IR 05000333/1982019
| ML20027B938 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 09/10/1982 |
| From: | Doerflein L, Kister H, Linville J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20027B933 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.3.22, TASK-TM 50-333-82-19, NUDOCS 8209300347 | |
| Download: ML20027B938 (14) | |
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U.S. NUCLEAR REGULATORY COMMISSION 50333-820706 50333-820713 50333-820715 50333-820716 Region I 50323-820723
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50333-820730 82-19 Report No.
50333-820804
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50333-820801 50-333 Docket No.
DPR-59 C
License No.
Priority Category
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Power Authority of the State of New York Licensee:
P. O. Box 41 Lycoming, New York 13093 J. A. FitzPatrick Nuclear Power Station Facility Name:
Inspection at:
Scriba, New York Inspection conducted: Augu t 1-3 1982 h
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Inspectors: J A s Lihville Senicr Resident Inspector da'te dsigned
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date/ signed L.T.Doerflein{ResidentInspector b
dat signed
[ /d 7/
Approved by:
A.s H. 'B. Kiste'r, Chief, Reactor Projects 6te / signed Section 1C Inspection Summary:
Inspection on August 1-31 1982 (Raport No. 50-333/82-19)
Areas Inspected: Routine and reactive inspection during day and backshift hours by two Resident Inspectors (156 hours0.00181 days <br />0.0433 hours <br />2.579365e-4 weeks <br />5.9358e-5 months <br />) of licensee action on previous inspection findings; licensee event report review; operational safety verification; surveillance observations; maintenance observations; licensee event followup; review of plant operations; IE Bulletin followup; TMI Task Action plan item followup; containment vent and purge review, and review of periodic and special reports.
Results: No violations were observed in nine of eleven areas inspected. Two violations were observed in two areas. Failure to purge the containment through the Standby Gas Treetment System (detail paragraph 11), and Failure to provide annual radiation protection requalification (detail paragraph 8).
e209300347 820914 PDR ADOCK 05000333 G
PDR Region I Form 12 (Rev. April 77)
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DETAILS 1.
Persons Contacted
- R. Baker, Superintendent of Power V. Childs, Senior Resident Engineer
- R. Converse, Operations Superintendent
- M. Cosgrove, Site Quality Assurance Engineer M. Curling, Training Superintendent
- T. Doray, Training Specialist
- W. Fernandez, Technical Services Superintendent
- H. Keith, Instrument and Control Superintendent R. Liseno, Assistant Operations Superintendent
- A. McKeen, Assistant to Radiological & Environmental Services Superintendent
- C. McNeill, Resident Manager E. Mulcahey, Radiological & Environmental Services Superintendent T. Teifke, Security & Safety Superintendent V. Walz, Senior Plant Engineer
- R. Wiese, Acting Maintenance Superintendent The inspectors also interviewed other licensee personnel during this inspection including shift supervisors, administrative, operators, health physics, security, instrument and control, maintenance and contractor personnel.
- Denotes those present at the exit interview.
2.
Licensee Action on P_revious Inspection Findings (Closed) VIOLATION (333/82-06-05): The inspector verified that an operator was stationed to continuously monitor drywell containment oxygen concentration using a portable oxygen analyzer when both installed oxygen analyzers had failed after the inspector informed the licensee that periodic monitoring did not meet the Technical Specification requirement. The inspector has verified that both installed containment oxygen analyzers have been restored to service. The inspector finds acceptable the licensee's comitment to maintain the installed equipment in a more reliable operating condition and to assure equivalent quality with respect to range and sensitivity of any temporary system which might have to be used in the future.
(Closed) VIOLATION (333/81-27-03): The inspector reviewed RPOP-7, Radiological Incident Review, Revision 0, dated July 1982 which was developed to document minor radiation protection incidents and to provide a feedback mechanism according to the licensee's February 18, 1982 response to this violation involving a failure to adhere to Radiation Work Permit requirements.
(Closed) UNRESOLVEDITEM(333/80-01-04): This item has recurred as reported in LER 82-17 and LER 82-38. The licensee's modification to correct this problem will be reviewed after it is installed during the 1983 refueling outage. This review will be conducted under open item (333/82-08-10). This item is closed for administrative purpose.
