IR 05000352/1988013

Insp Rept 50-352/88-13 on 880501-31.No Violations Noted. Major Areas Inspected:Plant Tours,Observation of Maint & Surveillance,Review of LERs & Periodic Repts,Review of Operational Events & Sys Walkdown

ML20195C988
Person / Time
Site:	Limerick
Issue date:	06/15/1988
From:	Linville J, Williams J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20195C986	List: IR 05000352/1988013 ML20195C982
References
50-352-88-13, NUDOCS 8806230009
Download: ML20195C988 (12)

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W U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report No. 88-13 Dockei, N License N NPF-39 Licensee: Philadelphia Electric Company 2301 Market Street Philadelphia, Pa 19101 Facility Name: Limerick Generating Station, Unit 1 Inspection Period: May 1 - May 31, 1988 Inspectors: T. J. Kenny, Senior Resident Inspector L. L. Scholl, Resident Inspector J. H. Williams, Project Engineer Reviewed by: d' 6 d H. Williams, Project Engineer Date

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Approved by:

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s Linville,04Miief, Pro 'e ts Section 2A

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Summary: Ro ne daytime (99 hours0.00115 days <br />0.0275 hours <br />1.636905e-4 weeks <br />3.76695e-5 months <br />) and backshift/ holiday (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />)

inspections of Unit 1 by the resident inspectors consisting of (a) plant tours, (b) observations of maintenance and sueveillance, (c) review of LERs and periodic reports, (d) review of operational events and (e) system walkdown .

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During this inspection period the licensee:

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Operated the plant at 85 to 90% power while monitoring the previously identified fuel failur Submitted several LERs (section 6.0), monthly operating report (section 5.0).

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Issued Upset Report UR-038 and Event Report No. 21 which document additional reviews of the April 9,1988 reactor scram (section 5.0).

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Reviewed several problems identified during Unit 2 construction for impact on Unit 1 operation (section 8.0).

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Performed a detailed investigation into the root cause for the blefn !

fuse on May 7 which caused a half scram and various system l 1solations (section 2.3). I

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8806230009 880616 PDR ADOCK 05000352 O DCD g

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DETAILS 1.0 Persons Contacted Within this report period, interviews and discussions were conducted with members of licensee management and staff as necessary to support inspection activit .0 Operational Safety Verification (71707, 70709, 71710 and 71881)

2.1 Documents Reviewed

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Selected Operators' Logs

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Shift Superintendent's Log

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Temporary Circuit Alteration Log

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Radioactive Waste Release Permits (liquid and gaseous)

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Selected Radiation Work Permits (RWP)

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Selected Chemistry Logs

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Selected Tagouts

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Health Physics Log 2.2 The inspector conducted routine entries into the protected areas of the plant, including the control room, reactor enclosure, fuel floor, and drywell (when access is possible). During the  :

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

I&C), mechanics, security personnel, supervisors and plant managemen The inspections were conducted in accordance with NRC Inspection Procedures 71707, 71709, 71710 and 71881 and affirmed the i licensee's commitments and compliance with 10 CFR, Technical i

Specifications, License Conditions and Procedure :

No violations were identified ,

2. Engineered Safety Feature (ESF) System Walkdown: (71710)

The inspectors verified the operability of the selected ESF system by performing a walkdown of portions of the system to confirm that system lineup procedures match p'.nt drawings and the as-built configuration. This ESF system walkdown was also conducted to identify equipment c7nditions that might degrade performance, to determine that instrumentation is calibrated and functioning, and to verify that valves are properly positioned and locked as appropriate. Portions of the standby liquid control system were inspecte . Safeguards Battery Seismic Lifetime and Qualification (Regionalif87-07J For batteries supplying vital loads, the following were evaluated for acceptability:

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(a) Estabitshment of seismic lifetime

