IR 05000219/1987021
| ML20236E602 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 07/22/1987 |
| From: | Anderson C, Woodard C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20236E572 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.E.4.2, TASK-2.F.1, TASK-TM 50-219-87-21, GL-82-05, GL-82-5, NUDOCS 8708030022 | |
| Download: ML20236E602 (10) | |
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U.S. NUCLEAR REGULATORY COMMISSION
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REGION I
Report No. 87-21 Docket'No.
50-219 License No.. DPR-16-Priority --
Category C
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. Licensee: GPU Nuclear Corporation Post Office Box 338 Forked River, NJ 08731
, Facility Name: Oyster Creek Nuclear Generating' Station
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Inspection At: Oyster Creek Inspection Conducted: June 8-12, 1987 Inspectors:
M Mad M
7/.20 /8 7 C. H. Woodard, Reactor Engineer d'a te
' Approved by:
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C. J/. Anderson, Chief, Plant Systems date
'Section,' Engineering Branch, DRS Inspection Summary:
Inspection June 8-12, 1987 (Report No. 50-219/87-21)
Areas Inspected: An announced routine safety' inspection was conducted by j
- Mr. C. H. Woodard of this office at the Oyster Creek Nuclear Generating Station for the purpose of reviewing the licensee's implementation of specific sections of NUREG-0737, Clarification of TMI Action Plan Requirements relative to containment isolation dependability and to certain accident-monitoring instrumentation. The inspection involved 38 hours4.398148e-4 days <br />0.0106 hours <br />6.283069e-5 weeks <br />1.4459e-5 months <br /> onsite by one region-based inspector.
Results: No violations or deviations were identified.
l 8708030022 G70724 PDR ADOCK 05000219
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DETAILS 1.0 Persons Contacted GPU Nuclear Corporation P. Fiedler, Vice President and Director of Operations
- J. Barton, Deputy Director of Operations
- J. Sullivan, Director of Plant Operations
- S. Fuller, Quality Assurance Manager of Operations
- A. Rowe, Director of Plant Engineering
- B. DeMerchant, Licensing Engineer D. Dryden, Document Control Specialist J. Kawalski, Manager Licensing D. MacFarlene, Manager, Audits United States Nuclear Regulatory Commission
- W. Bateman, Senior Resident Inspector
- J. Wechselberger, Resident Inspector
- Attendees at the exit meeting on June 12, 1987.
2.0 Purpose The purpose of this inspection was to verify and validate the adequacy of the licensee's implementation of the following tasks identified in NUREG-0737, Clarification of TMI Action Plan Requirements:
Task No.
Title II.5.4.2 Containment Isolation Dependability Positions 1, 2, 3 and 4.
II.F.1.4 Containment Pressure Monitor II.F.1.5 Containment Water Level Monitor II.F.1.6 Containment Hydrogen Monitor
i 3.0 TMI Action Plan Generic Criteria and Commitments The licensee's implementation of the tasks specified in Section 2.0 were reviewed against criteria contained in the following documents:
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NUREG-0737-Clarification of TMI Action Plan Requirements, dated l
November 198.
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NUREG-0578-TMI-2 Lessons learned Task Force Status Report and Short
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Term Recommendations, dated July 1979.
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, Regulatory Guide 1.97, Rev. 3-Instrumentation for Light-Water-Cooled
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Nuclear Power Plants to Assess Plant and Environs Conditions During and Following and Accident.
DyIter Creek Nuclear Generating Station, Updated Final Safety
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Ana' lysis Report, dated December 1984.
l NUREG-0800 U.S. Nuclear Regulatory Commission Standard Review Plan:
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Section 6.2.4 Containment Isolation System, Rev. 2 dated July 1981.
i Section 7.1 Instrumentation and Controls, Rev. 3 dated February 1984.
Table 7-2 TMI Action Plan Requirements for Instrumentation and Control Systems Important to Safety, Rev. 0 July 1981.
Letter from Eisenhut, NRC to all operating power plants, dated
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October 30, 1979.
