IR 05000352/1985027
| ML20133N235 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 08/05/1985 |
| From: | Beall J, Gallo R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20133N234 | List: |
| References | |
| 50-352-85-27, NUDOCS 8508130415 | |
| Download: ML20133N235 (7) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No. 85-27 Docket No. 50-352 License No. NPF-27 Priority
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Category A
Licensee: Philadelphia Electric Company 2301 Market Street Philadelphia, Pennsylvania 19101 Facility Name:
Limerick Generating Station, Unit 1 & 2 Inspection Conducted: June 3 - 28, 1985 7/o/PS'
Inspector:
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[.E.Beall,ProjectEngineer date Other Participating Inspectors:
H. Kerch A. Krasopoulos b
hI Approved by:
R. M. Gallo, Chief, Reactor Projects date Section 2A
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, Inspection Summary:
Inspection Report for Inspection Conducted June 3 - 28,
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'1985 (Report No. 50-352/85-27)
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Areas Inspected:
Routine inspections by the Limerick project engineer and region-based inspectors of:
followup on outstanding items; followup on selected low power license conditions; plant tour; monthly surveillance observation; review of event reports and review of plant modifications associated with the control room survey. The inspection involved 58 hours6.712963e-4 days <br />0.0161 hours <br />9.589947e-5 weeks <br />2.2069e-5 months <br /> of
' onsite inspection by the Limerick project engineer and region-based inspectors.
Results: No violations were identified.
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DETAILS 1.0 Persons Contacted Philadelphia Electric Company J. Cotton, Maintenance Engineer J. Franz, Superintendent of Operations E. Gibson, Quality Assurance Engineer G. Lauderbach, Quality Assurance Engineer G. Leitch, Station Superintendent
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Also during the inspection period, the inspector discussed plant status and operations with other supervisors, engineers and operators in the PECo organization.
2.0 Followup on Outstanding Items 2.1 (Closed) ESW Valves Close on ESW Pump Trip This item was identified as a potential concern by an NRC contractor during a technical review of the ESW system as described in a Brookhaven National Laboratory Technical Review Report forwarded to the licensee by Region I on October 4,1984. The licensee responded to the concern in a letter dated November 7, 1984. The contractor reviewed the licensee response as described in a further report, forwarded by Region I on May 30, 1985, which left the item open pending Region I review.
The item involves the response of ESW valves to the tripping of an ESW pump. In the Limerick design, ESW system loads are supplied through valves arranged in parallel with each valve tied to the operation of its associated pump such that when the "A" pump starts, the "A" valves open. The contractor's concern is that the system response to the tripping of one pump is to close its associated valves, that is, tripping the "A" pump will cause the "A" valves to close. With the "B" pump running, both the "A" valves and the "B" valves would pass flow, therefore closing the "A" valves would reduce redundancy. The contractor noted that this system response might conflict with the guidance of IEEE-279 (1971) and IE Bulletin 80-06, "ESF Reset Controls."
The inspector conducted an independent review of the cited documents, reports, correspondence, and design drawings and finds that the ESW system, as installed, meets the applicable acceptance criteria as described in the Limerick Safety Evaluation Report (NUREG-0991), section 9.2.1.
Both IEEE-279 (1971) and IE Bulletin 80-06 pertain to the response of system components to the originating ESF actuation signal being reset. In this case the signal would be the signal to start the diesel generators. These
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signals are loss of emergency bus voltage and detection of a potential LOCA. The Limerick ESW system continues to operate in the emergency mode after the diesel generator start signal is reset until deliberate operator action is taken to secure the system or until protective devices internal to the system trip the ESW pumps.
These protective devices were described in the ESW system design basis which was reviewed prior to the issuance of the Limerick SER.
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This item is closed.
2.2. (Closed) Unresolved Item Pertaining to Preservice Ultrasonic Reports (352/84-29-02)
This item was identified during an NRC independent measurements inspection performed during the period June 25 - July 20,1984. The inspector found that preservice ultrasonic reports did not include the results of examination, acceptance or rejection, by the person responsible for interpreting the results of the examination. The licensee has issued and implemented procedure RT-09-70002 Rev. 1,
" Procedure for Review of Vendor Supplied NDJ Materials and Data."
The inspector reviewed the procedure and supporting documentation of the implementation and has no further questions at this time.
This item is closed.
2.3 (Open) Unresolved Item Pertaining to the Incorporation of Environmental Qualification Requirements into the Limerick Preventive Maintenance Program (352/85-03-09)
This item was identified during the Operation Assessment Team Inspection conducted during the period January 24 - February 1, 1985. The inspector found that preventive maintenance (PM)
requirements of the Environmental Qualification (EQ) Reports (for both electrical and mechanical equipment) had not yet been entered into the computer system for PM scheduling and tracking. The licensee has committed to complete this item by the first refueling outage and this commitment was documented in a letter from the licensee dated June 17, 1985. During the current report period the inspector reviewed the licensee's progress in incorporating the EQ requirements into the PM program computer scheduling and tracking system.
The licensee's approach has been to review the EQ Report, identify the component requirements, enter each item in the Computerized History and Maintenance Planning System (CHAMPS) computer, and time the output so as to allow sufficient lead time for generating the required approved procedures for each item.
The CHAMPS output is a Maintenance Request Form (MRF) which is the document used by the licensee to initiate work items. The inspector
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reviewed a sampling of MRF's printed by the CHAMPS EQ output and noted that'each contained the component name, location, required action, frequency, date last performed, required plant restraints, source of the requirements (in these cases the EQ Report), and other data used by the licensee such as account numbers. The inspector discussed the licensee's progress in placing all the EQ items into CHAMPS and the schedule for completion.
