IR 05000219/1980035
| ML20003D110 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 01/28/1981 |
| From: | Briggs L, Keimig R, John Thomas NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20003D109 | List: |
| References | |
| 50-219-80-35, NUDOCS 8103190308 | |
| Download: ML20003D110 (8) | |
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O U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Region I Report No. 50-219/80-35 Docket No. 50-219 License No. DPR-16 Priority Category C
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Licensee:
Jersey Central Power and Light Company Madison Avenue at Punch Bowl Road Morristown, New Jersey 07960 Facility Name:
Oyster Creek Nuclear Generating Station Inspection at:
Forked River, New Jersey Inspection conducted: Dece er 1 - December 31, 1980
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Inspectors:
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J. A. Thomas, Resident Reac' tor Inspector date signed YY hm.O
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1. E. Brigt%, Reactor Inspector date signed
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date signed
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Approved by:
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R. R. Keimig, Chief, Reactor Projects dats signed Section No.1, RO&NS Branch i
Inspection Summary:
-Inspection On December 1 - December 31, 1980 (Report No. 50-219/80-35)
Areas Inspected:. Routine inspection by the resider.t inspectors -(39 hours4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br />) of:
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Licensee action on previous inspection findings; tours of the facility; log and record review; follow-up of onsite events; and follow-up of IE Bulletins and Circulars.
Resul ts : No items of noncompliance were identified.
I Region -I Form 12
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(Rev. April 77)
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DETAILS 1.
Persons Contacted J. Carroll, Director, Oyster Creek Operations K. Fickeissen, Manager, Plant Engineering J. Maloney, Manager, Plant Maintenance A. Rone, Engineering Manager W. Stewart, Plant Operations Manager J. Sullivan, Manager, Operations D. Turner, Radiological Controls Manager The inspectors also interviewed other licensee personnel during the course of the inspection including management, clerical, maintenance, and operatiens personnel.
2.
Licenses Action on Previous Inspection Findings (Closed) Unresolved Item (219/79-10-05):
Procedural changes regarding recirculation loop isolation, operator guidance, and operator training.
Procedure 301, Nuclear Steam Supply System, Revision 16, dated July 1, 1980, states in part. "During all modes of operation with irradiated fuel in the vessel except when the reactor head is removed and reactor cavity flooded above the main steam nozzles, at least two recirculation loop suction valves and their associated main discharge valves will be in the full open position." Technical Specifications Section 2.1, Safety Limit-Fuel Cladding Integrity, paragraph F, has been revised to contain the same statement.
Procedure 307, Isolation Condenser System, has been revised to contain a similar precaution on.the operation of the recirculation loop isolation valves.
The inspector interviewed control room operators and verified that adequate training has been conducted to insure _ operator awareness of the effects of recirculation loop isolation on reactor water level measurement.
The inspector had no further questions on this item.
(Closed) Unresolved Item (219/79-10-07):
Conflicting level alarm indica-tions and operator training on level instrumentation. The implementation of the corrective actions discussed in the previous item above will insure adequate level communication between the annulus and core regions of the reactor vessel to preclude disparity between level alarms sensing in the core region and level indications sensing in the annulus region.
The inspector had no further questions on tnis item.
(Closed) Unresolved Item (219/79-10-08): Unavailability of wide range Barton indication in control room.
A wide range level instrument has been installed in the control room which uses the upper GEMAC instrument tap (180 inches above the top of active fuel) and the standby liquid control sparger (below lower core support plate) as its sensing points.
This instrument is operable only when all recirculation pumps are secure.
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However it will provide adequate core region water level indication (-20 inches below top of active fuel to 180 inches above tcp of active fuel)
under accident conditions.
The inspector had no further questions on this item.
(Closed) Unresolved Item (219/79-10-09):
Scaling methods used for level indication, lack of consistency in instrument zero. This has been corrected by the addition of a dual scale to all reactor vessel water level indications in tne control room.
The added scale gives reactor water level in inches above the top of the active fuel, allowing easy correlation between instru-ments. The inspector had no further questions on this item.
(Closed) Follow Item (219/79-14-04):
Completion of replacement of flammable contact arm retainers in relays in control room safety panels 1F/2F, 3F, and 5F/6F.
Replacement of flammable contact arm retainers in G.E. CR 120A relays was a requirement of Inspection and Enforcement Bulletin (IEB) 78-01.