Generic Le,tter 82-05, Letter from Eisenhut, NRC to all operating
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power plants, dated March 14, 1982.
Licensee submittals to NRC on post-TMI related issues, dated May 6
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and June 30, 1983 and April 3, 1985.
NRC Safety Evaluations for NUREG-0737 dated September 29, 1983 and
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June 19, 1985 In addition specific review criteria and/or commitments relative to each of the task item numbers are included in Attachment 1 of this report 4.0 Inspection The inspection was organized into five distinct phases as follows:
1.
A review of NUREG-0737 task requirements and the licensee's plans for satisfying the requirements.
2.
A review of the licensee's documentation to perform the modifications.
3.
Physical walkdown inspection of the modifications including verification of the modification work packages, inspections, and tests,
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A review of the operational procedures and training for both the 4.
r surveillance and gperating personnel.
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A review of pland record drawings and documents to ascertain that I
key drawings and procedures reflecting the modification are in the control room for operations and are on file for trouble shooting and s
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maintenance.
4.1 Containment Isolation Dependability, NUREG-0737 Task II.E.4.2 F~'sitions1,2,3and4.
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Position 1 Position 1 specifies that the Containment Isolation System should address the recommendations of NUREG-0800, Standard Review Plan Section 6.2.4 in that a diversity of parameters hould be sensed for-the initiation of containment isolation.
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The inspector confirmed from a review of the FSAR, Technical Speci-fications, Operating Procedures and drawings identified in Section 3 and Attachment 1 that the licensee has incorporated a diversity of containment isolation signal parameters into the Containment Isola-tion System.
In addition to the high radiation signal parameters the licensee's system now includes the activating signals " Low-Low
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Reactor Water Level" and "High Drywell Pressure".
Within the scope )of this inspection, the licensee has completed the implementation of this Position 1 Task.
4.1.2 Position 2 Position 2 requires that "All plant personriei shall give careful consideration to the definition of essential and non essential sys-tems, identify each system determined to be essential, describe the basis for selection of each essential system, modify their contain-ment isolation designs accordingly and report the results of the reevaluation to the NRC."
The inspector confirmed licensee compliance with this position by a l
review of the following:
Documentation listed in Section 3.0 and Attachment 1.
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Licensee April 10, 1980 letter to the NRC which provided the basis, definition and identity of systems determined to be classified as non-essential and described the modifications being made to provide for automatic isolation.
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Reactor Protection Drawing 237E566, Revision 13 which includes i
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the currently approved modified design for the containment (
isolation valves and the actuating circuitry for these valves.
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Oyster Creek FSAR Section 7.3 which includes a description of the modified Containment Isolation System.
Within the scope of this inspection, the licensee has completed the implementation of this Position 2 Task.
4.1.3 Position 3 Position.3 requires that "All n)nessential systems shall be
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automatically isolated by the containment isolation signal."
J The inspector confirmed licensee compliance with this position as
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follows.
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A review of the documentation listed in Section 3.0 and Attachment 1.
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A review of the licensee's April 10, 1980 letter to the NRC in which all non-essential Containment Isolation Systems and
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penetrations are identif.3d.
A review of the licensee's automatic containment isolation
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system design including the actuating signals and the non-essential containment penetrations which are automatically isolated as described in Reactor Protection Drawing 237E566, j
Revision 13.
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A review of GPU Oyster Creek acceptance tests and turn-over documentation listed in Attachment 1.
Within the scope of this inspection, the licensee has completed i
the implementation of this Position 3 Task.
4.1.4 Position 4 Position 4 requires that "The design of control systems for automatic containment isolation valves shall be such that resetting the isola-
tion signal will not result in the automatic reopening of containment i
isolation valves.
Reopening of containment isolation valves shall require deliberate operator action."
i The inspector confirmed licensee compliance with this position as follows:
A review of documentation listed in Section 3.0 and Attachment 1..
