IR 05000352/1989003
| ML20244B474 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 04/07/1989 |
| From: | Linville J, Williams J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20244B473 | List: |
| References | |
| 50-352-89-03, 50-352-89-3, NUDOCS 8904190218 | |
| Download: ML20244B474 (14) | |
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Philadelphia Electric Company
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No. 89-03 Docket No. 50-352 License No. NPF-39 Licensee:
Philadelphia Electric Company l
Correspondence Control Desk
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P.O. Box 7520 Philadelphia, Pa 19101 Facility Name:
Limerick Generating Station, Unit 1 j
Inspection Period:
February 27 - March 26,1989 Inspectors:
T. J. Kenny, Senior Resident Inspector L. L. Scholl, Resident Inspector Y/
Reviewed by-A J
.6Mf411ams j6ct Engineer DAte/
Approved by:
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eY Linville, ief, Projects Section 2A Dfte, /
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Summary:
Routine daytime (1 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) and backshift/ holiday (27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br />)
inspections of Unit I by the resident inspectors consisting of (a) plant tours, (b) observations of maintenance and surveillance testing, (c) review of j
LERs and periodic reports, (d) review of operational events and (e) system t
walkdowns.
During this inspection period:
The licersee:
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commenced core reloading (section 2.3)
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completed reactor vessel nozzle weld ultrasonic inspections (section 5.1)
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briefed the NRC staff on the fuel failure investigation, the nozzle weld
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indication and Appendix R review status (sections 2.3 and 5.1)
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submitted several licensee event reports (LERs) and followup evaluations of events determined not to be reportable (section 7.0)
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Philadelphia Electric Company
l continued tie-in and testing of common systems including Emergency
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Service Water (ESW) and Residual Heat Removal Service Water (RHRSW)
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continued other scheduled outage activities.
The inspectors:
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reviewed LERs and special reports (section 7.0)
l observed ESW system flow balancing (section 4.1)
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inspected allegations related to contamination controls and background j
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investigations (section 8.0)
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closed an unresolved item of low pressure coolant. injection system instrument setpoints (section 3.0)
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reviewed licenses reevaluation of previously reported events (section 9.0).
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DETAILS j
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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 activity.
2.0. Operational Safety Verification s
2.1 Documents Reviewed
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Selected Operators' Logs Shift Superintendent's Log
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Temporary Circuit Alteration Log 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
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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 was possible).
During the inspection, discussions were held with operators, technicians (HP &
I&C), mechanics, security personnel, supervisors and plant management. The inspections were conducted in accordance with NRC Inspection Procedure 71707 and affirmed the licensee's commitments and compliance with 10.CFR, Technical Specifications, License-Conditions and Administrative Procedures.
2.3 Inspector Comments / Findings On March 2, the licensee determined that the air flow switches for
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the drywell hydrogen mixing unit coolers were not properly sealed thus violating their environmental qualification.
The potential for this problem was realized during cycle 2 operation but due to their location in the drywell, the switch inspection could not be performed until the refueling outage. The flow switches are not required for the fan operability and were jumpered out of the circuit when the potential for improper switch housing sealing was identified as discussed in NRC inspection report 50-352/88-20. At that time, the item was left as unresolved item 88-20-01.
This closes out the unresolved item and the inspector will review the licensee's actions upon issuance of the LER.
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On March 15, during a Unit 1. inspection of environmentally qualified equipment, the licensee discovered that a number of conduit-entries to electrical devices, located in high energy line break (HELB).
I compartments, were not properly sealed or protected against moisture ~
as required by the appropriate qualification installation instruction.
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' These electrical devices are associated with the steam leak detection-system and HVAC back pr' essure dampers and are required to function post-HELB to mitigate HELB consequences. Moisture. intrusion into these electrical devices could cause them to fail to operate. As a result, the required isolation to prevent reactor coolant pressure boundary leakage into adjacent areas would not occur and may be subject to unanalyzed environmental conditions. The resident inspector will review the licensee's actions when the LER is issued.
On March 15, the licensee met with members of the NRC staff to brief them on the status of their investigation into the root cause for the
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cycle 2 fuel clad failures and on their plans for configuring the core for cycle 3 operations such that the likelihood of additional fuel failures is minimized.
