IR 05000219/1980032
| ML19345F018 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 11/26/1980 |
| From: | Briggs L, Keimig R, John Thomas NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19345F016 | List: |
| References | |
| 50-219-80-32, NUDOCS 8102060528 | |
| Download: ML19345F018 (15) | |
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U.S. NUCLEAR REGULATORY COMMISSION m
0FFICE OF INSPECTION AND ENFORCEMENT
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Region I Report No. 50-219/80-32 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:
October 6 - October 31, 1980 Inspectors: p>W $,Ar
//br/go L. E. Brig 6s, Senior Resident Reactor Inspector date signed A-n /nho J. A. Thomas, Resident Reactor Inspector d' ate signed date signed
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Approved by:
9W;
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R. Keimig, ief, Regor Projects date signed Section No.
, RO&NS Manch Inscection Summary:
Inspection on October 6 - October 31,1980 (Report No. 50-219/80-32)
Areas Inspected:
Routine inspection by the resident inspectors (83 hours9.606481e-4 days <br />0.0231 hours <br />1.372354e-4 weeks <br />3.15815e-5 months <br />) of:
Licensee action on previous inspection findings; tours of the facility; log and record review; followup of on-site events; surveillance testing; implementation of TMI task action plan category "A" requirements; and followup of IE Bulletins and Circulars.
Results: No items of noncompliance were identified.
02060S'4D
Region I Form 12 (Rev. April 77)
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DETAILS
1.
Persons Contacted J. Carroll, Director, Oyster Creek Operations K. Fickeissen, Manager, Plant Engineering I. Finfrock, Jr., Vice President JCP&L, Director, Oyster Creek J. Maloney, Manager, Plant Maintenance 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 operations personnel.
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2.
Licensee Actibn on Previous Insoection Findings a.
Open Items as a Result of Inspection 79-11 and IE Bulletin 79-08 The following open items were reviewed by the inspector to verify that appropriate action had been taken. The ins reviewed applicable procedures, check-off lists (COL's)pector and Process and Instrumentation Drawings (P&ID's) as listed with each item.
In addition, selected system valve identification tags and valve positions were observed to compare with the applicable COL's and P&ID's.
(Closed) Unresolved (79-11-02) Licensee to assign permanent
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unique valve numbers to each valve and update COL's for the Stand-by Liquid Control System (SBLCS).
Procedure No. 304, Stand-by Liquid Control System, Revision 7,
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January 14, 1980; and,
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B&R Drawing No. 2013, Revision ll, May,1980.
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(Closed) Unresolved (79-11-03)
Several instrument valves were.not identified or shown on SBLCS P&ID.
B&R Drawing No. 2013, Revision 11, May,1980.
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(Closed) Unresolved _(79-11-04) SBLCS vent and fill procedural error.
Procedure No. 304, Stand-by Liquid Control System, Revision 7,
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January 14, 1980.
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(Closed) Unresolved (79-11-05) Licensee to assign permanent unique valve numbers to each valve and update COL's for the Core Spray System.
Procedure No. 308, Emergency Core Cooling System, Revision 12,
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June 18,1980; and, GE Drawing No. 885D781, Revision 13, March,1980.
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(Closed) Unresolved (79-11-06) Several instrument valves were not identified or shown on the Core Spray System P&ID.
GE Drawing No. 885D781, Revision 13, March,1980.
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(Closed) Unresolved (79-11-09) Licensee to assign permanent unique valve numbers to each valve and update COL's for the Containment Spray Systam.
Procedure No. 310, Containment Spray System, Revision ll,
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July 7,1980; and, GE Drawing No.148F740, Revision 14, December,1979.
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(Closed) Unresolved (79-11-10) Several instrument valves were not identified or shown on the Containment Spray System P&ID.
GE Drawing No.148F740, Revision 14, December,1979.
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(Closed) Unresolved (79-11-11)
Capped bypass pipe not shown en Stand-by Gas Treatment System (SBGTS) P&ID.
