IR 05000352/1986011
| ML20209G775 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 09/08/1986 |
| From: | Eselgroth P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20209G751 | List: |
| References | |
| TASK-1.C.1, TASK-1.C.8, TASK-TM 50-352-86-11, NUDOCS 8609150071 | |
| Download: ML20209G775 (13) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
i Report No. 86-11 Docket No. 50-352 License No. NPF-39 Priority --
Category C Licensee: Philadelphia Electric Company 2301 Market Street i
Philadelphia, Pennsylvania 19101 Facility Name:
Limerick Generating Station, Unit 1 Inspection Conducted: June 1 - July 31,1986 Inspectors:
E. M. Kelly, Senior Resident Inspector S. D. Kucharski, Resident Inspector B. M. Hillm E
, Reactof ngineer j
Approved by:
f-F' %
P.V. Eselg h, Chief, Reactor Date Projects S tion 2A Inspection Summary:
Inspection Report No. 50-352/86-11 for Inspection
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Conducted June 1 - July 31, 1986.
Area Inspected:
Routine daytime and backshift inspections (314 hours0.00363 days <br />0.0872 hours <br />5.191799e-4 weeks <br />1.19477e-4 months <br />) of Unit 1 by the resident inspectors consisting of followup on outstanding items; system walkdown of selected systems using PRA guidance; plant tours including fire protection measures; maintenance and surveillance observations; and review of LERs and periodic reports.
No violations were identified.
8609150071 860908 PDR ADOCK 05000352 G
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DETAILS 1.0 Persons Contacted Philadelphia Electric Company J. Corcoran, Engineer-In-Charge, Field QA J. Doering, Superintendent of Operations R. Dubiel, Senior Health Physicist P. Duca, Technical Engineer J. Franz, Station Manager G. Leitch, Superintendent, Nuclear Generation Division Also during this inspection period, the inspectors discu:, sed plant status and operations with other supervisors and engineers in the PECO, Bechtel and General Electric organizations.
2.0 Licensee Action on Previous Inspection Findings 2.1 (Closed) Unresolved Item 86-04-01; Corrective Actions for Calibration Testing Exceeding 2-Hour Technical Specification Limit.
During the performance of ST-2-076-401-1, which is a calibration functional test performed every 92 days for instrumentation associated with Reactor Enclosure and refueling floor isolation system, a two-hour time limit was exceeded due to ineffective communication between the I&C technician and operator and other distractions present due to control room activities.
Based on the licensee's evaluation of this event an ST Timer was placed in the control room. The timer will have a log associated with it, in which, the technician will record the ST number and their name.
The timer will be set to alarm prior to the required time limit which will enable the control room staff to notify. the appropriate' technician of the time constraint. This item is considered closed.
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2.2 (Closed) Inspector Follow-up Item 86-09-02; Review of the Storage
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and Testing of the MSRV's.
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The inspector reviewed the following Field Purchase Orders:
Purchase Order No.
Valves Serial No.
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-3 P.O. #8031-F-76278 501, 504, 505, 507, 508, 510, 511,
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512, 514
.P.O. #8031-F-76827 529, 530, 531, 533, 534, 535 P.O. #80,31-F-65216 502, 509
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Based on this review the inspectors noted that all the valves were stored according to the vendors recommendations.
The valves were reworked by the Target Rock Corporation based on recommendations presented in Service Information Letter (SIL) Number 196, September 30, 1976. The SIL recommended pilot seal leakage detection, pilot filter installation, second stage modification and pressure set point adjustments. After medifications were made the valves were reassembled and on January 11, 1985 the Target Rock Corporation performed a set point check and leakage test using nitrogen. This was followed by a steam test, the results of the steam test are as follows:
Serial No.
Nominal Lift Pressure SRV Set Pressure (PSG)
(PSIG)
501 1130 1125 504 1130 1126 505 1140 1136 507 1150 1151 508 1140 1132 510 1150 1154 511 1150 1142 512 1130 1123 514 1140 1138 529 1140 1142 530 1140 1136 531 1130 1120 532 1130 1123 533 1130 1137 534 1150 1150 535 1150 1152 502 1130 1133 509 1140 1138 All valves lifted within their design band, the inspector has no further questions, and considered this item closed.
2.3 (Closed) TMI Action Item I.C.1, SER 13.4 Supplement No. 1, Short-Term Accident and Procedure Review Limerick's Emergency Operating Procedures (EOP) are based upon the BWR Owners Group Emergency Procedures Guidelines (EPG), which are contained in General Electric Topical Report NED0-24934, Revision 2.