.(b) Maintenance of seismic qualification "Nuclear Environmental Qualification Report", N QR1-16374/QR8-31947 by C&D Batteries concludes that C&D Batteries KC-7 and LK-21 cells and battery racks are environmentally qualified 'for a 14 year life, when installed and maintained in accordance with the manufacturer's and IEEE recommendations, for service in the Limerick Generating Station. KC and LC type batteries were tested to demonstrate that they are capable of performing their safety function throughout their qualified life. The tests included thermal aging, seismic simulation and period capacity tests to demonstrate battery performanc C&D Manual, "Stationary Battery Installation and Operating Instructions", section 6.1 discusses battery life and notes that a stationary battery will not experience more than 200 evenly spaced discharge '

cycles in its useful lif Frequent or greater depths of discharge can shorten service life to less than 10 years even with preper maintenance and operating condition The licensee uses surveillance tests to evaluate battery performance and has no set lifetime establishe . Seismic qualification is maintained though inspection, test and maintenance of the cells, cell plates, terminals and battery rack .3 Inspector Commerits/ Findings The inspector selected aspects of the unit's operation to determine compliance with the NRC's regulations. The inspeccor determined

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i 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 May 1, at 5:53 a.m., three reactor enclosure isolations on low differential pressure were received due to problems associated with restarting the reactor enclosure ventilation fan The fans were I being started during restoration to a normal ventilation lineup I following the performance of standby gas treatment system (SBGTS)

troubleshooting. After several attempts the fans remained operating and the SBGTS was secured. The licensee investigated the problem with the reactor enclosure fans, however the system operated

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properly during several attempts to reproduce the occurrence. A possible cause is that sufficient time had not elapsed to allow air flows to stabilize and the system engineer is reviewing the operating procedure for potential improvements to prevent recurren On May 7, at 7:46 a.m., a 60 amp fuse in the Reactor Protection System (RPS) 'B' channel circuitry failed causing a half scram and various system isolations. Recirculation pump speed was reduced to limit the pump motor temperature increase while cooling water was isolated. This resulted in a power decrease from 90 to 80%. The fuse was replaced and the half scram and all isolations were reset within 20 minutes. All systems functioned as designed. No immediate cause for the failed fuse was evident and the fuse will be analyzed by the vendor to determine the failure mode. Fallowup troubleshooting determined that the circuit was normally loaded to approximately 48 amps as compared to tie ' A' channel RPS circui It was also discovered that the fuses in these circuits operated at relatively high temperatures. Readings obtained with a thermal heat detector, and confirmed by contact pyrometer readings showed that the 'A' channel fuse and end caps were in excess of 200 degrees F and the 'B' channel fuse and caps were approximately 150 degrees The 'A' channel fuse was replaced in order to determine if the high temperatures may be an indication of fatigue, however the new fuse also operated at relatively high (200 degrees F) temperatures within a day of installatio The licensee obtained and installed fuse clip clamps which appear to have been successful in improving the electrical contact between the fuse clips, and the fuse end caps. Temperatures on the 'A' channel fuse end caps decreased by approximately 100 degrees F and a temperature decrease was also noted on the 'B' channel. This resulted in a drop of all the fuse temperatures to less than 140 degrees The fuse clip clamps were installed in accordance with the Temporary Circuit Alteration (TCA) Procedure and will remain installed while an engineering evaluation is performed to determine what circuit modifications are appropriat At 9:55 p.m., on May 9, a reactor enclosure isolation occurred on a low differential pressure signal when the 'B' reactor enclosure exhaust fan was placed in service. An air leak on the 'B' fan discharge damper positioner starved the air supply to the remaining fans which caused them to trip resulting in the loss of building differential pressur The reactor enclosure recirculation and standby gas treatment systems started per design. The air leak was repaired and al. systems were returned to their normal alignmen On May 11, at 2:48 p.m., a control room isolation occurred when the

'C' chlorine detector momentarily spiked upscale. The control room j emergency fresh air system started as designed. The spike occurred i on only one of the four channels and is suspected to have been i i