B&R Drawing No. 2011, Sheet 1, Revision 10, May,1980.
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t (Closed) Unresolved (79-11-12) Licensee to revise vanual SBGTS initiation to shut V-28-21 and V-28-22.
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l Procedure No. 330, Stand-by Gas Treatment System, Revision 5,
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July 1,1980.
(Closed) Unresolved (79-11-13) Licensee to revise Procedure No. 330 to include manway and damper alignment.
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-- Procedure No. 330, Stand-by Gas Treatment System, Revision 5, July 1,1980.
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(Closed) Unresolved (79-11-14) Licensee to identify reactor protective system instrument rack valve numbers.
GE Drawing No. 148F712, Revision 8, April, 1980; and,
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Standing Order No. 27.
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(Closed) Unresolved (79-11-15)
Instrument valves for reactor protective system not shown or identified on P&ID.
-- GE Drawing No. 148F712, Revision 8, April, 1980.
(Closed) Unresolved (79-11-16) Licensee to assign' permanent unique valve numbers to each valve and update COL's for the Shutdown Cooling System (SDCS).
Procedure No. 305, Shutdown Cooling System, Revision 8,
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May 29,1980; and,
-- GE Drawing No.148F711, Revision 9, May,1980.
(Closed) Unresolved (79-11-17) Several instrument valves were not identified or shown on SDCS P&ID.
GE Drawing No. 148F711, Revision 9, May, 1980.
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(Closed) Unresolved (79-11-18) Several SDCS valves incorrectly numbered.
GE Drawing No.148F711, Revision 9, May,1980.
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l (Closed) Unresolved (79-11-19) Diesel generator line-up procedure did not require fuel oil valve line-up verification.
Procedure No. 341, Stand-by Diesel Generators, Revision 11,
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June 18, 1980.
(Closed) Unresolved -(79-11-20) Licensee to assign permanent unique valve numbers to each valve and update COL's for the Isolation Condenser System.
Procedure No. 307, Isolation Condenser System, Revision 11,
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l July 2,1980; and, GE Drawing No.148F262, Revision 8, July,1980.
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(Closed) Unresolved (79-11-21) Several instrument valves were not identified or shown on the Isolation Condenser P&ID.
GE Orawing No. 148F262, Revision 8, July, 1980.
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The inspector had no further questions relating to the above items.
b.
Other Inspection Findings (Closed) Inspector Follow Item (77-06-04) Licensee to increase height of curbing around acid and caustic tanks. The inspector verified that curbing around the subject tanks in Old Rad-Waste had been raised; however, the tanks in question have not been used since late 1978 when the New Rad-Waste facility was put into operation.
The inspector also verified that the tanks currently being used have dikes surrounding them that will contain the total volume of the tanks should any leakage occur.
The inspector had no further questions concerning this item.
3.
Plant Tour a.
During the course of the inspection, the inspector made observations and conducted multiple tours of:
-- Control Room; Reactor 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; Monitoring Change 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 instruments were functional and demonstrated parameters within Technical Specification limits.
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-- Radiation controls. The inspector made observations to verify that control point procedures and posting requirements were being followed.
Fluid leaks. No fluid leaks of significance were observed
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which had not been identified by plant personnel and for which corrective action had not been initiated.
Plant housekeeping conditions. Observations of plant house-
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keeping relative to fire hazards identified no notable conditions.
Piping vibration. No excessive piping vibration was noted
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during the plant tours.
Control room annunciators. Selected lit annunciators were
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discussed with control room operators to verify that the reasons for them were understood.
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 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.
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c.
The following acceptance criteria were used for the above items:
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-- Technical Specifications; 10 CFR 50.54(k); and,
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-- Inspector judgment.
d.