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A review of the licensee's automatic and manual electrical control system for the automatically-actuated to close contain-ment isolation valves as described in Reactor Protection Drawing 237E566, Revision 13. This drawing shows that any one of the containment isolation signals will automatically initiate closure of the containment isolation valves.
Once ',he signal occurs it seals in through a relay circuit such that even if the isolation signal is lost the isolation circuit is still acti-vated. As long as the isolation signal is present, the isola-tion circuit cannot be reset to open either from the reset switch on the local control panel or the switch in the control room.
Re-opening the isolation valves after the isolation signal is lost requires deliberate operator action.
A review of the acceptance tests and turn-over acceptance
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documentation listed in Attachment I which verified proper operation.
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A review of Oyster Creek Station Operating Procedure 312, Revision 39 Reactor Containment Integrity and Atmosphere Control.
Physical inspection of both the local and remote control panels
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for the controls and instrumentation for these isolation valves.
Within the scope of this inspection, the licensee has completed the i
implementation of this Position 4 Task.
4.2 Additional Accident Monitoring Instrumentation NUREG-0737 Section II.F.1 (NUREG-0660 Subparts 4, 5 and 6).
4.2.1 Subpart 4 Containment Pressure Monitor Subpart 4 position specifies that each operating reactor shall be provided with containment wide range pressure measurement and monitoring with continuous indication in the control room.
The inspector confirmed licensee compliance with this position as follows:
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A review of the requirement for this monitor as contained in the criteria documentation in Section 3.0 and in the licensee documentation listed in Attachment 1.
A review of the specific commitments for the pressure monitor
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included in licensee submittals to the NRC dated May 6 and June 30, 1983.
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A review of the NRC approval of the licensee submittals in the
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NRC Safety Evaluation dated September 29, 1983.
A review of installation specifications drawings, tests,
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acceptance and turn-over documents listed in Attachment 1.
Physical inspection of the local and remote indicating / control
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panels for the containment pressure monitor.
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A review of the record drawings and operating procedures listed in Attachment 1.
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Within the scope of this inspection the licensee has completed the implementation of this Subpart 4 Task.
4.2.2 Subpart 5 Containment Water Level Monitor Subpart 5 position requirc; that each operating reactor shall be provided with a wide range containment water level monitor with continuous indication in the control room.
The inspector confirmed licensee compliance with this position as
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follows:
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A review of the requirements for the monitor as contained in I
the criteria documents listed in Section 3.0 and in the licensee
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documentation listed in Attachment 1.
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A review of the specific commitments made by the licensee for the water level monitor included in submittals to the NRC dated May 6 and June 30, 1983 and April 3, 1985.
A review of the NRC approvals of the licensee submittals in NRC
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Safety Evaluations dated September 29, 1983 and June 19, 1985.
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Physical inspection of the local and remote indicating / control
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panels for the water level monitor.
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A review of the record drawings and operating procedures listed in Attachment 1.
Within the scope of this inspection, the licensee has completed the i
implementation of this Subpart 5 Task.
l 4.2.2.1 Containment Water Level Instruments Sensing Tubing
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The Containment Water Level Monitoring System is a dual channel system with separate instrumentation, wiring, cable and power supplies in order to meet the single failure acceptance criteria.
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However, the inspector noted that the water level impulse signal to both channels of instrumentation is from a single upturned water
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collecting tube in the bottom of the torus which splits outside the torus to provide each instrument with a level signal (approved by NRC June 19, 1985 letter).
Scrap, debries, etc. could collect in this up-turned sensing signal tube and go unnoticed during calibrations and operation of these instruments since the torus water level is generally a static level. Obstructions in the sensing line could impair the dynamic response of these instruments in the event of an accident.
Each of the dual channel level instrument detectors located outside the containment has capped stub lines which could be backflushed to assure that the sensing line is open. The licensee stated they would investigate means to assure that the water level impulse signal line is not subject to a single failure' causing the loss of both instruments (87-21-01).
4.2.3 Subpart 6 Containment Hydrogen Monitor System Subpart 6 position requires that each operating reactor shall be provided with continuous indication of containment hydrogen concentration in the control room.
