IR 05000219/1981019
| ML20039B687 | |
| Person / Time | |
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| Site: | Oyster Creek |
| Issue date: | 11/27/1981 |
| From: | Briggs L, Greenman E, John Thomas NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20039B686 | List: |
| References | |
| TASK-3.D.3.3, TASK-TM 50-219-81-19, NUDOCS 8112230418 | |
| Download: ML20039B687 (12) | |
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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Region I Report No. 50-219/81-19 Docket No.
50-219 License No. DPR-16 Priority Category C
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Licensee:
Jersey Central Power and Light Company Padison Avenue at Punch Bowl Road Morristown, New Jersey Oyster Creek Nuclear Generating Station Facility Name:
Inspection at:
Forked River, New Jersey Inspection conducted: October 6 - November 2, 1981 It!)v 8(
Inspectors:
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' date signed U. Thoma, esid nt Reactor Inspector cbs%
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L. E. Brig'gs, Reactor Inspectfr
'da te' si gned date signed ~
Approved by:
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G'. Gree'nman', Chief, Reactor dat6 segned Projects Section 2A Inspection Summary: Inspection on October 6 - November 2,1981 (Report No. 50-219/81-19)
Areas Inspected: Routine inspection by the resident inspector (96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />) and one region based inspector (15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />) of: licensee action on previous inspection findings, review of plant operations, log and record review, surveillance testing,
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onsite_ event followup, review of TMI Task Action Plan items, in-office LER review, l
and review of periodic and special reports.
Results: Violations: None cited.
DCS IDENTIFICATION NUMBERS
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50-219/810825 50-219/810924 50-219-810920 50-219-811005
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Region I Form 12 PDR ADOCK 05000 G
(Rev. April 77).
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DETAILS 1.
Persons Contacted J. Carroll, Director, Oyster Creek Operations K. Fickeissen, Manager, Plant Engineering M. Laggart, Supervisor, Licensing A. Rone, Engineering Manager W. Stewart, Plant Operations Manager J. Sullivan, Manager, Operations D. Turner, Radiological Controls Manager
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The inspector also interviewed other licensee personnel during the course of the inspection including management, clerical, maintenance and operations personnel.
2.
Licensee Action on Previous Inspection Findinos (Closed)
Inspector Follow Item (219/79-23-06) Check for air distribution test procedure: The subject of this item was the inability to perform the Standby Gas Treatment System (SGTS) HEPA filter air distribution tests required by Technical Specification 4.5. K.1.a ( 3). A safety evaluation by the NRC:NRR in support of Amendment 52 to Provisional Operating License DPR-16 stated that the design of the SGTS uses a small filter unit with only one HEPA filter which should ensure uniform flow distribution at the design flow rates. The air flow distribution tests are not warranted for these small units.
Based on this evaluation, Amendment 52 issued on February 11, 1981, deleted the HEPA filter air flow distribution test requirement from Technical Specifications.
(Closed) Item of Noncompliance (219/79-24-02) Non-fire retardant wood crates on reactor building elevation 119: The licensee has established acceptable administrative controls on the use of wood in safety related areas.
Procedure 120 revision 10, dated April 24, 1981, " Fire Hazards" states, "All wood used in safety related areas will be fire retardant. A continuous fire watch will be provided should it become necessary to bring untreated wood, such as equipment l
containers, or flammable material into a safety related area." The inspector determined by frequent observations cf safety related areas that this administrative control has been adequately implemented.
(0 pen) Unresolved Item (219/81-01-03) Evaluate and correct out of specification instrument trip points resulting from instrument drift: The inspector reviewed the results of an analysis performed on four sample ITT Barton differential pressure switches documented in a
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report entitled, "Barton Snap Acting Switches - Analysis of Set Point Drift-Problem", dated May 19, 1981. The analysis consisted of a calibration and error study of two new switches and two
. switches which were removed from field installation. The report concluded that the subject switch appears to be a reliable instrument with a long term repeatibility of 2.5 to 4.5 percent of full scale. Substitution of.better ranged instruments would reduce the problem. The report stated that, in general, the existing equipment was not selected with. ideal ranges. This item will.
remain open pending review of further corrective action.
3.
Review of Plant Operations a.
Periodic inspection tours of ~ selected plant areas were conducted-
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to verify compliance with Technical Specifications (TS) and the licensee's administrative procedures in the areas of housekeeping -
and cleanliness, fire protection, radiation control, physical security, and operational control. Areas' toured included the following:
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Control Room Turbine Building
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Augmented Off-Gas Buildino
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New Rad-Waste Building Cooling Water Intake and Dilution Plant Structure
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Monitoring Change Areas
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4160 Volt Switchgear,' 460 Volt Switchgear, and Cable Spreading
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Rooms
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Diesel Generator Building
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Battery Rooms Maintenance Work Areas
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Yard Areas b.
