IR 05000352/1988015

Insp Rept 50-352/88-15 on 880601-0705.No Violations Noted. Major Areas Inspected:Plant Tours,Observations of Maint & Surveillance,Review of LERs & Periodic Repts,Review of Operational Events & Sys Walkdowns

ML20151H885
Person / Time
Site:	Limerick
Issue date:	07/14/1988
From:	Linville J, Williams J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20151H869	List: IR 05000352/1988015 ML20151H866
References
50-352-88-15, GL-82-33, IEB-88-004, IEB-88-4, IEIN-87-059, IEIN-87-59, NUDOCS 8808020109
Download: ML20151H885 (12)

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U.S. NUCLEAR REGULATORY COMMISSION f

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REGION I

Report No. 88-15 1 i

Docket No, 50-352 '

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License No. NPF-39 <

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Licensee: Philadelphia Electric Company ,

2301 Market Street '

Philadelphia, Pa 19101 j Facility Name: Limerick Generating Station, Unit 1 I Inspection Period: June 1 - July 5, 1933

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1 Inspectors: T. J. Kenny, Senior Resident Inspector L. L. Scholl, Resident Inspector '

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Reviewed by: - / // D h. Wil % s, Project Engineer 7 '// k Date

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-- i Approved by: 0Af?fs e /) 7 Lf V

-' ATates Linvil Nhiefifr6Jects Section 2A

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Sum na ry: Iac/ tine daytime (135 h s) and backshift/ holiday (29 hours3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br />)

! inspections of Unit I by the resident inspectors consisting of (a) plant

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tours. (b) observations of maintenance and surveillance, (c) review of LERs

and periodic reports, (d) review of operational events and (e) system '

i walkdowns, j

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Operated the plant at 85 to 90*4 power while monitoring the I previously identified fuel leakage.

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Submitted several LER$ (section 6.0), the monthly operating report !

(section 5,0), j

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Conducted a Reactor Core Isolation Cooling System maintenan !

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Replaced defective motor control center bus bar !

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Perfortred minimum flow tests on core spray pump l

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Corrected feedi.ater hester level control problem {

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j DETAILS l

l i 1.0 Persons Contacted l 4 ,

! Within this report period, interviews and discussions were conducted with I members of licensee management and staff as necessary to support j inspection activit ,

2.0 Operational Safety verification (71707, 70709, 71710 and 71881)

2.1 Documents Reviewed l -

Selected Operators' Logs

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Shift Superintendent's Log i l -

Texporary Circuit Alteration tog  ;

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Radioactive Waste Release Permits (liquid and gaseous)  !

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j Selected Radiation Work Permits (RWP) (

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Selected Chemistry Logs i

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Selected Tagouts [

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heatth Physics Log  !

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2.? The inspector conducted routine entries into the protected areas of i j the plant, including the control room, reactoe enclosure, fuel t i floor, and drywell (when access is possible). During the  !

i inspection, discussions were held with operators, technicians (HP & >

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I&C), mechanics, security personnel, supervisors and plant f managemen The inspections were conducted in accordance with NRC !

Inspection Procedures 71707, 71709, 71710 and 71831 and affireed the

, licensee's comitments sad compliance with 10 CFR, Technical Specifications, License Conditions and Administrative Procedure ,

l No violations were identifie {

{ l 2. EngineeredSafetyFat1reJESF)SystemWalkdown:

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The inspectors verified the ^perability of the selected ;

ESF system by performing a walkdown of portions of the j

, system to confirm that system lineup procedures match

plant drawings and the as-built configuratio This ESF i

] system walkdown was also conducted to identify equipment

conditions 1, hat might degrade performance, to determine that instrumvntation is calibrated and functioning. and to

{ verify that valves are properly positioned and locked as appropriate. The inspectors also utili:ed methods prescrioed in a study prepared for the NRC by Brookhaven

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National Laboratory using the Limerick Probabilistic Risk

Assessment (PRA). The study, entitled PRA-Based System Inspection Plan, dated May 1936, proYides inspection l guidance by prioritizing plant safety systems with respect I

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! to their importance to risk. An abbreviateo system

checklist on Table 11-3 which identifies c.omponents that i are consicered to have a high contribution to risk as t j determined in the PRA was also used. Accessible portions of the standby liquid control systen, were inspecte The following procedures, drawings and tests were also I reviewed

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S48. Standby Liquid Control System Set-up for I Normal Operation

$48.9 A Routine Inspection of Standby Liquid  ;

Control System [

548,1.A (COL) Equipment Alignment to Place Standoy Liquid *

Control System in Nermal "Standby" '

Condition  !