A presentation was also made to update the staff on the status of the 10 CFR 50 Appendix R safe shutdown analysis review. A followup meeting at which compensatory measures for problem areas will be presented is planned for April 6 at the NRC Region I office.
On March 18, the licensee commenced core reload in accordance with plant procedures. The inspector observed fuel handling activities to be generally well controlled.
3.0 Update of Open Items and Bulletins a.
(Closed) 50-352/88-09-03.
Low Pressure Coolant Injection (LPCI)
Valve Differential Pressure Instrument Loops. When plant shutdown reduces reactor pressure to below normal RHR discharge pressure, the measured differential pressure across the injection valves becomes negative and is no longer within the calibrated range of the instrument loop.
The shutdown condition caused a trip unit gross failure condition and false annunciation of RHR out-of-service in the main control room.
In a November 5, 1986 letter to the NRC, the licensee proposed a change to the trip setpoints of the LPCI differential pressure l.
permissives in order to increase the range of the instrument loops.
Amendment No. 16 to the Facility Operating License was approved by the NRC on February 9, 1989 and revised the instrument range to
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correct the gross failure indication when the plant is shut down.
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The licensee has subsequently recalibrates the instruments as permitted by the amendment, j
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- n Philadelphia Electric Company 7-The importance of the low pressure permissive calibrations.for the RHR-LPCI injection is emphasized in the Limerick Unit 1 PRA. The dominant failure mode for low pressure injection leading to core damage involves postulated miscalibration of the pressure channels.
which causes failure'of-the differential pressure permissive; thereby preventing LPCI' flow into the reactor vessel. The PRA also assumes L
that the miscalibration is combined with a failure of control room operators _to-recognize that the injection valves HV-51-017A through D have not automatically. opened.
t The inspectors observed portions of the channel B recalibration and reviewed the results of the 18 month calibration and trip unit check surveillance test for all four channels (ST-2-051-420-1 through 423-1).
Based on the issuance of Amendment 16 and the satisfactory recalib-ration of the instruments, this item is closed.
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(0 pen)Bulletin 85-03.
Motor-0perated Valve Common Mode Failures during Plant Transients due to Improper Switch Settings.
As requested by action item e. of Supplement 1 to Bulletin 85-03,
" Motor-operated Valve Common Mode Failures during Plant Transients due'to Improper Switch Settings," the licensee's letters dated May 27 and September 12, 1988, identified the additional. valves to be addressed in their program in response to the original bulletin.
Review of these responses indicates that the licensee's selection of the additional valves to be addressed in their program in response to the original bulletin meets the requirements of action item e. of the supplement to the bulletin and is acceptable.
The results of the inspections to verify proper inclusion of these valves in the bulletin program will be addressed in subsequent inspection reports.
_ 4.0 Surveillance /Special Test Observations (61726, 64704)
During this inspection period, the inspector reviewed in progress surveillance testing as well as completed surveillance packages. The
inspector verified that surveillance were performed in accordance with
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licensee approved procedures and NRC regulations.
The inspector also i
verified that instruments used were within calibration tolerances and j
that qualified technicians performed the surveillance.
j The following surveillance were reviewed:
ST-2-051-420-1 ECCS-RHR Injection Valve Differential ST-2-051-421-1 Pressure-Low (Permissive) Channel A, (B, C, D)
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Philadelphia Electric Company
i ST-2-051-422-1 Calibration / Functional Test
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ST-2-051-423-1 Calibration / Functional Test-RT-6-097-410-0 Daily Refueling' Platform Checkout ST-6-107-630-1 Core Alteration Testing for Offloading, Shuffling and Reloading the Core 4.1 ESW Loop 'B' Flow Balancing The inspectors also' witnessed portions of.the Emergency Service Water (ESW). system Loop B flow balance and reviewed the results of the completed test. The system flow balance was performed in order
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to establish design flows through Unit I and 2 components.to support
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two unit operation following the completion of the current refueling outage. The system was balanced in accordance with procedure 2FB54.18, Flow Balance Procedure, Emergency Service Water-Loop B.