Corrective action has been completed and is documented in detail in I&E Inspection Report No. 50-219/80-25.
The inspector had no further questions on this item.
(Closed) Unresolved Item (219/80-25-01):
Revise procedure 201.2 to require test of turbine bypass valves prior to reaching full operating pressure.
The inspector reviewed Procedure 201.2, Plant Heatup to Hot Standby, Revision 8, dated July 24, 1980. This revised procedure requires verification of turbine bypass valve operability alter main condenser vacuum is established and vacuum trip number 2 is reset.
The inspector had no further questions on this item.
3.
Plant Tour a.
During the course of the inspection, the inspector made observations and conducted multiple tours of:
Control Room;
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Reactor Building:
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Turbine Building;
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Augmented Off-Gas Building;
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Cooling Water Intake Structure;
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New Rad-Waste Building;
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Old Rad-Waste Building;
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Monitoring Change Areas;
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Maintenance work areas; and
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Yard Areas.
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b.
The following determinations were made:
Monitoring instrumentation. The inspector verified that selected
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instruments were functional, demonstrated parameters within Technical Specification limits, and demonstrated proper correlation between channels.
Radiation controls. The inspector made observations to verify
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that control point procecures and posting requirements were being followed.
Systems and equipment in all areas toured were cbserved for the
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existence of fluid leaks and abnormal piping vibrations.
Plant housekeeping conditions including general cleanliness
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conditions and control of materials tc prevent fire hazards were observed in the areas listed. Maintenance of fire barriers in these areas was also observed.
Lit control board annunciators were reviewed with c ntrol room
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operators to verify that the reasons for the alarmed conditions were understood and that corrective action, if required, was being taken.
Selected valves in safety related systems were checked to verify
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proper system alignment.
By frequent observations during the inspection,-the inspector
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verified that control room manning requirements of 10 CFR 50.54(k)
and the Technical Specifications were being met.
In addition, the inspector observed shift turnovers to verify that continuity of system status was maintained.
No items of noncompliance were identified.
c.
The following acceptance criteria were used for the above items:
Technical Specificaticas;
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10 CFR 50.54(k); and,
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Inspector judgment.
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Shift Logs and Operating Records a.
The inspector reviewed the following plant procedures to determine the licensee established requirements in this area in preparation for review of selected logs and records:
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.
b.
Shift logs and operating records were reviewed to verify that:
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 aonormal conditions provided sufficient
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detail to communicate equipment status; 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, Logs and records.were maintained in accordance with the procedures
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in a. above.
c.
The review included the following plant shift logs and operating records as indicated and discussions with licensee personnel.
Caviews were conducted on an intermittent selective basis:
Control Room Log, December 1 through December 31;
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Control Room Alarm Sheets;
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Control Rod Status Sheets;
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. Technical Specification Log;
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Reactorf Auxiliary Log;
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Reactor Log;
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Control Room Turnover Check List;
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Standing Orders, all active; and,
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Operational Memos and Directives, all active.
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No items of nonccmpliance were identified.
5.
Follow-up of On-Site Events at At 3:27 p.m. on December 5,1980, the licensee notified the NRC:RI office that unusual environmental conditions, during the week of November 30, 1980, had led to extremely low tide conditions and critically low intake canal water levels. The licensee stated that at 5:35 p.m. on December 4, 1980, both screen wash pumps had lost suction and were tripped. The subsequent debris buildup caused a further reduction of 1"take water level.
In anticipation of a plant shutdown (scram) due to a loss of cooling water (Low Condenser Vacuum),
both operating dilation pumps were tripped (5:58 p.m.) and a controlled plant shutdown initiated at 6:00 p.m. on December 4,1980. The tripping of the dilution pumps decreased intake canal flow and allowed the water level to recover sufficiently to resume plant operations. The shutdown was terminated after a power reduction of 25 MWe. This event, reported in Nonroutine Environmantal Report No. 50-219/80-13, was reviewed on site by the inspector on December 8, 1980. The licensee has purchased new screen wash pumps which are scheduled for installation at a later date. The new pumps will ex:end deeper into the intake water and should preclude further events of this type.
In addition, licensee survaillance of the discharge canal following this event did not disclose any adverse effects on marine life. The inspector had no further questions concerning this event.
b.