IV 48 Instrumentatter. Valve List L ST-5-045-800-1 SBLC Sodium Pentaborate Concentration l Analysis t Orawing M-48 P&!D Standby Liquid Control i Procedure A-8 Locked Valve List  :

The inspector also reviewed the tank chemistry analysis L results and verified they were in compliance with i Technical Specification 4.1.5 for frequency and  !

specification. Technical Specification surveillance l

requirements for tank and piping temperature monitoring i and for tank level were also verifie [

The inspector identified a discrepancy that several l instrument root valves di& not appear on the system check  !

off list or instrument lineu The licensee iraediately i verified the valves were in the correct position and added L them to the instru ent valve lis l l

The inspector had no further questicas concerning this j system, l

2.3 Inspector Coments/Findinn (93702)  ;

The inspector selected aspects of the unit's operation to determine compliance with the NRC's regulations. The inspector determined that the areas inspected and the licensee's actions did not constitute a health and safety hazard to the public or plant personnel. The following are noteworthy areas the inspector researched in depth:

On June 1, at 7:58 p.m., during a severe thunder storm, a spike occurred on 'D' channel of the control room chlorine detector causing a control room ventilation isolation. All other channels of the detection system were norma l The licensee determined that the alarm was spuricus, reset the tripped condition and returned the

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systems to normal. The licensee notified the NRC of this event on !

the Emergency Notification System (ENS) as require !

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On June 7, the licensee reduced power from 904 to 85% in order to !

further reduct power in the fuel bundles which may be susceptible to !

crud induced localized corrosion (CILC) failures. Recommended  !

maximum power levels provided by General Electric are bar.ed on operating experience at other plants. When operated at power levels [

in accordance with these recommendations CILC induced failures have l not been experienced. The Si power reduction ensures power levels !

are in accordance with the guidelines provided by General Electri !

At 11:45 a.m., on June 9, while performing a special test on 0-13 Emergen:.y Diesel Generator, the diesel inadvertently started. The licensee was parforming a special test to evaluate a slow start capability of the diesel and during the installation of special test :

aquipment the diesel started when a control circuit tsrminal block i connection was loosene Subsequent licensee investigation of the event also revealed a faulty tachometer (which initiates auxiliary l start features as the diesel accelerates to speed). During the i replacement of the tachometer the instrument technician was (

demonstrating the operation of the tachometer relay to the QC l inspector and was unaware that thf operator nad already replaced the '

fuses for the relay. This resultei in an inadvertent start of the !

diesel's associated e ergency service water (ESW) pump. The licensee event riport associated with this event will be reviewed in a future l

recor The NRC was notified of both the diesel and ESW pump start :

on the ENS as require !

On June 10, pcwer was reduced to approximately 70% when problems ;

were experienced with the SC feedwater heater drain valve controller resulting in a high feedwater heater level condition. Control circuit calibrations were nerformed and a defective level switch was reoaired. The heater was returned to service and reactor power was increased to $5% power late on June 11. Several hours later on  :

June 12 the SC feedwater heat.er automatically isolated due to high !

water level. Additional drain valve controller tuning was performed to eliminate excessive level oscillaticns, the heater was returned to l

service and power was returned to 834 by approximately 9:00 a.m. on !

June 13. No further level control prchlems were observe !

t On June 13, the licensee announced it is currently rebidding the {

plant security contracts at both the Peach Bottom and Lieerick '

nuclear power stations with the primary objective of the rebidding to increase the efficiency and the consistency of the cc'apany's nuclear plant security operations by placing both under a single contracto The bids are expected to be awarded by July 8 and the new Limerick contract will begin on October 1. The NRC is monitoring licensee actions in this area for any potential effects on the plant securit )

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On June 20. PECo announced it will implement a new comprehensive drug policy which provides for random testing of all executives and {

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employees granted unescorted access to its nuclear facilities, and l the mandatory termination of anyone found selling, distributing or i using drugs on PECo premise The new policy provides for random !

drug testing at least once during the next year with no prior notification for employees with nuclear unescorted access ch aranc Such clearance allows an employee to be admitted to certain a?eas of i a nuclear plant without escort, j Initial testing will be done between August 1 and September 30, 198 Periodic drug testing will be conducted thereafter with l employees being informed in advance that they will be tested once l during a specific three-month period, but without advance notice of ;

when during the period the test will occu !