The following observations were made:
(1)
Through daily observation of Control Room monitoring instrumentation and ' annunciators, log review, and direct
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observation of selected equipment, the inspector verified that systems were in conformance with Technical Specification (TS) Limiting Conditions for Operation (LCO). Applicable portions of the follcwing LC0's which could be verified by control room observation were checked daily:
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TS 3.1.B APRM System TS 3.1.C LPRM System
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TS 3.2.C Standby Liquid Control System
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TS 3.3.A Pressure Temperature Relations
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TS 3.3.D Reactor Coolant System Leakage
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TS 3.3.F Recirculation Loop Operability
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TS 3.4.A Core Spray System
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TS 3.4.B Automatic Depressurization System
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TS 3.4.C Containment Spray and Emergency Service Water System
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TS 3.4.D Control Rod Drive Hydraulic System
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TS 3.5.A Primary Containment TS 3.5.B Secondary Containment
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TS 3.7.A Auxiliary Electric Power TS 3.7.C Standby Diesel Generators
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TS 3.8.A Isolation Condensers
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TS 3.13.A Relief and Safety Valve Position
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Indicators Instrumentation used to verify the above was examined to insure that displayed parameters were within normal and expected limits and that proper correlation existed between redundant instrument channels.
Instrumentation used above included valve position and breaker indication lights, visible portions of instrument recorder traces, and annunciator and alarm panels.
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(2)
During daily control room tours, the inspecter verified that the manning requirements of 10 CFR 50.54k and the licensee's administrative procedures were. met. Shift turnovers were periodically observed to confirm that they were conducted in an '
orderly manner and that sufficient information was exchanged
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to insure the continuity of system status. :The inspector verified that the access control requirements.of the physical'
security plan were met.
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Through frequent control room observation, the' inspector -
verified that evolutions in progress were being performed-in accordance with approved procedures. Control room operators and supervisors were periodically questioned on. evolutions in =
progress, instrument parameters, and alarmed annunciators to-determine that they were aware of current plant-status and that the reasons for abnormal indications and alarmed conditions were understood and that required corrective actions were-being taken.
(4)
Local plant instrumentation was selectively examined to verify that instruments necessary to support safe plant operation and fulfill technical specification limiting conditions for operations were in service and that acceptable correlation
between channels. existed. Safety system actuation sensors were.
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examined to insure that activities in the area did not impair system operability.
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(5)
Monitoring and Change Areas were observed to ensure that entrances to the radiation controlled area (RCA) were properly posted, personnel entering the RCA' were wearing proper dosimetry, and that personnel and materials leaving the RCA
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were properly monitored for radioactive contamination.
Monitoring instruments were observed to ensure that proper operaticnal checks and calibrations had been performed.
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(6)
During tours of the facility, the inspector made observations to verify.that control point procedures were followed, that
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contamination areas and airborne radioactivity areas were -
properly posted, that high-radiation areas were properly posted and locked when required, and that personnel complied
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with the' requirements of applicable radiation work permits (RWP).
The following RWP's were reviewed for completeness _ and work performed under the RWP was observed periodically:
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RWP 161981: Removal'and replacement of emergency
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battery lights and hangers.
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RWP 158281:
LRD rebuild room modifications.
' RWP 164881:
HP Support, trash removal, stock step-
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off pad supplies, retrieval of tools.
RWP 165181: Observation and inspection of RWP areas.
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(7)
Valves and components in safety ~related systems were. observed to verify proper system alignment. Accessible major flow path valves in the Core Spray,~ Containment Spray, Control-Rod Drive Hydraulic, Emergency Service Water, and Isolation Condenser systems were examined for proper alignment by direct observation and by observation of remote position
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indicators. All breakers' in the 4160 Volt and' selected breakers in the 460 Volt and 125 Vdc electrical systems were periodically examined for proper alignment. Systems and.
components were examined for evidence of abnornal vibration and. fluid leaks. Selected pipe hangers and seismic restraints were visually examined for indications of mechanical interference or fluid leaks.
(8)
Equipment Control procedures were examined for proper
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implementation by verifying that tags were properly filled out, posted,- and removed as required, that jumpers were properly installed and removed, and that equipment control logs and records were complete.
. The following electrical jumpers were verified to be properly installed:
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Check-off Sheet 81-371 dated October 26, 1981: Jumper 74 in MCC 1B23-C05 to allow operation of batter.y room exhaust fan 1-20 while supply fan 1-20 was secured for naintenance.
Check-off Sheet 81-369 dated October 22, 1981:
Jumpers
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32,42,55,71,75,85 and 87 in control room panel 10XF and TB-lEl-53B to provide temporary control circuits for valves V-7-29 and V-7-31.