The inspector had no further questions at this time; this item remains open pending completion and further NRC review.
2.4 (Closed) Low Power License Condition 2.C(8)(a), Part 3, Control Room Enhancements The requirements associated with the Detailed Control Room Design Review (DCRDR) are contained in NUREG-0737, Supplement 1. On June 21, 1984, the NRC/NRR forwarded to the licensee the results of an audit conducted in December 1983 of the licensee's DCRDR. On June 25, 1984 and November 2, 1984, the licensee submitted reports which listed Human Engineering Deficiencies (HEDs) identified during the DCRDR.
Resolution of the HEDs was the subject of discussions between licensee and NRC personnel and of correspondence from the licensee.
In a June 10, 1985 letter, the licensee stated that all the DCRDR identified HEDs had been resolved.
The inspector conducted an independent review of the licensee's program to resolve the HEDs noted during the DCRDR. The inspector selected a sample of HEDs identified in both the June 25, 1984 and the November 2,1984 submittals. The inspector examined the HEDs for clarity of problem description, completeness of proposed resolution, identification of training or procedure requirements, and adequacy of assigned priority and schedule. The control room panels were examined to verify that the proposed changes to labels, buttons, color coding, etc. had been made and that the required training had been completed. The inspector reviewed the methodology employed by the licensee in conducting two 100% in-situ verification audits and discussed the acceptance criteria with the licensee's QA auditors.
No deficiencies were identified; this item is closed.
3.0 Plant Tour
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-Periodically during the inspection period, the inspectors toured the Unit I containment, the reactor enclosure, the control enclosure, the turbine enclosure, and the diesel generator enclosures. The inspectors exemined preventive maintenance, surveillance testing, equipment tagging, house-keeping, radiological control practices, portal monitoring, security, lighting, power block control points, fire protection equipment, and
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general plant operations. The inspectors routinely toured the control room to verify proper control' room manning, procedural compliance, safety system availability, and nuclear instrumentation operability. Interviews and discussions were routinely conducted with licensee operators and staff concerning the status of off-normal alarms, compliance with technical specifications, and general plant conditions.
During one tour the inspector noted that portions of the fire proof coating of the structural steel in the wall separating Rooms 430-431 and Rooms 434-435 had been chipped away exposing the underlying steel beam.
The rooms involved contain emergency auxiliary switchgear for Unit 1 (Rooms 434 and 435) and for Unit 2 (Rooms 430 and 431). The cabinets in the Unit 2 rooms supply certain loads for_ systems common to both units and are energized even though Unit 2 is still in the construction phase.
The inspector identified a similar condition in the Unit 1 cable spreading room. In this room the structural steel fire proofing had been degraded by the emplacement of weldments for cable tray supports. In response to the inspector's concerns, the licensee provided an analysis which concludes that under hypothetical fire scenarios the room tempera-tures remain below the level that could compromise the structural integrity of the steel due to low quantity of combustibles for the emergency auxiliary switchgear room and due to the operation of the installed water sprinklers for the cable spreading room. The inspector reviewed'the licensee's analysis and has no further questions at this time.
No violations were identified.
4.0 Monthly Surveillance Observation The inspector observed and reviewed surveillance test ST-6-095-904-1 "125 VDC Safeguard Battery Weekly Inspection." The inspector verified that the test had been properly approved by shift supervision, an approved procedure was being used, test instrumentation was calibrated, and test acceptance criteria were met.
No unacceptable conditions were identified.
5.0 Review of Licensee Event Reports The inspector reviewed the licensee event reports (LERs) listed below to determine if the information provided was accurate and submitted in a timely manner; if the event cause was properly identified and corrective actions were appropriate; if the report described a potentially generic issue; and if the report satisfied the licensee's reportability requirement *
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The following LERs were found to be acceptable:
LER No.
Event Date Description 85-36 3/27/85 Removal of telephone conduit fire seals without fire watches in place 85-37 3/26/85 Various ESF actuations including LPCI injection and diesel start due to valving error during instrument line backfilling 85-38 3/28/85 Inadvertent isolation of RHR shutdown cooling 85-39 3/30/85 Inadvertent isolation of containment atmosphere sampling 85-40 3/30/85 Various ESF actuations including CS & LPCI injection and diesel start due to valving error during surveillance testing 85-41 4/1/85 Spurious Reactor Enclosure HVAC isolation 85-42 4/2/85 Inadvertent control room HVAC isolation due to a broken tape in the chlorine detector 85-43 4/2/85 Failure to meet hourly fire watch requirements of Technical Specifications 85-44 4/10/85 Inadvertent control room HVAC isolation due to hand held radio transmitter 85-45 4/1/85 Failu*;e to meet SPDS schedule for opedability 85-46 4/23/85 Inadvertent scram signal due to valvidg error 85-47 4/20/85 Three reactor instrument piping excess flow check valves inoperable while in startup mode 85-48 4/30/85 Inadvertent isolations of engineered safety features due to a blown fuse 85-49 5/1/85 Inadvertent isolations of engineered safety features due to a blown fuse during surveillance testing
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85-50 5/6/85 Inadvertent control room HVAC isolation due to a broken tape in the chlorine detector 85-51 5/7/85 Inadvertent RWCU isolation during surveillance test 6.0 Exit Meeting The NRC inspector discussed the issues and findings in this report throughout the inspection period and at an exit meeting held with Messrs.
G. Leitch and J. Franz on June 26, 1985. At this meeting the represen-tatives of the licensee indicated that the items discussed in this report did not involve proprietary information. No written material was provided to the licensee during this period.