This test establishes the required throttle valve positions and verifies adequate flow through all components in all modes of operation.
The flows met the requirements of the test, however, the licensee is revising the required flows for several components to values below those currently in the Final Safety Analysis Report, Table 9.2.3.
The Table will be revised as follows:
Required emergency diesel generator flow is reduced from 700
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gallons per minute (gpm) to 450 gpm (Licensing Document Change Notice (LDCN FS01649)
RHR system motor bearing oil cooler changed from 6 gpm to 5.1
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gpm per cooler (LDCN 1657)
Control room chiller condenser flow reduced from 815 gpm to 600
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gpm (LDCN F5-1658)
High Pressure Coolant Injection (HPCI) system room cooler flow was reduced from 200 gpm to 190 gpm (LDCN FS-845).
Reactor Core Isolation Cooling (RCIC) pump room unit cooler to
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be reduced from 80 gpm to 60 gpm (LDCN FS-1629, not available for review at time of this report).
The flow requirements were revised prior to the test when the emergency service water system computer analysis predicted that the original design flows may not be met during two unit operation.
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The inspector reviewed LDCNs FS-1649, FS-1657, FS-1658 and FS-845 and their associated safety evaluations.
The evaluations appear to adequately justify the reduced cooling water flows and also document the receipt of vendor concurrence (for operating at reduced flow.c)
for the affected equipment.
The inspector also reviewed the 'B'
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ESW pump performance.
The test review board had identified that the pumps did not achieve the developed flow specified in FSAR Table 9.2-2; i.e. 6,400 gpm at a total head of 240 feet.
During the test the 'B' pump operated at approximately 5,250 gpm and the 'D' at approximately 5,850 gpm at 240 feet head.
The inspector compared the pump performance data obtained in the flow balance procedure with the acceptance criteria in the quarterly 'B' loop ESW Pump, Valve and Flow Test ST-6-011-232-0 l
figures 1 and 2.
The comparison showed that the data for the 'D'
pump, although within the acceptable range, was below the reference total developed head curve. Also, the data for the
'B' pump in the high flow range (5,000-6,000 gpm) was found to be out of the acceptable i
range and in the alert range due to lower than expected flow rates.
Previous performance of the quarterly pump tests had yielded acceptable results, however, due to the system being lined up for single plant operation the data was obtained at lower flow rates.
The' pump performance concerns were discussed with the licensee engineering on March 23.
The acceptability of the pump performance is unresolved (50-352/89-03-01) pending resolution of the following:
The significance of the deviation between the FSAR 9.2-2 pump
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design parameters and the actual pump performance; i.e.
head-flow data.
The accuracy of the loop 'B' flow indicator FI-11-013B was
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questioned because the sum of the individual component flows was consistently higher than the loop actual flow as determined by FI-11-0138.
Section XI of the ASME Code, Table IWP-4110-1 requires that flow rate instruments used in inservice testing be accurate to +2% of full scale.
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The significance of the
'B' pump data which was in the alert range of the developed head curve.
The validity of performing the quarterly inservice testing at
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flow rates which are significantly below the maximum possible demand on the system.
The update of the emergency diesel generator load tables (FSAR
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section 8) since the power consumption of the control room chillers increases by 6% with the reduced condenser cooling j
flow.
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The inspector also reviewed procedures SP-65, Requirements for
Declaring RHR service water loops operable after the second refuel l
outage tie-in and SP-66, Requirements for declaring ESW loops
operable after the second refuel outage tie-in. The purpose of
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these procedures is to specify actions that must be performed in j
order to declare the systems operable. The inspector found the
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procedures to be a very comprehensive list of verification of I
construction completeness, preoperational testing, system lueups j
and surveillance testing.
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5.0 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 utilized on the repairs were in compliance with the licensee's QA program.