In SER ORAFT SUPPLEMENT 2, Section 13.5.2.3, the BWR owners group EPGs were reviewed and accepted by NRR.
Limerick's E0Ps were developed utilizing a Procedures Generation Package (PGP) which includes a plant Specific Technical guidelines, writer's guide, E0P Validation program, and a training implementation program.
The NRC staff has concluded the PGP for Limerick meets the guidelines of
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supplement I to NUREG-0737 and provides acceptable methods for accomplishing the objectives of NUREG-0899.
The Inspector reviewed the following licensee symptomatic procedures to determine if they were in agreement with the BWR Generic Emergency procedures Guidelines.
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T-101 Reactor Pressure Vessel Control Rev 0, 12/30/83
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T-102 Containment Control Rev 0, 12/30/83
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T-111 Level Restoration Rev 0, 12/30/83 T-112 Emergency Blowdown Rev 0, 12/30/83
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T-113 Blowdown Cooling Rev 0, 12/30/83
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T-114 Spray Cooling Rev 0, 12/30/83
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T-115 Alternate Shutdown Cooling Rev 0, 12/30/83
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T-116 Reactor Pressure Vessel Flooding Rev 0, 12/30/83 T-117 Level / Power Control Rev 0, 12/30/83
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No descrepancies were noted.
Training in use of this procedure consists of a combination of classroom instruction, and simulator exercises. The effectiveness of the training has been reflected in the emergency preparedness appraisal report.
This item is closed.
2.4 (Closed) TMI Action Item I.C.8; Pilot Monitoring of Selected Emergency Procedures for NTOL.
The pilot monitoring program was established on an interim basis for evaluating Licensee's Emergency Operating Procedures, pending NRC Staff approval of Generic Technical Guidelines and development of a long term program for upgrading Emergency operating procedures. As noted in paragraph 2.3 above, the NRC staff has accepted the BWR Ow:.cs Group Emergency Operating Procedures Guidelines, and in Generic Letter 82-33, the NRC staff issued it's long-term plan for the upgrading of emergency procedures.
Limerick's SER, Section 13.5.2.3 calls for no further action to be taken on this item. This item, therefore, is closed.
3.0 Review of Plant Operations 3.1 Summary of Events The plant achieved full rated powar operation on June 21, 1986 after completion of a planned outage. Operation continued until July 4, 1986 when the plant was shutdown due to an increase in the Drywull floor drain leakage nearing 5.0 gpm. The increase in drywell leakage was attributed to leakage past the packing of the
'A' LPCI Manual isolation valve.
In addition to replacing the
' A" LPCI valve packing, maintenance replaced the ' A' recirculation pump seal and inspected the MSIV external springs as recommended by General Electric.
On July 13 the plant returned to full power operatio.
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3.2 Operational Safety Verification The inspector toured the control room daily to verify proper manning, access control, adherence to approved procedures, and compliance with LCOs.
Instrumentation and recorder traces were observed and the status of control room annunciators was reviewed.
Nuclear instrument panels and other reactor protective systems were examined.
Effluent monitors were reviewed for indications of releases.
Panel indications for onsite/offsite emergency power sources were examined for automatic operability.
During entry to and egress from the protected area and vital island, the inspector observed access control, security boundary integrity, search activities, escorting and badging, and availability of radiation monitoring equipment including portal monitors. No unacceptable conditions were found.
The inspector reviewed shift superintendent, control room supervisor, and operator logs covering the entire inspection period.
Sampling reviews were made of equipment trouble tags, night orders, and the temporary circuit alteration and LCO tracking logs.
The inspector also observed shift turnovers during the period. The operations activities were observed for conformance with the applicable procedures and requirements; no unacceptable conditions were noted.
3.3 Station Tours The inspection toured accessible areas of the plant throughout this inspection period, including: the unit 1 reactor and turbine-auxi-liary enclosures; the main control and auxiliary equipment rooms; emergency switch gear and cable spreading rooms, and the plant site parameter. During these tours, observations were made relative to equipment condition, fire hazards, fire protection, adherence to i
procedure, radiological controls and conditions, housekeeping, security, tagging of equipment, ongoing maintenance and surveillance and availability of redundant equipment. No unacceptable conditions were identified.
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3.4 System Inspections Using PRA Guidance Reactor Core Isolation Cooling (RCIC) System The inspector performed a modified system walkdown of the Reactor Core Isolation Cooling (RCIC) System utilizing the method prescribed in a study prepared for the NRC by Brookhaven National Laboratory. The walkdown consisted of first, in the control room, to see if the system is operable and if all the valves are in their correct position.