Check-off Sheet 81-368 dated October 21, 1981:
Jumper 87
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in panel 9XR-TBA to allow testing of canal water temperature monitoring system. This jumper was verified by the inspector to be properly cleared following its removal upon completion of maintenance on October 21, 1981.
During the conduct of inspection tours, the interiors of cabinets and control panels were examined for the presence of
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uncontrolled jumpers, lifted leads, or tags. Tags found on systems and components were examined to verify. that.the
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component was in the condition specified on the tags and that tags were properly. filled out and authorized.
Equipment control logs were examined to verify that jumpering or tagging of system components did not remove redundant safety systems from service or violate technical specification limiting conditions for operation.
(9)
Plant housekeeping conditions including general cleanliness, control of material to prevent fire hazards, maintenance of fire -barriers, and storage and preservation of equipment were
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examined. The inspector examined the placement of temporary hoses and extension cords,' and the locations of scaffolding erected for maintenance or modification jobs to verify that -
safety related equipment operability was not impaired.
On October 20, 1981, the inspector observed two JCP&L employees performing maintenance on a hydraulic snubber on core spray system piping in' the south-west corner of the reactor building 23 foot elevation.. One of the individuals had removed his coat and hung it on the manual operating lever arm on one of the reactor building to suppression.
chamber vacuum breakers. The inspector promptly notified thc Group Shift Supervisor and station management and the coat was removed from the valve operating lever. This act did not impair the operability of the vacuum breaker.
However, the inspector expressed concern that similar actions in other areas of the plant could have impaired the operability of safety related equipment. Further review of this event showed that the individuals were aware of the implications of using safety related components as rigging pc.nts, staging platforms, or. attachment points for hoses, extension cords, drop lights, etc. They were, however, unaware of the safety function of the vacuun breaker valves.
.The facility management critiqued this event with the individuals involved and with the appropriate supervisors.
The inspector had no further questions on this item.
(10)
During daily entry and egress from the protected area, the inspector verified that access controls were in accordance
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with the security plan and that security posts were properly manned. During facility tours, the inspector verified that protected area gates were locked or guarded and that isolation zones were free of obstructions. The inspector examined vital area access points to verify that they were properly.
locked or guarded and that access control was in accordance with the security plan.
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Acceptance criteria for the above areas include the current revisions of the following:
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Technical Specifications
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Procedure 106, Conduct of Operations
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Procedure 108, Equipment Control
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Procedure 115, Standing Order Control Procedure 119, Housekeeping
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Procedure 120, Fire Hazards Procedure 122, Security Guidelines for Plant Personnel
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Procedure 903.2, Personnel Monitoring s
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Procedure 903.6,'. Personnel Regulations Procedure 915.1, Restriction of Access into Radiation
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Control Areas
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Procedure 915.4, Contamination Control
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Procedure 915.6, Radiation Work Permit
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Oyster Creek Physical Security Plan
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_ I,nspector judgment 4. ' Shift' Logs and Operating Records a '. s The inspector reviewed tne current revisions of the following plant
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procedu'res to determine the licensee established requirements in this area in preparation for review of selected logs and records:
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Procedure 106, Conduct of Operations;
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Procedure 108, Equipment Control; and,
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Procedure 115, Standing Order Control.
The inspector had no questions in this area.
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Shift logs and operating records were reviewed to verify that:
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Control Room logs were filled out and signed;
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Equipment logs were filled out and signed;
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Log entries involving abnorral conditions provided sufficient detail to communicate equipment status;
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Shift turnover sheets were filled out, signed, and reviewed;
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Operating orders did not conflict with Technical Specification
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requirements; and,
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Logs and records were maintained in accordance with the procedures in a. above, c.
The review included the following plant shift logs and operating records as indicated, and discussions with licensee personnel.
Reviews were conducted on an intermittent selective basis:
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Control Room Log, all entries;
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Group Shift Supervisors Log, all entries; Technical Specification Log;
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Reactor Auxiliary Log;
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Control Room Turnover Check List;
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Reactor Building Tour Sheets;
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Turbine Building Tour Sheets;
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Equipment Tagging Log; Lifted Lead and Jumper Log;
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Defeated Alarm Log;
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Standing Orders;
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Operational _ Memos and Directives No unacceptable conditions were identified.
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Surveillance Testina During control room tours selected completed surveillance tests were reviewed to verify that the tests were completed as scheduled, test results were reviewed by supervisory staff and forwarded for canagement review, and '; hat appropriate corrective actions were initiated as required for identified deficiencies.. Portions of selected ongoing surveillance activities were observed to verify that approved procedures were used, the work was performed by qualified personnel, that test instrumentation was calibrated, and that redundant systems 'or components were available for service if required. Activities reviewed included the following:
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Procedure 617.4.006, revision-2 January 8,1981, " Scram Discharge Volume Vent Piping Blockage Test", completed October 15, 1981.