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(Closed) UNRESOLVEDITEM(333/82-02-01): The licensee received a Technical Specification change in Amendment 68 to the facility operating license to correct the High Pressure Coolant Injection (HPCI) system high steam flow isolation setpoint based on calculations contained in Mr. R. Ram's memo to the Resident Manager dated April 1, 1982 which is an attachment to safety evaluation JAF-SE-82-044.
Following the startup from the refueling outage in March 1982, the licensee performed testing on March 15, 1982 to verify the new HPCI high steam flow isolation signal setpoint, and on Marci 22, 1982 to verify the acceptability of the original Reactor Core Isolation Coaling (RCIC) system high steam flow isolation signal setpoint.
Inspector review of this testing is documented in paragraph 6 of inspection report 50-333/82-06.
(Closed)
INSPECTOR FOLLOWUP ITEM (333/82-01-01): The inspector verified by observation the installation of the three missing High Pressure Coolant In-jection system drain line restraints reported in LER 81-81.
(Clo:,ad) UNRESOLVED ITEM (333/79-01-05): The inspector reviewed Quality Control Inspection Report No. F-80-1147 to verify that those cells showing degradation in the LPCI batteries were replaced during the 1980 refueling outage. The licensee maintains the LPCI batteries in accordance with Technical Specification 4.9.F and IEEE 450-1980, and since the 1980 refueling outage has replaced three additional degraded cells in the "B" LPCI battery.
3.
Licensee _ Event Report (LER) Review The inspector reviewed LER's to verify that the details of the events were clearly reported. The insoector determined that reporting requirements had been met, the report was adequate to assess the event, the cause appeared accurate and was supported by details, corrective actions appeared appropriate to correct the cause, the form was complete and generic applicability to other plants was not in question.
LER's 82-31*, 82-32*, 82-33*, 82-34*, 82-35*, 82-36, 82-37*, and 82-38* were reviewed.
- LER's selected for onsite followup.
LER 82-31 reported that control rod 02-23 was inoperable due to a failed insert valve.
Inspector coverage of this event is covered in paragraph b of inspection report 50-333/82-15.
LER 82-32 reported a failure to inert the containment within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of going into the run mode due to nitrogen delivery and procedural problems.
The inspector will review the revision to the containment inerting procedure which clarifies the use of nitrogen directly from a contractor truck when it is issued.
(333/82-19-01)
LER 82-33 reported a station reserve line out of service because of a lightning strike.
In reviewing this event the inspector found a lack of understanding of the automatic reclose feature of the reserve line breakers.
The inspector will review the lesson plan being developed as a result of the Plant Operations Review Committee review of this event.
(333/82-19-02)
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LER 82-34 reported that licensed thermal power of 2436 MWt may-have been exceeded because both feedwater flow transmitters were found out of procedural
. tolerance during calibration. The inspector will review the details of the licensee review of this event in the followup report.-
(333/82-19-03)
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LER 82-35' reported that the torus level was out of specification during
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startup operations. The-inspector noted that the event date was incorrectly
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reported as August.4, 1982 instead of-July 16, 1982 and that the level was out of specification by as much as 0.69 inches as opposed to 0.1 to 0.2 inches indicated in the report. The' licensee plans to submit a revised LER by September 30,-1982. The inspector will review the revised LER later.
(333/82-19-04)
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LER 82-37 reported that F safety relief val _ve (SRV) was inoperable due to i
high tailpipe temperature.
Inspector review of this event is documented in.
paragraph 4.a(5)(b) of inspection report 50-333/82-15. The shutdown did not occur on August 15, 1982 as indicated in.that report because the licensee received a technical specification change which permitted continued operation
with only 10 SRV's operable in the safety mode.
LER 82-38 reported A main steam line 10 percent scram switch inoperable..The
' inspector will review the installation of the modification to prevent this
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recurring problem at the next refueling outage.
4.
Operational Safety Verification
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Control Room Observations (1) Daily, the: inspectors verified selected plant parameters and i
equipment availability to ensure compliance with limiting conditions i
for operation of the plant Technical Specifications.. Items checked included:
Power distribution limits;
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j Availability and proper. lineup of safety systems;
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Availability and proper alignment of onsite and offsite
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emergency power sources; Reactor Control panel indications;
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Primary containment temperature and pressure;
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Drywell to suppression chamber differential pressure;
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Standby Liquid Control Tank level and concentration, and
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Stack monitor recorder traces.