On December 13, 1980 at approximately 1:30 p.m., a water leak was discovered in the area of the demineralized water transfer pumps. The i
l source of the leak was determined to be a rupture in the two inch re-
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circulation line on the number one demineralized water transfer pump.
Approximately 7500 gallons of demineralized water leaked through a broken weld in the aluminum piping before the pump was secured and valved out of service. Since the pump is located in a contamination l
controlled area, samples of the spilled water and the wetted soil were collected for analysis. Subsequent isotopic analysis showed that no release or spread of radioactive contamination had occurred as a result
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of the incident.
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The cause of the pipe rupture is believed to be fatigue in the welds of the aluminum pipe.
Tne inspector expressed concern for the possibility of a similar occurrence in the condensate transfer system piping. The condensate transfer pumps are located adjacent to the demineralized water transfer pumps and run continuously during plant operation to supply water to various locations in plant support system!..
The piping at the condensate transfer pumps is similar in arrangement and construction to the piping at the demineralized water trarisfer pumps.
A similar rupture in this piping could result in an uncontrolled release of radioactively contaminated water to the environment.
The licensee agreed to review this concern at a meeting of the Operations Experience Assessment Committee (OEAC).
The licensee's resolution of this concern will be reviewed in subsequent inspections.
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No items of noncompliance were identified.
6.
IE Bulletins and Circulars a.
Bulletins Licensee actions concerning the following IE Bulletins were reviewed to verify that: The Bulletin was forwarded to appropriate on-site management; a review for applicability was performed; information discussed in the licensee's reply was accurate; corrective action taken_was as described in the reply; and the reply was within the time period described in the bulletin.
IEB 80-17, Failure of Control Rods to Insert During a Scram at a
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BWR.
The inspector verified that procedures have been implemented to provide clear guidance to the control room operator regarding when he should initiate the standby liquid control system (SLCS)
without obtaining prior supervisory approva!.
Procedure 506.5,
. Scram System Failure, gives this guidance. Procedure 106, Conduct of Operations,-ensures that the SLCS key is readily available to the licensed operator in the control room.
By correspondence dated August 29, 1980, the licensee responded to the requirements of Supplement 3 to this bulletin.
The inspector verified that Procedure 501, Annunciators and Alarms, requires a manual scram in the event of multiple rod drift-in alarms or a marked change in the number of control rods with high temperature alarms. The inspector verified that startup procedures require performance of a functional test, using water, of the SDV high level alarm, rod block, and scram switches ' prior to startup following a reactor scram.
The inspector reviewed the licensee's action to require an immediate scram when control rod drive (CRD)
air _ header pressure drops to within 10 PSI above the opening pressure cf the scram valves.
Initially, the licensee revised
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procedures to require dispatching an operator to the reactor building to monitor CRD air pressure locally upon receipt of a low pressure alarm (75 psi), and to initiate a manual scram when the local pressure gauge indicated 60 psi. This was determined to be unacceptable due to the possible time lag between receipt of the alarm and arrival of the operator at the local pressure gauge. This time lag could present the potential for scram system degradation caused by water leakage through scram valves into the SDV due to reduced air pressure. The licensee subsequently reduced the low CRD air pressure alarm to 60 psig and revised procedure 501, Annunciators and Alarms to require an immediate manual scram upon rece:pt of the alarn.
In addition, standing orders have been issued specifying remedial action to be taken if water is found in the scram discharge volume (SDV)
system.
The inspector had no further questions on the above items. This bulletin will remain open pending completion of additional actions required by IEB 80-17 and its supplements.
b.
Circulars Licensee actions concerning the following IE Circulars were reviewed to verify that the circular was received by licensee management, that a review for applicability was performed, and that action taken or planned is appropriate.
All of the following circulars were detennined to be not applicable to the Oyster Creek facility:
I6C 78-18, UL Fire Test. This circular was issued for information
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only and required no cor*ective action.
IEC 79-01, Administration of Unauthorized Byproduct Material to
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Humans.
' IEC 79-06, Failure to Use Syringe and Bottle Shields in Nuclear
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Medicine.
IEC 79-16, Excessive Radiation Exposures to Members of the General
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Public and a Radiographer.
The -inspector had no further questions on the above items.
7.
Exit Interview-At periodic intervals during the course of this inspection, meetings were held with senior facility management to discuss inspection scope and findings.
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