Procedure 604.4.002, revision 0, March 15,1977, " Reactor Building
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to Suppression Chamber Self Actuating Vacuum Breaker Operability Test", completed October 23, 1981.
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Procedure 619.3.006, revision 8, September 23, 1981, " Reactor Triple Low Water Level Test and Calibration", ccmpleted
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October 23, 1981.
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Procedure 602.4.006, revision 1, September 10,1979, " Main Steam Isolation Valve 5 Percent Closure Test", completed October 27, 1981.
No unacceptable conditions were noted.
6.
Follow-up of On-Site Events a.
On October 19, 1981 at about 10:30 p.m., an automatic scram occurred
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from 64_ percent reactor power due to inadvertent closure of one Main Steam Isolation Valve (MSIV). ' A control room operator attempted to i
perform surveillance procedure 602.4.006, " Main Steam Isolation Valve
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5 Percent Closure Test". The procedure requires that the operator-press the MSIV " Test" button for the selected valve. This bleeds off the operating air pressure to the valve and allows the valve to move in the "close" direction. Releasing the button when the valve
reaches five percent closed (indicated by extinguishing of the valve open light) reenergizes the valve's pilot solenoid and reopens 'the valve. The control room operator turned the MSIV manual control switch
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to the "close" position rather than depressing the " test" button.
This caused full closure of the selected MSIV. The resulting primary
pressure increase caused a reactor high pressure scram.
This event was critiqued by the station management and reviewed with all operations personnel. The cause of this event was operator error due to inattention to the details of the evolution in progress.
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At about 1:50 p.m. on October 21, 1981, a 130 foot length of three inch electrical conduit attached to the outside of the north wall of the reactor building fell when the anchor bolts failed. The conduit carried wiring for instrument and control curcuits to the augmented off-gas and radwaste buildings. The breaking of these wires caused the off-gas holdup volume isolation valve, V-7-31, to fail shut.. Since the main condenser noncondensible gases are discharged through V-7-31, an immediate reactor power reduction and subsequent shutdown were required to prevent a loss of condenser vacuum and possible automatic scram. The failed anchors were expandable lead anchor bolts that were installed during off gas system modifications in 1976. The licensee conducted a review of all plant electrical drawings and a visual inspection of similar modifications to confirm that no other anchors of this type are used in the plant. Temporary repairs to the control circuits were completed and operations resumed on October 22, 1981.
The inspector had no further questions on the above item.
7.
In-Office Review of Licensee Event Reports (LER's)
The inspector reviewed LER's received in the NRC:R1 and Resident Office to verify that details of the event were clearly reported including the accuracy of the description of cause and adequacy of corrective action.
The inspector also determined whether further information was required from the licensee, whether generic implications were involved, and whether the event warranted on-site followup. The following LER's were reviewed:
~LER'
EVENT 81 -44/3L Standby Gas Treatment System exhaust fan 1-8 was removed from service for corrective maintenance 81-45/3L Isolation Condenser valve V-14-32 failed during performance of routine surveillance test 81-46/3L Three hydraulic snubbers in the shutdown cooling system failed during functional testinc 81-47/3L Diesel Generator number 1 failed to achieve peak load during surveillance testing.
The inspector had no further questions on the above items.
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8.
Review of TMI Task Action Plan Items a.
Item III.D.3.3.1 Provide means to determine presence of radiciodine.
The inspector determined that equipment and procedures are in place to monitor iodine levels. The licensee uses a silver zeolite sample media. Sample cartridges can be analyzed in the field using an HP-210
G-M probe and a portable scaler or in the on site counting lab on a Ge-Li counter.
The following. procedures were reviewed:
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Procedure 905.54, revision 9. June 30,1980, " Emergency
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Monitoring Kit - Counting of Particulate and Charcoal Filters -
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Air Sample Calculation".
Procedure 905.65,-revision 0, January 29,1980, " Post Accident
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In-Plant Radiation Measurement".
The inspector determined that training was provided for health
physics technicians on dose assessment, personnel decontaminatic,i, and onsite and offsite radiation surveys.
No unacceptable conditions were identified.
9.
Review of Periodic and Scecial Reports.
The fo' lowing periodic and special reports submitted by the licensee
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were reviewed by the inspector. 'The inspector determined that information was reported to the NRC as required.. planned corrective action appeared adequate to resolve identified problems and that reported information was valid.
September 1981
_ Monthly Operating Data Report ---
10.
Exit Interview At periodic intervals during the course of this inspection, meetings were held'with senior facility management to discuss inspection scope and findings. A summary of the inspection findings was also provided to the licensee at the conclusion of the inspection on N)vember 3, 1981.
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