The following specific observations were made by the inspector:
The inspector expressed concern over the number of contamination
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controlled areas and high radiation areas throughout the plant and the fact that this may hinder the equipment operators in the performance of routine tours. The licensee has assigned crews to perform general decontamination of the plant to reduce the number of contamination controlled areas and the number of areas requiring the use of respiratory protection. Progress in this area will be reviewed by the resident inspector on a continuing basis.
The inspector had no further questions on the above.
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4.
Shift Logs and' 0perating Records
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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, Revision 8, July 3,1980;
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Procedure 108, Equipment Control, Revision 22, July 24,1980; and,
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Procedure 115, Standing Order Control, Revision 7,
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September 15, 1979.
The inspector had no questions in this area.
b.
Shift logs and operating records were reviewed to verify that:
Control Room logs were filled cut and signed;
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Equipment logs were filled out and signed;
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Log entries involving abnormal conditions provided sufficient
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detail to communicate equipment status;
-- Shift turnover sheets were filled out, signed, and reviewed; 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
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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:
-- Control Room Log, Octcber 6 through October 31; l
Control Room Alarm Sheets;
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Control Rod Status Sheets;
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Technical Specification Log;
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-- Reactor Auxiliary Log; 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 noncompliance were identified.
5.
Follow-up of On-Site Events The inspector conducted a detailed review of documentation and held discussions with plant personnel to fully evaluate the circumstances leading to the September 11, 1980 report from the NRC's Office of Congressional Affairs (OCA) to Region I that radioactive waste had reportedly been dumped by the licensee at the Southern Ocean County Landfill during the first week of August 1980.
Results of a prelim-inary inspection indicated 'that no radioactive material had left the licensee controlled area. This information was communicated to the OCA and the alleger on September 15, 1980. Independently, the Ocean County Radiological Officer upon hearing of the allegation, conducted a radiological survey of the Tandfill area and confirmed that no radioactive material was present.
The following records were reviewed on site:
H.P. Form No. 915.11, Unusual Incident Report, UIR No. 575-80, dated
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August 5, 1980; Ra'diation Protection Survey Record (RPSR August 6,1980, survey of dumpster full o)f trash;No. 9263-80, dated
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RPSR No. 9253-80, dated August 5,1980, survey of trash truck;
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RPSR No. 9285-80, dated August 7,1980, survey of dumpster contents;
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- -- RPSR No. 9371-80, dated August 12, 1980, survey of contents of scrap metal dumpster; RPSR No. 9384-80, dated August 12, 1980, survey of old warehouse;
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- -- RPSR No. 9412-80, dated August 13, 1980, survey of scrap metal dumpster by pretreatment building; RPSR No. 9437-80, dated August 14, 1980, survey of dumpster by
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pretreatment building;
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RPSR No. 9427-80, dated August 14, 1980, survey of scrap metal
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dumpster; RPSR No. 9455-80, dated August 15, 1980, survey of dumpster; and
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RPSR No. 9459-80, dated August 15, 1980, survey of outside of LSA
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The inspector determined that three dumpsters, two small trash dumpsters and one large scra) metal dumpster, located by the pretreatment building, were involved in tae incident. The inspector determined that during a routine tour by a member of the licensee's Radiological Assessment Group on August 5,1980, three radioactively contaminated (low level) items were identified and removed from the large scrap metal dumpster. The three items and their contamination levels were:
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One four inch black iron pipe elbow with loose surface contamination
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of 8000 dpm beta-gamma, and fixed contamination reading one (1)
millirem per hour;
-- A piece of plastic bag with loose surface contamination of 2000 to 3000 dpm beta-gamma, and no detectable fixed contamination; and,
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A 25 foot section of extension cord with. loose surface contamination
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of 2000 to 3000 dpm beta-garra, and no detectable fixed contamination.