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(2) The inspectors directly observed the following plant operations to ensure adherence to approved procedures:
Routine Power.0peration
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Issuance of RWP's and Work Regnest/ Event / Deficiency forms
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(3) Selected lit annunciators were discussed with control room operators to verify that the reasons for them were understood and corrective action, if required, was being.taken.
(4) Shift turnovers were observed to ensure proper control room and shift manning. Shift turnover checklists and log review by the oncoming and offgoing shifts were also observed by the inspectors.
No violations were observed, b.
Shift Logs and Operating Records (1) Selected shift legs and operating records were reviewed to:
Obtain information on plant problems and operations;
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Detect changes and trends in performance; Detect possible conflicts with Technical Specifications or
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regulatory requirements; Determine that records are being maintained and reviewed as
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required, and Assess the effectiveness of the communications provided by
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the logs.
(2) The following logs and records were reviewed:
Shift Supervisor Log
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Nuclear Control Operator Log
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Night Orders
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Shift Turnover Check Sheet
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Protective Tag Record Log
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Jumper Log
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Liquid Radwaste Discharge Log
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Gaseous and Particulate Sample Logs
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Weekly Chemistry Status Log
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Air Sample Log
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(3) Findings:
On August 16, 1982, the inspector reviewed the placement of jumper 82-156 which involved installing two air pressure gages, one on each of the starting air compressors for C Emergency Diesel Generator (EDG). The jumpers were placed under work request WRED 93/20925 because the installed gage on the local panel was out of calibration. The job was classified as QC Category M by the licensee because the air compressors them-selves are not safety related..However, the inspector stated that the air lines up to the compressors are safety related and that this jumper involved replacement of the air sensing line leading to the compressor.
In addition, the inspector considered tinis a com)onent substitution rather than a jumper meaning that it should lave been classified and evaluated at least as a minor modification rather than a jumper. The licensee acknowledged the inspectors concern about this matter and removed the temporary gages.
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Plant Tours (1) During the inspection period, the inspectors made observations and conducted tours of plant areas including the following:
Control Room
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Relay Room
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Reactor Building
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Turbine Building
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Diesel Generator Rooms i
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Electric Bays
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Pumphouse-Screenwell
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Standby Gas Treatment Building
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Radwaste Building
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Crescent Rooms (
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Battery Rooms
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Cable Tunnels i
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Plant Perimeter l
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(2) During the plant tours, the inspector conducted a visual inspection of selected piping between containment and the isolation valves for leakage or leakage paths. This included verification that manual valves were shut, capped and locked when required and that motor operated or air operated valves were not mechanically blocked.
Other items verified during the plant tours included:
(a) Fire Protection Conditions No significant fire hazards existed.
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Extinguishing equipment, fire alams, actuating controls,
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fire fighting equipment and emergency equipment was operable.
Ignition sources and flammable material were properly
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controlled.
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(b) Housekeeping / Cleanliness Conditions Critical clean areas like the refueling floor were
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properly controlled.
Combustible material was properly controlled.
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(c) Radiation Protection Controls Surveys were properly performed.
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Radiation Protection instruments were calibrated and
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Radiation Work Permits were complete, appropriate and
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followed.
Methods used to control exposures of those working in
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high radiation areas were appropriate.
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Activities in radioactive waste system areas were conducted in accordance with approved procedures.
(d) Physical Security Plan Implementation The security organization appeared to be properly manned
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and capable of performing its assigned function.
Isolation zones were clear.
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Persons and packages were checked prior to entry into
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the protected area.
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Vehicles were properly searched and escorted or
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controlled within the protected area.
Persons within the protected area displayed photo
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identification badges and persons requiring escorts were properly escorted.
Compensatory measures were employed when required by
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secerity equipment failure or impairment.
Protected area and vital area barriers were not degraded
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and accese to these areas was properly controlled.
(e) Verification of adherence to selected Technical Specification Limiting Conditions for Operation.
No violations were oberved, d.
Tagout Verificati_on, The inspectors verified that the following safety related protective tagout records (PTR's) were proper by observing the positions of breakers, switches and/or valves.