t On June 21, at 4:32 a.m., a Unit 1 security guard was observud

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sleeping by two Unit 2 construction watchmen. Security supervision was notified and guard was immediately relieved. Sweeps of the j entire plant were conducted and no discrepancies were found. The l NRC was notified via the ENS as required, j At 2:07 p.m., on June 27, a channel B reactor enclosure isolation i occurred. The inadvertent isolation occurred while perf Y q a l surveillance test on A reactor enclosure isolation chan . The j test was repeated but the licensee could not duplicate the proble ;

it is suspected that an isolation valve, which is required to be !

shut to perform the test, was not shut fully. After repositioning i the valve the surveillance was performed satisfactory. The NRC was notified via the ENS as require [

On July 4, Mr. Martin McCormick, Jr. assumed the duties of Plant !

Manager. He previously held the position of Maintenance Division !

Manager for nuclear operations and electric production. Since the !

reassignment of the previous Limerick piant manager Mr. G. Leitc (

Vice President-Limerick, had assumed the plant manager dutie I h

The inspector reviewed the potential impact that the recent dry beather could have on plant operation. The use of water from the Schuylkill River for cooling tower makeup is restricted when flow

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j decreases below 530 cubic feet per second (cf s) or when dissolved oxygen levels drop below specified limits. To compensate for low flow or dissolved oxygen levels water is released from upstream j reservoirs. Based on existing reservoir reserves plant operat. ions could continue for approxiettely 100 days af ter river conditions necessitate release from the reservoirs. If the Titus and Cromby I fossil generating stations are also operating approximately 75 days I of reserve water is available to support Limerick, j

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During the inspection period dissolved oxygen levels required reservoir releases for several days and by the end of the period river flow had decreased below 530 cfs requiring releases on a daily basis.

I 2.4 Followup on Outstanding Inspection items (ClosedL0eficiency 87-07-0_1. This deficiency was written when, I during a simulator examination, the senior reactor operators had difficulty in reentering Emergency Operating Procedures (TRIPS)

after exiting them follcwing subsequent reentry conditions. Generic Letter 82-33 (NUREG-0737) Supplement ! requires that operators be trained in accordance with Emergancy Protection Guidelines for upgraded TR!p procedures. The licensee has conducted training in this area for all licensed personne The resicent inspector reviewed documentation of the training, which conforms with the Generic Letter $2-33, that has been conducted twice since the March inspection of 1937. This item is close [ Closed) Deficiency 87-07-0 This deficiency was written when coe.tunication 01' authority and responsibility between the shif t supervisor and shift superintendent was demonstrated unsatisfactor T:ie event contributsd to the failure of four SRO candidates. The resident inspector has reviewed documentation of retraining in this area including a video tape and a procedure OPS MAN 6.2 which delineates methods of conm.unication to ensure reliable and accurate transmissien of informatio This item is close Closed) Deficien:y 87 07-04 This deficiency was written when RO

[a W $R6 candidat,es cemonstrated unsatisfactory performance using procedure GP-8, Containment Isolation Rese The inspector verified that the licensee has conducted additional training with all licensed personnel, in the use of procedure GP The licensee has entered this training into the two year retraining cycle to reinforce the itcensed personnel knowiecge concerning containment isolatio This item is close '

2.5 fuel __ teak As discussed in report 50-352/88-03, the liennsee identified evidence of a fuel leak on March 25. During the inspection period reactor power has been limited to approxirately 85', in order to minimize the potential for additional cladding degradatio Oose equivalent todine in the reactor coolant and offgas system activity remained rel4tively constant, during the period and are less than 5*, of the Technical Specification limits. The resident inspectors will continue to monitor licensee action I

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2,$ NRC Bulletin 88-07: power Oscillations in Boiling Water Reactors

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The power oscillation event which occurred at LaSalle Unit 2 on !

March 9, 1938 w&s reviewed with licensee personnel to determine if ;

there were any immediate iafety implications for Limerick, The !

oscillations at the laSale plant occurred following a dual l recirculation pump trip. The incediate action in Limerick procedure :

0T-112, Recirculation Pump Trip, instructs the operator to insert control rods as required to prevent a reactor scram, This action is !

recommended by General Electric in Safety Information Letter ($1L) !