Secondly the inspector went to the 177 foot elevation in the Reactor enclosure (RCIC Room) to verify that the RCIC pump suchtion valve from the Condensate Storage Tank (CST), a manual valve, is open and locked. No violations were identifie.
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4.0 Event Followup 4.1 Control Room Emergency Fresh Air System Actuation At 10:15 a.m. on June 3 both toxic gas analyzer channels alarmed high which resulted in the manual initiation of the Control Room Emergency Fresh Air System (CREFAS) as required by Special Event Procedure, SE-2, " Toxic Gas".
The control room was cleared of all non-essential
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personnel and self-contained breathing apparatus were donned in accordance with emergency procedure, EP-101. Chemistry performed a habitability check of the control room and sampled the control enclosure HVAC intake plenum in accordance with EP-330, " Emergency Response Facility Habitability". Based on the results of that check no abnor-mal presence of Ethylene Oxide, Phosgene or Ammonia was detected in the control room or control enclosures HVAC intake plenum. At that point the toxic gas analyzers were declared inoperable.
Subsequent investigation by the licensee found that maintenance being performed on the drywell chillers involved venting of retrigerant which contained freon gas.
Freon gas behaves similar to the infrared characteristics of ethylene oxide, one of the six toxic gases monitored by the analyzers. The freon refrigerant was discharged and detected by the analyzer in the Control Enclosure HVAC intake plenum.
The resident inspector discussed this event with the licensee who has agreed to revise the system procedure S87.3A Drywell Chiller maintenance to include a precaution statement which requires control room notification prior to venting freon during charging of the Drywell Chillers. The procedure will also include detailed steps for the preliminary evacuation and dehydration of the Drywell Chiller evaperator. The licensee has also agreed to investigate, with the aid of the vendor, an alternate method of venting freon during the evacuation and. dehydration process. The inspector had no further questions.
4.2 Loss of Reactor Protection System (RPS) Power Power was lost to the 'B' reactor protection system channel at 8:57 a.m. on June 9 while attempting to transfer power supplies from
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I the alternate supply to the normal static inverter following maintenance on the RPS-UPS inverter.
This de-energization resulted in the actuation of the outboard Nuclear Steam Supply Shutoff System (NSSSS) isolation, a half scram signal, and the isolation of the
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Containment Instrument Gas and the Drywell Chilled water systems.
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The Reactor Water Cleanup (RWCU) system which was operating in the shutdown cooling mode at the time of the event, tripped due to the NSSSS logic de-energization.
In addition, the following systems also received isolation signals but due to plant conditions (OPCON4), no equipment actuations occurred:
Reactor Enclosure HVAC; Refueling floor HVAC; Residual Heat Removal (RHR) shutdown cooling; and Containment Sampling Systems.
This event was caused by an incomplete connection between a logic card and its mating connector.
The incomplete connection is the result of an alignment problem with the pin connectors on the logic card. When the power supply was transferred back through the inverter, the logic card with the incomplete connection did not permit the transfer.
Investigation by the licensee showed that the design of the pin connectors on the Gate Drive Booster Card is such that improper alignment of the connectors can easily occur during installation of the card.
The licensee has proposed a procedural change to RT-11-02023 Revision 0, " Testing and Calibrating of the 30KVA Static Inverters IAD160 and 1BD160" and RT-11-02024 Revision 0, " Testing and Calibrating of the 30 KVA Statis Inverters 100161 and 00D162",
to test the continuity of the connections of the Booster Cards, whenever they are installed, before attempting a power source transfer. The inspector had no further questions.
4.3 Chlorine Tank Bulge At 10:20 a.m. on June 10, the Auxiliary Operator (AO) notified the Control Room of an unusual bulge in the concave portion of the 'C'
chlorine tank. The
'C'
chlorine tank is part of a series of pin tanks used to chlorinate the cooling tower water.
Each tank is capable of holding 2000 pounds of chlorine. As a precautionary measure the control room manually isolated the 'B' chlorine tank train and all the tanks in that series were valved closed to eliminate the possibility of their draining should the problem tank fail or leak. The licensee then waited until after the Unit 2 construction workers left the site to drain the tank. The chlorine was transferred to an empty tank and the "C" tank was removed from the site.
The vendor representative attributed the problem to the tank being overffiled.
The concave portion of the tank is designed to pop out (safety feature) if overfilled or overpressurized.
The licensee plans to prepare an upset report as a followup to this event.
No violations were identified.
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5.0 Licensee Reports 5.1 In-Office Review of Licensee Event Reports The inspector reviewed Unit 1 LERs submitted to the NRC Region I office to verify that details of the event were clearly reported, including the accuracy of description of the cause an adequacy of corrective action.