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8901729 Replace Scale and Recalibrates LPCI Differential Pressure Instrumentation - Channel A
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8901730 Replace Scale and Recalibrates LPCI Differential Pressure Instrumentation - Channel B 8901731 Replace Scale and Recalibrates LPCI Differential Pressure
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Instrumentation - Channel C
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8901732 Replace Scale and Recalibrates LPCI Differential Pressure Instrumentation - Channel C
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8881859 D-11, Emergency Diesel Generator Five Year Inspection and Maintenance 5.1 N2 Nozzle deld Indication As discussed in NRC Inspection Report 89-04 section 2.3, an indication was identified in one of the jet pump riser inlet nozzle welds during ultrasonic inspections.
The licensee has completed the 23 planned nozzle inspections, including all 10 jet pump riser inlet nozzles, with only one indication identified.
On March 8, the licensee briefed the NRC staff on the finding and the proposed plan of action which is to monitor the indication during the next operating cycle.
No weld overlay or other repair method is planned at this time.
The nozzle indication will be monitored using the crack arrest verification system (CAVS) and also with an acoustic monitving system.
The resident inspectors and Region I specialist inspectors will monitor licensee actions during the next operating cycle.
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Similar indications are being investigated on Unit 2 nozzle welds.
Refer to NRC Inspection Report 50-353/89-11 for additional information.
6.0 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 deportability and validity of report information. The following periodic report was reviewed:
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Unit 1 Monthly Operating Report - February 1989 The inspector had no questions regarding this report.
6.1 10CFR21, Report for Battery Charger Circuit Boards During the performance of preventative maintenance on the safeguards battery chargers, control circuit boards were replaced in accordance with the manufacturer's recommendation. When the chargers were retested following the maintenance they could not achieve the required output current.
Followup investigation revealed that the replacement circuit boards were not identical to the installed boards and a modification is required for the chargers to function properly. The original cards were operable and have been reinstalled pending modification of the replacements.
7.0 Licensee Event Report Followup (90712, 92700)
The inspector reviewed the following LERs to determine that deportability l
requirements were-fulfilled, that immediate corrective action was taken, I
and that corrective action to prevent recurrence was accomplished in l
accordance with technical specifications.88-009, Revision 3 The revision to this LER discusses additional engineering studies that have further identified the reason for a Reactor Water Cleanup System (RWCS) isolation.
Initially it was determined that a leaking relief. valve caused the temperature in the vicinity of the temperature sensing element to increase and isolate the system.
Further licensee engineering studies have determined that the sensing element is not in an area that is representative of the ambient room temperature.
Therefore the licensee has issued a modification package to move the temperature sensors to a more suitable location. The study also concluded that a modification, to add a high point vent to the RWCS, would be necessary in order to vent. air from the system which may be introduced during maintenance. The' licensee indicates this modification will be performed after the current outage.
The inspector will follow the licensee's progres _
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88-042, Revision 4.
This LER is tracking the cable separation problems that have been discussed in previous inspection reports.
Revision 4 identifies another separation problem and its corrective action, which was wrapping the cable in question.
The inspector has no further questions at this time.89-010, Refuel Floor Containment Isolation and Standby Gas Treatment System Initiation.
This event occurred on February 4, 1989 and was discussed in NRC inspection report 89-04. The inspector has reviewed the LER and has no further questions.89-012, Fire Protection for the RCIC System.
This event was discussed in j
inspection report 89-04 and the inspector indicated that the LER would be reviewed to assess the licensee's corrective actions. After review, the inspector determined that the procedure changes to administratively I
control the affected valve in the RCIC system was satisfactory. The l
valve will be placed in the open position and will be left in that l
position whenever the reactor is not in a cold shutdown condition. The i
inspector has no further questions at this time.
89-13, Inoperability of the triaxial peak recording acceleragraph while removed from the Reactor Vessel Head during refueling activities. This is a normal activity during refueling however technical specifications (TS)
state that whenever this instrumentation is out of service for more than 30 days a report must be made.
The licensee has submitted a change to amend the TS to allow the above normal practice without a special report.
The inspector has no further questions concerning this LER.89-014, Refuel Floor Secondary Containment Isolation and Standby Gas Treatment System Isolation. This event occurred on February 11, 1989 and was discussed in NRC inspection report 89-04.
The inspector has reviewed the LER and has no further questions.89-015, Missed Surveillance Requirement.