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- 5 caused by electrical noise. A planned modification to the isolation logic will eliminate isolations caused by momentary spikes occurring on a single channe On May 15, during the weekly control rod exercise test, control rod 18-55 did not initially respond to the rod "insert" signal. After several unsuccessful attempts at inserting the rod the control rod drive (CRD) cooling water i olation valve was shut momentarily and the rod inserted normall The cooling water was unisolated and the rod was again successfully repositioned leading the licensee to suspect that the CRD cooling water check valve may have stuck open and then was freed. Similar symptoms were seen on April 17 for rod 22-15 when the CR0 was being vented and on May 2 for rod 22-47 during a power i.1 crease. Based on these similar occurrences the licensee began an investigation to determine the cause and the potential safety implication The effect of a stuck open check valve on the ability to scram a control rod was reviewed and it was determined that the rod would scram with a negligible effect on the actual insertion time. At normal reactor pressures the CR0 mechanism internal check valve would seat and reactor pressure would scram the rod. Based on previous testing performed by General Electric, under conditions which would approximate a stuck open check valve, the hydraulic control unit scram accumulator successfully inserted the control rod with a minimal increase (approximately 0.1 second) in scrr'. time. The licensee is preparing a procedure which wil? provide detailed instructions to the plant operations staff or. what actions to take in the event a control problem is encountered and will specify data to be taken to permit analysis and resolution of the proble On May 21, at 6:47 p.m. , a reactor enclosure isolation occu. red when the 'B' exhaust fan blade pitch adjustment moved to the minimum position due to the loss of instrument air to its controlle The air loss was caused by damaged tubing which was subsequently repaired and normal ventilation was restored at 10:00 p.m. The standby gas treatment and reactor enclosure recirculation systems started and restored the required differential pressure, however the systems initiated conservatively in approximately 70 seconds instead of the normal 100 second time delay. The licensee has replaced the time delay relay which controls the time delay The inspectors discussed a concern with the licensee that the continuous fire watch on the 313 foot elevation of the reactor enclosure was posted at the boundary of a radiation area resulting in him having a partially obstructed view of the area he was monitoring. Subsequently it was determined that upon performance of additional radiation surveys an area with low radiation levels was present and allowed posting of the watch directly at the area being monitored for fire .- . .. _ - . - . . . . . _ . _ . - . . --.

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. 6 2.5 Fuel Leak As discussed in report 50-352/88-08, the licensee identified evidence of a fuel leak on March 25. During-the inspection period reactor power has been limited to approximately'90% in order to minimize the potential for causing additional cladding degradation while a more detailed review is performed to determine if operation at higher power is practical, Dose equivalent iodine in the reactor coolant and offgas system activity have remained relatively constant during the period and are well below the Technical Specification limits. Dose equivalent iodine is less than 1% of the limit while offgas activity is less than 4% of the limit. The resident inspectors will continue to monitor licensee action .

3.0 Surveillance Observations (61726)

During this inspection period, the inspector reviewed in progress surveillance testing as well as completed surveillance packages. The

! inspector verified that surveillances were performed in accordance with licensee approved procedures and NRC regulation The inspector also verified that instruments used were within calibration tolerances and that qualified technicians pe.dormed the surveillance The following surveillances were reviewed:

ST-6-107-590-1 Daily Surveillance Log ST-6-092-313-1 Monthly 0-13 Diesel Run ST-6-092-314-1 Monthly D-14 Diesel Run ST-6-095-901-1 Div. I 125/250 VDC Safeguard Battery Weekly Inspection ST-6-095-902-1 Div. II 125/250 VDC Safeguard Battery Weekly Inspection ST-6-095-903-1 Div. III 125 VDC Safeguard Battery Weekly Inspection ST-6-095-904-1 Div. IV 125 VDC Safeguard Battery Weekly Inspection ST-1-078-301-0 'A' CREFAS Functional Test During the inspection period the licensee informed the inspector that the measureaent of the cor rol room boundary air in-leakage had not been performed within the frequency previously committed to in a May 13, 1987 letter to the NRC. Tne testing was overdue by 73 days. The air in-leakage test is normally required to be performed at 18 months intervals per Tethnical Specification 4.7.2 with an allowable leakage of 525 cubic feet per minute (cfm). A license amendment was approved which permits leakage up to 2100 cfm until the issuance of the Unit 2 operating license. This amendment enables boundary penetrations to be opened to j