Subsequent to the removal of the contaminated items, but prior to the licensee's posting and survey of the three dumpsters, the licensee's contracted trash removal service arrived on site and commenced loading
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one of the srall trash dumpsters. The licensee observed the loading of trash and stopped the operation. The licensee directed the driver of the trash truck to an on-site radioactive materials area (RMA) where the entire contents of the truck were unloaded. The trash truck was then surveyed for both loose and fixed contamination. No contamination was detected, and the truck was released from the site in an empty condition. All dumpsters and the contents of the trash truck were subsequently surveyed by the licensee.
Several contaminated items were identified in the scrap metal dumpster during the licensee's surveys (identified with an asterisk in the above listing of reviewed documents).
The additional items identified were:
Three sections of pipe with loose internal contamination of up to.
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3000 dpm beta-gamma, and fixed contamination of 300 cpm above l
background; One section of angle iron with loose surface contamination of
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70,000 dpm beta-gamma, and no fixed contamination; and, A steel plate with fixed contamination of 150 cpm above background l
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and no loose contamination.
No contaminated material was identified in the two small trash-dumpsters or in the contents of the trash truck. The removal of the large scrap metal dumpster contents, where all the contaminated items were located, is not part of the licensee's contract with.the trash removal service
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trash removal service. The inspector was satisfied that appropriate
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corrective action had been taken, for this event, but expressed concern about the licensee's system of controls that'would allow radioactive l
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material to leave the RMA without proper survey and tagging. The licensee stated that the system was controlled by procedure with friskers available at all RMA exits; however, all RMA exits are not continuously manned by health physics technicians.
In addition all personnel, prior to working on site, are instructed during an indoctrination course not to remove material from RMA's without proper surveys by H.P. Technicians and to follow plant procedures. The licensee again stressed the importance of the above during the Monthly Safety Meeting held in August. The meeting involved station and contractor union personnel through the first line supervisor level. The licensee, as a permanent corrective action, has committed to reduce the number of RMA exits to two (2).
Each exit will be manned whenever the exit is in use and locked when not in use. This change, which will require multiple RMA exit lock additions, is scheduled to occur coincident with the manning of the new maintenance building, on or about December 31, 1980.
This item is unresolved pending the licensee's satisfactory implementation of RMA exit controls and inspection by the resident inspectors (219/80-32-01).
In addition, the licensee committed to perform daily surveys of dumpster contents as an interim measure until strict RMA exit control is established.
Survey results will be reviewed by the resident inspectors as part of the routine inspection program.
6.
Surveillance Testing a.
Selected surveillance testing was observed to verify the following:
Surveillance procedures conformed to techn! cal specification
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requirements and had received proper licensee review and approval; Testing was conducted in accordance with the approved procedures;
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Removal of systems from service and subsequent restoration
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complied with technical specification requirements;
-- Test instrumentation in use was calibrated;
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Test data was accurate, complete, and met technical specification
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requirements; Test documentation was reviewed and discrepancies rectified; and
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-- Testing was performed by qualified personnel and met the surveillance schedule.
b.
The following surveillance testing was observed:
Containment Spray System Automatic Actuation Test, Procedure
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607.3.002, Revision 10; LPRM Test and Calibration, Procedure 620.3.001, Revision 2; and,
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Ultrasonic Test of Scram Discharge Volume Header Piping, Procedure
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617.5.007, Revision 1.
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No items of noncompliance were identified.
7.
Follow-up of TMI Task Action Plan Category "A" Reauirements Implementation of the licensee's commitments on the following TMI Action Plan Requirements was inspected on site. The items are identified by NUREG 0578 item number and NUREG 0660 Task Action Plan (TAP) number.
NUREG 0578 number 2.1.2.b/ TAP I.A.l.1, Shift Technical Advisor
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(STA). The STA program was instituted following the 1980 refueling outage and a training program is presently in progress. The training program will be completed and all STA's will take both oral and written examinations by December 31, 1980. There are presently no procedures that explicitly delineate the duties and responsibilities of the STA. These procedures are being developed and will be completed by December 31, 1980.