This event occurred on February 22, 1989 and was discussed in inspection report 89-04. The inspector has reviewed the LER and has no further questions.
8.0 Allegation Followup 8.1 As a result of an anonymous allegation, involving supervisors climbing in posted contaminated areas without proper anti-contamination (AC) clothing, the licensee conducted an inspection into the alleged practice.
The results were as follows.
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Within the area in c,uestion, there wer9 two contamination incidents, which resulted from walving :nto contaminated water from a backed-up floor drain.
There were no contaminations documented as a result of climbing in the area.
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'The Radioactive Work Permit (RWP) allows supervisors to. enter
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the area wearing booties and gloves, for inspection during work activities, but the RWP states that no climbing is permitted.
The inspection concluded that the practice of climbing, without proper AC clothing did not result in the spread of contamination.
However, the licensee has issued a letter to all parties who work in contaminated areas, reiterating the no climbing pol.cy without
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proper AC clothing.
8.2 On December 30, 1988, the Region I office received an anonymous allegation indicating that a contractor-employee had been directed, by. his employer, to improperly rely on false indication provided to him during a background investigation by an applicant for a security position at the Limerick Generating Station. The regional office transmitted a letter to the licensee, dated January 23, 1989, denoting the alleged concern and directed the licensee to conduct an investi-gation.' As a result of the investigation, the allegation was found
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to be unsubstantiated.
The investigation results were provided in a letter from the licensee to the regional office.-The inspector reviewed the letter and concluded that the allegation may have been a misunderstanding of a communication with the' wrong individual which was later corrected by contacting the proper individual for the correct information. The inspector had no further questions regarding this allegation.
9.0 Reevaluation of Previously Reported Items In NRC inspection report 88-26, three four hour reports were discussed.
The following paragraphs describe the results of the licensee's re-evaluation of these events.
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The report of December 21, 1988, declared Emergency Diesel Generator (EDG) D-13 inoperable due to a low fuel oil level in D-13 fuel oil storage tank. Technical specifications state that the minimum volume l
of oil allowable is 33,500 gallons. The licensee had been using a
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tank level of nine feet six inches to correspond to 33,500 gallons.
However, this level calculation had been conservative and had not taken into account the convex tank ends and the cylindrical shape of the tank.
New calculations indicate there had been 34,667 gallons in the tank (nine feet three and three quarter inches) at the time the report was made.
The new calculations also show that 33,500 gallons corresponds to a level (nine feet one quarter inch). The licensee has informed all of the operators of this and has made improvements to the affected logs to more accurately inform the operators of the correct tank levels and volume _ _ _ _ _ _ _
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The report' of January 17, 1989, declared the reactor vessel level indication nonconservative due to a level loss in the. reference leg of 14.5 inches. The licensee also declared the EDGs inoperable due to the nonconservative low level start signals for the EDGs.
The licenraa nas evaluated the level loss in the reference legs and has determined that.that much evaporation of water could not have
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occurred. The calculated evaporation loss with a vacuum on the I
vessel during the cooldown would have only been about a two inch j
loss.
The engineer conducting the evaluation concluded that the indication discrepancies were a result of the reactor vessel being subjected to i
an approximate negative 0.5 psig or about 14 inches of water (nngative)
during the time I&C performed the pressure measurements, thereby invalidating their calculations and temporary instrument calculations (the temporary level device was being installed at the time the problem was discovered). Thus the temporary installed instruments were indicating lower levels than actual because their calibrations were based on 0 psig not negative psig. The reporting of the EDGs being inoperable was conservative. The EDGs were capable of performing their intended function at that time.
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The report of January 21, 1989, identified the isolation of the Fuel Floor.HVAC following loss of power when fuses were removed and the failure of the standby gas treatment system to initiate due to testing in progress.
Because of plant conditions and parts of these systems were removed from service for testing and maintenance and in accordance with NUREG 1022, Supplement 1, the event was not reportable and the system was not required for the operating condition. The inspector has no further questions regarding this event.
The above events will not be followed up by an LER.
The inspector has no further questions at this time.
10.0 Exit Interview 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 March 28, 1989. No written inspection material was provided to licensee representatives during the inspection period.
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