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facilitate Unit 2 cable pulls and it was during the processing of the amendment when tho licensee committed to performirg the leakage test at six month interval In addition to the increased testing, administrative controls were put in place to closely monitor and control the number of boundary penetrations which may be opened without exceeding 2100 cfm leakag The test was overdue because of an oversight on the part of the system engineer who was tracking the test performance at the increased frequency. The computerized surveillance test tracking system was programmed for the 18 month frequenc The administrative control system was maintained during this period and when performed the leakage test was sseisfactory thus there were no apparent safety concerns as a result o' the oversigh The licensee has revised the computerized tracking schedule for this test to ensure the six month frequency is not exceeded in the future. The inspector had no further question l No violations were identified, l

4.0 Maintenance Observations (62703)  ;

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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 inspecior also verified.that the replacement parts and quality control utilized on the repairs were in compliance with the licensee's QA progra Work Order Number Description 8686303 Clean and Examine 0134-R-H 480V Motor Control Center 8686309 Clean and Examine D134-R-E 480V Motor Control Center 8686299 Clean and Examine 0114-R-C 480V Motor Control Center i

8686297 Clean and Examine D114-R-G 480V Motor Control Center 8607338 Clean and Examine D114-R-G 480V Motor Centrol l Center The licensee experienced a problem involving pillow block bearings on the control room emergency fresh air system fan operating at slightly higher than normal temperatures due to excessive greasing. The inspector reviewed the lubrication procedure with licensee personnel and was __ - - _ _ . _ ~ ~ . _ . . . _ _ _ _ _ . . . - _ _ . , _ , _ _ . _ _ _

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informed that the procedure is being revised to grease the bearings while the motor is running and to decrease the frequency of greasing to prevent

. recurrenc 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 information. The following periodic reports were reviewed:

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Unit 1 Monthly Operating Report - April 1988

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Independent Safety Engineering Group Event Report #21 - Reactor Scram Caused by IRM Upscale

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Upset Report UR-018-Unplanned Scram due to Cold Water Injection during a Scheduled Plant Shutdown Low in the IRM Range l

The inspector had no qsestions concerning these reports, l 6.0 Licensee Event Report followup (90712, 92700)

The inspector reviewed the fcllowing 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 identified a wiring change which was improperly implemented for the drywell sump flow detection systems. The result of the change was that drywell leakage rates in excess of the Technical Specification limits may-not have been annunciated in the main control roo The systems detect drywell unidentified and identified leakage by j monitoring the floor drain and equipment drain tank levels over a '

timed interval. At the end of the interval a leakage rate is calculated in gallons per minute and control room indicators are updated and annunciators actuated if required. If leak rates are high and tne tank fills and drains within the timed interval, the calculated flow rate could be erroneous.

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Another portion of the circuitry monitors the number of gallons collected in the tank from the start of the timing interval and will annunciate when a fixed number of gallons is collected regardless of the elapsed time. This feature which ensures any gross leak rate is detected and annunciated was inadvertently disabled during a system modification which made it subject to the same fixed timing interval-as the circuits which calculate an actual leak rate, The result again was that excessive leakage rates may not have been detected by this circui Upon discovery, the circuit was immediately returned to its original configuration oy use of a temporary circuit alteration. During the time that the circuit was improperly modified leak rates up to the technical specification identified leakage limits would have been accurately detected and unidentified leakage of up to twice the ,

technical specification limits would have been identifie Backup leakage monitoring systems were also available to identify drywell leakag The cause of the event was a combination of personal errors by a contract design engineer, a utility employed responsible system engineer and utility employed site system engineer in that a deficient design was provided and inadequately documented such that independent reviews did not identify the proble In general, modification designs have been good in the past and this appears to be an isolated cas The inspector had no further questions88-006 This LER reported a design deficiency, discovered during an engineering review, which resulted in three inoperable fire protection system water curtain It was calculated that three fire protection system water curtains were unable to achieve a discharge density of 0.30 gpm/ft2 at floor level as required by the Limerick Final Safety Analysis Report and the Fire Protection Evaluation ;

Report. The cause of this event was a deficiency in the design '

assumptions by the fire protection sprinkler contractor. The water curtain piping and nozzles are being redesigned-to meet the required coverage. A previous licensee review of selected calculations of the sprinkler contractor did not reveal any significant design deficiencies. Additionally, the licensee obtained a consultant to perform a review of other LGS fire protection systems designed by the sprinkler contractor to provide assurance that the deficiency was an isolated cas l l