The inspector determined whether further information was required from the licensee, whether generic implications were involved, and whether the event warranted on-site followup. The following LERs were reviewed:
LER NUMBER REPORT DATE SUBJECT 86-017 May 16 Internal Fire Protection Seals Missing in Electrical Gutters
!86-021 June 2 Inadvertent ESF Actuation of SGTS due (Note A)
to Improper Use of Jumpers86-022 June 5 Spurious Control Room Emergency Fresh Air System Actuation 86-023 June 11 Engineered Safety Feature (ESF)
(Note A)
Actuation During Surveillance Test 86-024 June 11 Unplanned Isolation of Reactor Enclosure HVAC due to Personnel Error 86-025 June 11 Communication Error during Testing causes ESF Actuation 86-026 June 13 Unplanned Isolation of the Reactor Enclosure HVAC Due to a Blown Fuse 86-027 June 20 Technical Specification Fire Watch Violation Due to Personnel Error 86-028 July 3 Control Room Emergency Fresh Air System Actuation due to False Toxic Gas86-029 July 9 ESF Actuation Caused by an Incomplete Connection on a Logic Card
'86-031 July 17 Special Report - Exceed Local Leak Rate Test Allowable limits.
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.86-032 July 21 Overdue Surveillance Test due to (Note A)
Personnel Error Notes:
A: Addressed in Detail 5.2 of this report.
5.2 Onsite Followup of Licensee Event Reports For those LERs selected for onsite followup as noted in Section 5.1, the inspector verified the reporting requirements of 10 CFR 50.73 and technical specifications had been met, that appropriate corrective action had been taken, that the event was reviewed by the licensee, and that continued operation of the facility was conducted in accordance with Technical Specification limits.
5.2.1 LER 86-021:
Inadvertent Engineered Safety Feature (ESF)
Actuation of SGTS On May 3, 1986 while the plant was in Operational Condition 3, hot shutdown, an inadvertent initiation of the Engineered Safety Feature (ESF) systems including the Standby Gas Treatment System (SGTS) and the Reactor Enclosure Recircu-Tation System (RERS) occurred during the performance of a Surveillance Test due to an improperly installed jumpers and a blown fuse. The probable cause of the fuse burning out is the inadvertent grounding of the jumper by the test engineers.
The inspector reviewed the corrective actions and held discussions with the licensee about preventative actions taken. The test engineers involved were counseled to the importance of properly installing jumpers during surveillance tests. The inspector had no further questions.
5.2.2 LER 86-023:
Inadvertent Engineered Safety Feature (ESF)
Actuation during Surveillance Test On May 12, 1986, while the plant was in Operation Condition 4, cold shutdown, various ESFs actuations occurred when an initiating signal (Loss of Coolant Accident) was generated during the performance of surveillance test ST-s-42-462-1, Accident Monitoring -
Reactor Vessel Pressure Calibration / Functional Test (PT-42-103B,XR-42-aR6238).
The 'B' Core Spray (CS) pump and the 'B' Residual Heat Removal (RHR) received start signals but were blocked out of service at the time of the event. The false signal was caused by a personnel error when the technician performing the test used an
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alternative test connection on PT-42-103B when venting the
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transmitter.
PT-42-103B has a common reference leg with other instrumentation including a Reactor Protection System (RPS) level transmitter and two CS/RHR level transmitter.
The inspector reviewed the corrective actions plans with the licensee which a change to Surveillance test ST-s-42-462-1.
The ST now provides a figure showing where the connection is to be made and a corresponding step in the procedure relating to the figure. The inspector had no further questions.
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5.2.3 LER 86-032: Overdue Surveillance Test Due to Personnel Error On June 21, 1986, with the Unit in Operational Condition (OPCON 1) at 100% power, it was discovered there was a failure to comply with Technical Specifications because daily Surveillance Test (ST-6-043-320-1) Daily Jet Pump and APRM Flow Unit Operability Verification had not been performed within the required time.
The ST was performed as required during the day shift on June 21, 1986 but due to changing reactor recirculation flow the results of the test were invalid and had to be redone. At the shift turnover meeting the next shift was not informed that the ST had not been completed satisfac-tory. The overdue ST was satisfactorily performed by the night shift, who discovered the problem.
The resident reviewed the procedural changes that' were made by the licensee.
ST-6-107-590-1, Daily Surveillance Log-0PCONS 1, 2, and 3, Revision 13, July 14, 1986 now has
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a sign off for the performance of ST-6-043-320-1 that it has been satisfactorily completed. The licensee has also changed ST-6-043-320-1 by adding a sign off step to assure that the recirculation flow remains constant during the surveillance test. The inspector had no further questions.