NUREG 0578 number 2.2.1.a/ TAP I. A.l.2, Shift Supervisor's
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responsibilities specified, delegate non-safety related duties.
Shift Supervisor's responsibilities are defined in Procedures 106,
" Conduct of Operations" and 106.2, " Conduct of Operations During Emergency Conditions". These procedures clearly specify that the Shift Supervisor's primary responsibility is for the safety of ggogh shift supervisor also has been relieved of some adminis-es.
NUREG0578n'mber2.2.1.c/TANI.C.2,Shiftandreliefturnover u
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procedures.
Procedure 106, " Conduct of Operations", provides formal check lists to be used to accomplish a shift or relief turn-over. The check lists require verification that critical plant parameters are within limits, list the availability and aligrrent of safety systems, and require annotation of all abnormal plant conditions including systems or components in a degraded mode of operation permitted by technical specifications and equipment on accelerated surveillance frequency.
NUREG 0578 number 2.2.1.a/ TAP I.C.3, Clearly define supervisor
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t and operator responsibilities.
Procedures 106, " Conduct of Operations" and 106.2, " Conduct of Operations During Emergency Conditions", define the duties and responsibilities of the Shift
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l Supervisors and Control Room Operators. The procedures delineate a clear line of command authority and specify when the supervisor or operator has the responsibility to shutdown the reactor.
-- NUREG 0578 number 2.2.2.a/ TAP I.C.4, Establish authority and limit access to the control room.
Procedure 106, " Conduct of Operations" gives the Shift Supervisor the responsibility and authority to clear the control room of unnecessary personnel when their presence inter-feres with normal control room operations.
Procedure 106.2, " Conduct of Operations During Emergency Operations", restricts access to the control room during emergencies.
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NUREG 0578 number 2.1.8.a/ TAP II.B.3, Post accident sampling. The
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licensee has implemented procedures for the collection of primary coolant and containment atmosphere samples during accident conditions when the present sample stations are accessible and.has purchased shielded receptacles for handling highly radioactive samples. A design review of the sampling system has been completed and the necessary equipment has been ordered to allow remote sampling during an accident. The equipment is scheduled for delivery in December 1980 and will be installed during the 1981 outage. This action was reviewed and considered acceptable by NRR as specified in a NRR letter to the licensee dated May 8,1980.
NUREG 0578 number 2.1.3.a/ TAP II.D.3, Valve position indication.
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Installation of acoustic monitors on all safety and relief valves was completed during the 1980 refueling outage. The monitors give positive indication of flow through the valves and provide indication and alarm functions in the control room.
Seismic and environmental qualification of the system has not yet been completed. Operation of the system is governed by Procedure 413, " Operation of the Safety Valve / Relief Valve - Valve Monitoring System".
NUREG 0578 number 2.1.7.a and 2.1.7.b/ TAP II.E.1.2, Auxiliary Feed
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' system initiation and flow indication. The Oyster Creek plant, being a BWR, does not have an auxiliary feed system; thus, this item is not applicable.
x NUREG 0578 number 2.1.1/ TAP II.E.3.1, Emergency power for pressurizer
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heaters. The Oyster Creek plant, being a BWR, does not have g
pressurizer heaters; thus, this item is not applicable.
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NUREG 0578 number 2.1.5.c, Recombiner procedures review. The Oyster
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Creek plant's containment atmosphere control system does not use hydrogen recombiners; thus, this item is not applicable.
NUREG 0578 number 2.1.4/ TAP II.E.4.2 Isolation dependability. The
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inspector reviewed plant operating and emergency procedures to verify that resetting of a containment isolation signal is not permitted until all affected valve control switches are placed in the " shut" position. The diverse isolation signals consist of high drywell pressure and low-low reactor water level.
NUREG 0578 number 2.1.3.6/ TAP II.F.2, Inadequate core cooling instru-
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ments. The licensee has implemented emergency procedures for dealing with pipe breaks in various systems in the coolant pressure boundary both inside and outside of the primary containment. These procedures provide guidance on the use of existing instrumentation to recognize conditions that would lead to inadequate core cooling.