No. 320 Revision 1, to minimi:e the possibility of experiencing '

sustained power oscillations, f

Additional procedure clarifications are being prepared in response f to the bulletin and will be reviewed in a future report, i i

2.7 Motor Control Center Bus Bar Failure On June 13, a fault occurred on the 0144-R-H motor control center l (NCC) which caused its feeder circuit breaker te trip. The fault was due t,o arcing between the MCC bus bars which occurred when the B reactor enclosure cooling water (RECW) pump was started. The I fault apparently was the result of a deteriorated electrical connection where the motor controller stabs engage the vertical bus bar The defective motor controller was removed and the MCC was re-energi:ed within about one hour. Due to damage caused by the fault, the bus bars which supply the B RECW pump were replaced on June 1 i i

The vertical bus bars are constructed of aluminum, and it is j suspected that when large electrical leads are operated, the al nimum !

bus cannot adequately conduct the heat from the contacts, and gradual deterioration of the cennection results. The vendor, Eaton l j

Corporation (Cutler-Har..er), indicated potential reasons for stab interface failure could be: l 1, High ambient levels of vibratio . Long high current starting time . Escesstse wear on alpinum bus pining due to fr(quent unit insertion and withdrawal,

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4 Vertical bus or stab clip plating deterioration. This could be I original plating defects or degradation of plating due to l corrodents in the attospher J The vender recomends that the alumin o bus bars be replaced with copper bars to alleviate this condition.

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Ouring the corrective maintenance the licensee also measured the  !

contact resistance for all of the motor controllers in the 0144-R-It l MCC. The resistance readings were in the range of 0.001 to 0.004  ;

ohms although one reading was found to be 0.011 chms. After removing and reinstalling the controller, this contact resistance (

dropped to the same appruximate value as the other controller [

The licensee subsequently inspected the bus connections for Il motor controllers installed in other MCC's, These controllers supply l

power for so,te of the larger plant loads. Of the 11 inspected,  :

only the 11, instrument air compressor, showed signs of contact deterioration and required the bus bars be A ce These bus bars )

were replaced on June 25, with cepper bus t< Several non-safety i MCC leads have also experienced contact drJ-< + 4 tio j This conditicn was previously reported in accordance with 10CeR50.55(e) on the Limerick Unit 2 docket 50-353. That report was  ;

made due to similar failures which were experienced at a licensee [

fossil generating station (Eddystone). l

The MCCs at Lieerick were previously modified to provide an I additional atount of stab engagement on the bus bars and also have seismic citps installed to better secure the motor controller in the }

i FCCs. Essed on these features, contact failures were not expecte j As of the end of the inspection period all the size 4 motor f controllers, in safety related MCCs, have been inspected and found satisfactory or were repaire }

The licensee in continuing to >

evaluate the results of their inspections and continuity {

reasurenents. Additional inspections of non-safety related size i four starter connections are planned and inspection of si:e three starter connections are also being considere The resident inspectors will ccntinue to monitor actions taken to I resoh e this problem, i 3.0 Survetllance_ Observations (61726)

During this inspection period, the inspector reviewed in progress surveillance testing as well as completed survettlance packages. The inspector verified that surveillances were perfermed in accordance with licensee approved procedures and NRC regulation The inspector also l i

l verified that instruments used were within calibration tolerances and I j that qualified technicians performed the surveillances, t

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. . 9 The following surveill.inces were reviewed:

ST-6-052-231-1 A Loop Core Spray Pump Valve and Flow Test ST-6-107-590-1 Daily Surveillance Log ST-6-092-312-1 Mont'ly D-12 Diesel Run ST-6-092-314-1 Monthly D-14 Diesel Run ST-6-020-232-1 0-12 Diesel Generator Fuel Oil Transfer Pump, Valve and Flow Test SP-ST-014 A Loop Core Spray Pump Minimum Flow Verification Inspec'. ion SP-ST-014 performed flow rate checks on the A and C core spray pump minimum now lines utilizing a clamp-on ultrasonic flow indication. The A and C pumps shaie a common minimum flow path. The test was written to gather data to investigate the potential for pump damage due to inadequate minimum recirculation flow as identified in NRC Information Notice No. 87-59: Potential RHR Pump Loss. The test results show that the minimum flow with the pumps opera-ing at shutoff head meets the vendor recommendations The subject of adequate minimam pump flows has subsequently become the subject of NRC Bulletin No. 88-04: Fotential Safety-Related Pump Loss and the inspectors will follos the licensee actions related to this subjec No violations were identifie .0 Maintenance Observations (62703]