The inspector confirmed licensee compliance with this position as follows:
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A reviaw of the requirements for the hydrogen monitoring system as contained in the criteria documents listed in Section 3.0 and in the licensee documentation listed in Attachment 1.
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A review of the specific commitments made by the licensee for the hydrogen monitoring system included in submittals to the NRC dated May 6 and June 30, 1983.
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A review of the NRC approvals of the licensee submittals in NRC Safety Evaluations dated September 29, 1983 and June 19, 1985.
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A review of the installation specifications, drawings, tests, acceptance and turn-over documents listed in Attachment 1.
i Physical inspection of the local and remote indicating / control
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panels for the Hydrogen Monitoring System.
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A review of the record drawings and operating procedures listed j
in Attachment 1.
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Within the scope of this inspection, the licensee has completed the implementation of this Subpart 6 Task.
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4.2.3.1 Operational Problems with the Hydrogen Monitoring
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System Equipment
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A review of operations / calibration / surveillance history of the Hydrogen Monitoring System for the period November 1984 to April 1986 revealed that this system is subject to drifting.
The system is supplied by Comsip.
The vendor Comsip acknowledges that the equipment can be expected to undergo approximately 2% instrument drift per week (The Hydrogen Monitor is the Comsip Model K-IV Unit).
The licensee's technical specification (and the NRC standard technical specification)
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However, due to the fact that the instru-ment is subject to drift and based upon experience, the licensee is conducting a system surveillance once a week in order to maintain a higher confidence level in the system accuracy. This weekly surveillance by the licensee revealed 54 out of tolerance readings for this instrumentation during the November 1984 to April 1986 period.
The licensee has attempted to improve the Hydrogen Monitoring System and is actively seeking more accurate / reliable equipment from Comsip or others to either modify or replace the existing equipment.
5.0 Exit Meeting The inspector met with licensee representatives (denoted in paragraph 1.0) at the conclusion of the inspection on June 12, 1987 at the nuclear generating station.
The inspector summarized the scope of the inspection and the inspection findings.
At no time during this inspection was written material provided to the licensee by the inspector.
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ATTACHMENT NO. 1 Documents Reviewed by the Inspector
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Generic criteria and commitment documents listed in Section 3.0.
GPU Oyster. Creek Installation Specifications and Task Assignments
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438 Series for Additional In-Containment Instrumentation as Required by NUREG-0578.
GPU Oyster Creek Instrumentation Surveillance and Calibration
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Procedures 604.3.017, 018, 019, and 020.
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GPU Oyster Creek Station Operating Procedure 312, Rev 39 for Reactor Containment Integrity and Atmosphere Control.
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GPU Oyster Creek Records Center Files for NUREG-0737 Modifications, Licensee Record Number B/A 402052-Includes the following: Job Orders, Procedures, NDE Requests and Reports, NCR's and QDRS, FCN's, As-Installed Drawings.
GPU Oyster Creek Station Operating Procedure 105.3, Rev. 3
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Maintenance of Environmental Qualified Equipment.
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GPU Oyster Creek Master List ML-0C-666, Rev. 1-Electrical Equipment Environmental Qualification.
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Burns and Roe Inc. Oyster Creek Station Construction / Installation Drawings for the Additional Instrumentation. Drawings include the following M0151, M0153, M0187, E0220, E0224, E0225, E0221, E0222, E0112, E0110, E0111, E0223, E0219, E0218, E0217, E0216.
GPU Oyster Creek Startup and Test-Functional Test Procedures and
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Turnover Tag Acceptance Approvals, including the following STP 230/2, 300/0.4, 330/1, 400/0, 430/1, 430/3.
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GPU Oyster Creek Station Operating Procedure 12, Rev. 6-Plant Modification Control.
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NRC Safety Evaluation for NUREG-0737 dated September 29, 1983 and June 19, 1985.
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NRC Integrated Plant Safety Assessment SEP Oyster Creek dated January 1983.
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