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NUREG 0578 number 2.2.2.b and 2.2.2.c/ TAP III.A.1.2, Upgrade
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emergency support facilities. The inspector examined the Technical Support Center (TSC) in the conference room of the auxiliary office building and the Operational Support Center (OSC) in the instrument shop. The facilities provided in the TSC and OSC appear to be adequate for the performance of their intended support functions.
Procedure 106.2, " Conduct of Operations During Emergency Conditions",
and Procedure 106.3, " Technical Support Center", adequately address the manning and activation of the TSC and OSC.
NUREG 0578 number 2.1.6.a/ TAP III.D.l.1, Primary coolant sources
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outside containment. The licensee has established adequate surveill-ance procedures for leakage reduction on the following systems:
reactor water clean-up system, shutdown cooling system, containment spray system, core spray system, standby gas treatment system, coolant sample system, reactor pressure vessel instrumentation system, isolation condenser system, reactor building and drywell equipment drain system, recirculation system instrumentation, reactor building and drywell floor drain system, and control rod drive hydraulic system.
Surveillance on these systems wac completed during the 1980 outage.
NUREG 0578 number 2.1.8.6/ TAP II.F.1, Additional accident monitoring
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instrumentation. The licensee has installed a high range noble gas monitor in the base of the plant ventilation stack.
Construction is in progress on a building to house a permanent system which is due for installation by April, 1981.
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The inspector had no further questions on the above items.
8.
IE Bulletins and Circulars a.
Bulletins Licensee actions concerning the following IE Bulletins were reviewed by the inspector 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 g
was within the time period described in the Bulletin.
IEB 79-01, 79-Ol A, 79-OlB, Environmental Qualification of Class
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IE Equipment. The Oyster Creek plant is in the Systematic Evaluation Program (SEP). Therefore, these bulletins are not applicable.
The inspector had no further questions on this item.
IEB 79-07, Seismic Stress Analysis of Safety Related Piping.
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By correspondence dated April 24, 1979, the licensee responded to this bulletin. The response states that no areas have been found where seismic stress, load, or intermodal responses were summed algebraically. The response was reviewed to verify that no algebraic summation was utilized.
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The inspector had no further questions on this' item.
The following IEB's were reviewed and determined to be not
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applicable to the Oyster Creek plant:
(1) IEB 80-04, Analysis of a PWR Main Steam Line Break With Continued Feedwater Action; (2) IEB 80-05, Vacuum Condition Resulting in Damage to Chemical Volume Control System Holdup Tanks; (3) IEB 80-07, BWR Jet Pump Assembly Failure; (4) IEB 80-12, Decay Heat Removal System Operability; and, (5) IEB 80-18, Maintenance of Adequate Minimum Flow Thru Centrifugal Charging Pumps Following Secondary Side High Energy Line Rupture.
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The inspector had no further questions on the above items, b.
Circulars Licensee actions concerning the following IE Circulars were reviewed to verify that the circular was received by lice ~ ee management, that a review for applicability was performed, and that action taken or planned is appropriate.
All of the following circulars were determined to be not applicable to the Oyster Creek facility:
IEC 80-06; Control and Accountability Systems for Implant
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Therapy Sources; IEC 80-07, Problems with HPCI Turbine 011 System;
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IEC 80-17, Fuel Pin Damage Due to Water Jet from Baffle Plate
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IEC 80-19, Noncompliance with License Requirements for Medical
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Users; and, IEC 80-20, Changes in Safe-Slab Tank Dimensions.
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l The inspector had no further questions on the above items.
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Unresolved Items
Unresolved items are matters about which more infonnation is required in order to ascertain whether they are acceptable items, items of noncompliance, or deviations, The unresolved item identified during this inspection is discussed in paragraph 5.
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.
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