5.3 Review of Periodic and Special Reports Upon receipt, periodic or special reports submitted by the licensee were reviewed by the inspector. The reports were reviewed to determine that the report included the required information, that test results and/or supporting information were consistent with design predictions and performance specifications, and whether any information in the report should be classified as an abnormal occurrence.
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x The following periodic and special reports were reviewed:
Special Reporting Requirement - Inoperable Seismic Monitoring
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at Limerick Generating Station, July 31, 1986 Monthly Operating Report for April, 1986
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Monthly Operating Report for May, 1986
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Mcnthly Operating Report for June,1986
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These reports were found acceptable.
6.0 Maintenance and Surveillance Activities 6.1 Maintenance on Mechanical Seal of 'A' Recirculation Pump During this inspection period the licensee performed maintenance activity on the 'A' Recirculation pump, due to the failure of the first stage of the two stage mechanical seal.
Pressure in the second
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stage was 1000 psig which indicated the first stage was leaking. On July 7, the licensee disassembled the recirculation pump and removed the mechanical seal, replaced it with one from Peach Bottom Atomic Power Station (PBAPS).
The first seal (S/N 0739) to arrive failed I
the preliminary leak test, and another seal (S/N 1032) was obtained.
The inspector reviewed the maintenance request form package (MRF #
j 8604011) for the second seal (S/N 1032) which included the completed
procedure M-043-002, ' Maintenance Procedure for the Replacement of a Reactor Recirculation Pump Mechanical Seal', Revision 1, July 6, 1986. Work was performed in accordance with the procedure.
Maintenance data record forms were properly documented and approved, and all the required hold points were checked by Q.C. and properly signed off as required.
The inspector witnessed the checking of the mechanical seal for
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proper axial movement, which is a check for proper assembly and also the testing of seal (S/N 1032) for proper pressure reduction and flow prior to installation into the pump.
Personnel performing the work were knowledgeable and adequately performed their assignments.
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i The licensee's QA department identified several unacceptable findings
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concerning the shipment and testing of the replacement seals.
First,
the seals were transferred from PBAPS to Limerick without proper
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administrative controls being established for the transfer. Secondly,
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regarding the first seal, (S/N 0739), the axial distance from the
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bottom face of the spring holder to the bottom of the
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follower, was not within the acceptance criteria.
Finally, when seal, (S/N 0739), failed the pressure reduction and flow test, the seal was disassembled without the appropriate generation of a MRF, notification of QC, and documentation of the disassembly. The NRC inspectors will review the resolution of the QA concerns in future inspections.
6.2 Surveillance Activities The inspector observed performance of or reviewed the result of the following tests:
ST-4-103-011-1, Accessible Snubber Visual Inspection - RHR #11,
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Revision 1, May 5, 1986 ST-4-103-650-1, Inaccessible Sr.ubber Visual Inspection - Recirc
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ST-2-078-406-0, Chlorine Detection System - Control Enclosure
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Air Intake Chlorine Detector, Channel C Calibration / Functional Test (AITS-78-016C, AI-78-016C), Revision 1, July 22,1986
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ST-5-078-609-0, Anacon Chlorine Probe Replacement and
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Functional Test (AC-78-016C), Revision 1, July 7, 1986 ST-5-078-349-0, Adding Electrolyte to Anacon Chlorine Probe
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(AE-78-016C), Revision 1, June 24, 1986 ST-2-042-462-1, Accident Monitoring - Reactor Vessel Pressure
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Calibration / Functional Test (PT-42-103B, XR-42-1R623B),
Revision 5, June 19, 1986
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ST-6-107-885-1, Thermal Limits Determination, Revision 9, May 16, 1986
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The tests were observed to determine that procedures conformed
to Technical Specification requirements; proper administrative controls and tagouts were obtained prior to testing; testing was i
performed by qualified personnel in accordance with approved procedures and calibrated instrumentation; test data and results were accurate and in accordance with Technical Specifications; and, equipment was properly returned to service following testing.
The inspector also reviewed RT-4-103-480-0, Snubber Visual Inspection Orientation Program, Revision 0, May 1, 1986 which is
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used to familiarize inspection personnel with typical snubber installations and provide instructions for proper inspections.
The inspector witnessed the inspection of several snubbers by the licensee and observed practices that were pointed out in the training program.
No unacceptable conditions were identified.
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i 7.0 Exit Meeting
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i The NRC resident inspector discussed the issues in this report throughout
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the inspection period, and summarized the findings at an exit meeting
held with Mr. John Franz and others of your staff on August 8,1986. At
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this meeting, the licensee's representatives indicated that the items j
discussed in this report did not involve proprietary information, t
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