

(1) Test the UVTA of each RTB in service to determine that adequate force margin exists by attaching a 20 oz. weight to the trip ba This test is required to be performed with the breaker in the as-found condition prior to any lubrication or other maintenance and must be successfully performed three times in successio The inspectors observed testing of the reactor trip bypass breakers on November 8, 1985, and RTBs on November 12, 198 The licensee used a 19 oz. weight, traceable to the NB Discussions with the bulletin's technical contact indicated that this was acceptabl Each of the four 0B-50 breakers was successfully tested three times without incident per one-time modified preventive maintenance pro-cedure (2) The monthly surveillance test procedure for the reactor protection system is required to be modified to add the UVTA force margin test prior to any lubrication or adjustment of the VVTA. If a RTB fails the force margin *.est, the redundant breaker is to be similarly tested within eight hours. This procedure modification is to remain in effect until the automatic shunt trip modification is implemente The licensee revised the maintenance surveillance procedures (MSP)

used to test the reactor protection system logic -trip breakers and bypass breakers, to incorporate the above requirements. On November 26, 1985, the inspector observed the bi-monthly check of RTB B L.id bypass breaker A per those MSPs. No problems were observe ,

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(3) The plant operating staff is to be provided written instruction containing the content of this bulletin, to be reviewed by each licensed operator at the start of their next duty shift. Addition-ally, an RTB is to be declared inoperable if the UVTA either does not successfully pass the test or otherwise, may not be capable of

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performing its inteaded safety functio Through discussians with shift personnel, the inspector determined that the currest operating staff was informed of the bulletin's in-

, formation and requirements. Most non-control room licensed person-nel had alsr been informed as evidenced by completed training ros-ter Standing orders were issued to ensure remaining operators completed the reading prior to assuming shift duties. Licensee action is satisfactor (4) The NRC is to be notified via the emergency notification system within four hours of any RTB being declared inoperabl Since this is a new reporting condition, the Shift Reporting Matrix was appropriately update No other actions were require . Inoffice 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 de-scription of cause and adequacy of corrective action. The inspector deter-mined whether further information was required from the licensee, whether generic implications were indicated, and whether the event warranted onsite followup. The following LERs were reviewed:

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LER: 85-18 Vital Bus III Inverter Input Fuse Failure Causing Reactor Tri LER: 85-19 Reactor Trip Due to Loop Flow Instrumentation Spik The actual event leading to LER: 85-18 was discussed in detail 3.b.1 of NRC Inspection Report 334/85-22. Since those ' initial discussions with the I&C supervisor concerning the root cause, the licensee has reported that it is now believed to be attributable to high ambient temperature in the switchgear rooms which caused the control rectifier to misfire. Due to previous tem-perature related problems in this area, a station modification is being con-sidered to improve ventilation in this area (Inspector Follow Item 85-24-07).

LER: 85-19 reported a reactor trip due to a loop flow instrument spike. The spike was caused by a low flow signal due to a procedure deficiency for repair and calibration of loop flow transmitters. This resulted in a reactor trip from 100% power on October 25, 1985 (see detail 3.b.3 of Inspection Report 334/85-22). The licensee has' initiated RCS loop flow surveillance procedural changes to correct the valve identification discrepancy before March 198 Incorporation of these procedure changes will be tracked as Inspector Follow Item (85-24-06).

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Exit Interview Meetings were held with senior facility management periodically during the course of this inspection to discuss the inspection scope and findings. A

, summary of inspection findings was further discussed with the licensee at the l conclusion of the report period.

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e 4 - - y w n , e --r,- --,w-.-r-- -,w---.c m > qu+ .,--w,- 4 y y r3,-rw , e p - - - - - - + , - - - - ~,