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Five additional calculations were selected for review based on their potential- for having a h.br probability of improper design assumptions. No significant errors were discovered during the consultant's review. The licensee immediately instituted continuous fire watches as compensatory measures. .The inspector had no further question This report addresses a reactor scram which occurred during a controlled shutdown with reactor power low in the intermediate range. Past shutdowns typically concluded with a manual scram from approximately 20*J powe This shutdown was being accomplished by normal control rod insertion so that power was reduced gradually to avoid a pressure transient, thus minimizing the fission product release into the reactor coolant through the fuel cladding defec Since the operators do not normally perform this type of shutdown the operator did not anticipate the relatively rapid power increase which could occur due to moderator temperature chang The inspector also reviewed the operations department upset report and the Independent Safety Engineering Group (ISEG) event report related to the scram and had no further question ,88-013 Reported the details of a scram which occurred while shutdown with all rods inserted and was caused by a failed IRM detecto l Reported a control room isolation caused by a chlorine detector I which was wetted by rain wate Reported an inadvertent emergency diesel generator start which occurred during troubleshootin These events were reviewed in inspection report 50-352/88-08, section 2.3, and following the review of the associated reports the inspector had no further questions on LERs G8-013,88-014 or 88-01 .0 Drug Allegation (RI Allegation 88-43)

An allegation of drug use by a contractor employee was determined to be-unsubstantiated. The employee was interviewed.by the licensee and agreed to submit to a drug test with negative results. Yhis allegation is close '

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8.0 Unit 2 Findings with Potential Unit 1 Implications The following Unit 2 problems were reviewed for potential impact on Unit 1 operation:

8.1 Bellville washer failures were identified in Unit 2 Cutler-Hammer Motor Control Center (MCC) bus bar connections. All of.the accessible bus bar connections on the Unit 1 MCCs were inspected by the licensee for evidence of Feating and for cracks in the bellville washers. The inspector also reviewed several work packages involving periodic inspection and cleaning of MCCs. No deficiencies were identified during these inspections which could have been indicative of failed washers. The licensee is continuing to investigate this concern and is awaiting additional information from the vendor with which to perform a final assessmen .2 Unit 2 problems associated with missing and undersized welds on instrument racks were reviewed on Unit 1 and although no missing ,

welds were identified some undersized welds were present. General i Electric reviewed the deficiencies and determined that the weld I

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deviations were acceptable and that the existing dynamic qualification is adequat .3 On Unit 2 locking clips were found not to be engaged over the plug-in type relays in various control panels. The clips were part of the relay assembly used to perform seismic qualifications and therefore without the clip engaged the ability of the relays to remain operable during a seismic event is not clear. Without the

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l clip eagaged a significant force is required to unplug the relays and the licensee is having seismic qualification testing performed to determine operability of the relays without a locking clip. A 100*s inspection of the relays was performed on Unit 1 and all clips not engaged were immediately engage I The inspector discussed a concern with the licensee regarding how and when station management is informed of quality control findings so that the impact on continued safe plant operation can be assessed. The inspector was informed that procedure QCI-01, Preparation of Nonconformance Reports (NRs) is being revised to provide instructions to notify the station of NRs and includes a

"Station Acknowledgement" signoff on NR form QAD 9. The inspectors will follow licensee actions in the resolution of this issu .0 Assurance of Quality During this assessment period there were licensee actions in support of assurance of quality as evidenced by the following.

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In-depth troubleshooting to determine the root cause for the spurious blown reactor protection system fuse (section 2.3). l i

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Prompt station management attention in reviewing the potential safety implications associated with sticking control rod drive cooling water check valve Cont qued close monitcring of the fuel cladding defect and potential raditlogical concerns associated with the lea .0 Exit Interview The NRC resident ir.spectors discusscd the issues in this report throughout the inspection period, and summarized the findings at al exit i meeting held with the Vice president, Limerick Generating Station, :q June 2, 1988. No written inspection material was provided to licensee l representatives during the inspection perio !

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