The inspector reviewed the following safety related maintenance activities to verify that repairs were made in accordance with approved procedures, and in compliance with NRC regulations, and recognized codes and standards. The inspector also verified that the replacement parts and quality control utilizec on the repairs were 'a compliance with the licensee's QA progra Work Order Number Description 8803565 Motor Control Center 0144-R-H Bus Bar Replacement i

8883553 Bench Test RCIC PSV-050-1F018  ;

8803263 RCIC Turbine Overspeed Trip Repairs 8803732 Inspaction of D134-R-H-01 Bus Connections l 8803741 Inspection of 0134-R-H-03 Bus Connections

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Reactor Core Isolation Cooling (RCIC) System Outage l On June 20, the RCIC system was taken out of servir.e to replace worn parts in the mechanical overspeed trip mechanism. During this repair i numerous other minor repairs and preventative naintenance items were also l performed. The repair work continued around the clock and the test personnel immediately performed the retest when the final work was completed. The pre-outage preparation, planning and in process management was good and resulted in a minimal system octage duratio No violations were identifie .0 , Review of Periodic and Special Reports (90713)

Upon receipt, the inspector reviewed periodic and special reports. The review included the following: inclusion of information required by the NRC; test results and/or supporting information consistent with design predictions and performance specifications; planned corrective action for resolution of problems, and reportability and validity of report informatio The following periodic report was reviewed:

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Unit 1 Monthly Operating Report - May 1988 The inspector had no questions concerning this repor .0 Licensee Event Report Followup (90712, 92700)

The inspector reviewed the following LERs to determine that reportability requirements were fulfilled, immediate corrective action was taken, and corrective action to prevent recurrence was accomplished in accordance with technical specification This LER reports various engineered safety feature actuations due to the loss of power to a Reactor Protection System logic panel as a '

result of a blown fus This event was previously reviewed and documented in NRC inspection l report 50-352/88-13. In addition to the LER stated action of '

inspecting and replacing the fuse clips, if necessary, the inspector .

confirmed that the station has requested an engineering review of the  !

design to determine if a plant modification is appropriate to correct the conditio The inspector had no further questions or concerns related to this LE I

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. . 11 88-017 This LER reports an unplanned isolation of the reactor enclosure along with the actuation of the standby gas treatment system, reactor enclosure recirculation system, and nuclear steam supply shutoff system due to the failure of a system solenoid valv This event was reviewed at the time of occurrence (Inspection Report 50-352/88-013). The inspector had no additional questions upon review of the repor and 88-021

These LERs report automatic actuation of the control room emergency fresh air supply (CREFAS) system, an engineered safety feature, resulting from a chlorine concentration signal believed to be caused

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by rainwater contacting a chlorine analyzer prob The inspectors are following licensee actions related to a plant modification to eliminate CREFAS actuations du; to instrumentation spike The modification has been delayed because the chlorine probe manufacturer went out of business earlier this yea This LER reported several locking springs on Agastat relays which were found missing or unsecured as discussed in previous report 50-352/83-13, section 8.3. This condition may affect the ability of plant systems to perform their safety related functions during or af ter a seismic even The licensee is continuing to evaluate the effect of the unsecured clips on the relay's seismic qualification and will report the results in a supplement to the LER. The inspectors had no further questions 2.t this tim l 88-020 This LER concerns an isolation of the reactor enclosure secondary containment on low differential pressure due to the inability of the exhaust air fans to maintain differential pressure as a result of a severed instrument air line tube. The tubing supplied air to the blade pitch controller on the B reactor enclosure exhaust fan which is not a safety related component. This event was previously reviewed in report 50-352/88-13, section 2.3. The inspector had no '

further questions concerning this even l 7.0 Assurance of Quality During this assessment period the following issues relate to licensee assurance of quality: 1 I

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Continued close monitoring of the fuel leak (section 2.5)

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Prompt action to inspect and correct safety related MCC bus bar problems following the 0144-R-H failure (section 2.7), however the potential for such a failure had been identified in February 198 Well managed RCIC maintenance outage resulting in a minimum system inoperability period (section 4.0).

8.0 Exit Interview (30703)

The NRC resident inspectors discussed the issues in this report throughout the inspection period, and summarized the findings at an exit meeting held with the Plant Manager, Limerick Generating Station, on July 7, 1988. No written inspection material was provided to licensee rupresentatives during the inspection perio l

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