PTR 820809 on the "C" Emergency Diesel Generator Air Compressor
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93-AC-C1.
PTR 820837.on the Fire Protection Water System
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PTR 820839 on the Fire Protection Water System Air Compressor
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76-AC-11.
PTR 820855 on the "A" and "C" Emergency Diesel Generators
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PTR 820871 on the Standby Liquid Control System
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j No violations were observed.
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Radioactive Waste Systems Controls (1) The inspector witnessed selected portions of two liquid radioactive
releases to verify the following:
The required release approvals are obtained.
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The required samples are taken and analyzed.
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The radioactive waste system was operated in accordance with
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approved procedures, i
The inspector observed the release of Batch 4529, A Laundry Drain Tank, on August 6, 1982 and Batch 4530, B Laundry Drain Tank, on August 10, 1982.
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(2) On August 24, 1982, the inspector observed the survey of radioactive waste shipment number 08-82-047L. The inspector also reviewed the shipment records and observed that the shipment
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was properly labelled.
No 'iolations were identified, f.
Emergency System Operability _
The inspectors verified operability of the Emergency Diesel Generators and Emergency Power, Low Pressure Coolant Injection, and the Standby Gas Treatment Systems. The following items were included in the system verification:
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Confirmation that each accessible valve in the primary flow path was in the correct position.
Confinnation that power supplies and breakers are properly aligned
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for components that must activate upon an initiation signal.
Visual inspection of the major components for leakage and other
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conditions which might prevent fulfillment of their functional requirements.
The inspectors also verified the operability of the Containment Atmosphere Dilution (CAD) system by performing a complete walkdown of the accessible portions of the system. The following were included in the system verification:
Confirmation that the licensee's system lineup procedures match
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plant drawings and the as-built configuration.
Verification that valves are in the proper position, have power
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available and are locked (sealed) as required.
Verification that system instrumentation is properly valved in.
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Verification that there are no obvious deficiencies which might
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<!egrade system performance such as inoperable hangers or supports.
During the verification of the Containment Atmosphere Dilution System, the inspector noted the folhawing discrepancies between the as-built condition, drawing OP-37-1 A and drawing FM-18A:
Valves CAD 61, SV 201 A and SV 201B are not shown on drawing FM-18A.
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Vent valves installed between each of the truck fill /inerting
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connections and their isolation valves (CAD 11A, llB and 60) are not shown on drawing 0P-37-1 A or FM-18A.
Valves PCV 134A and B are not shown on drawing FM-18A.
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10.3 The Dewar fill connection, which comes off the "B" nitrogen
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storage tank, is incorrectly shown as coming off the "A" nitrogen storage tank on drawing FM-18A.
The inspector will verify the correction of these discrepancies during a later inspection.
(333/82-19-05)
5.
Surveillance Observations The inspector observed portions of the surveillance procedures listed below-to verify that the test instrumentation was properly calibrated, approved procedures were used, the work was performed by qualified personnel, limiting conditions for operation were met, and the system was correctly restored following the testing:
PSP-1, Reactor Water Sampling and Analysis, Revision 5, dated December 8,
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1981, performed August 4,1982.
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F-ISP-83, Condenser Lcw Vacuum, Revision 7, dated March 1981, performed August 5, 1982.
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F-ISP-3-1, Reactor Low-Low Water Level, Revision 11, dated March 1981, performed August 25, 1982.
F-ST-26I, Reactor Water Cleanup Isolation Logic System Functional Test,
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Revision 3, dated May 19, 1982, performed August 30, 1982.
The insputor also witnessed all aspects of F-ST-24A, RCIC Pump Operability Test, Revision 10, dated August 18, 1982, perfomed August 25, 1982.
Observations were made to verify that:
The surveillance procedure confoms to technical specification
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requirements and had been properly approved.
Limiting Conditions for Operations for removing equipment from service
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are met.
l Testing was performed by qualified personnel.
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Test results met technical specification requirements.
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i The surveillance test documentation was reviewed.
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Equipment was properly restored to service following the test.
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No violations were observed.
6.
Maintenance Observations The incectors observed portions of various safety related maintenance activities. Through direct observation and review of records, they determined that:
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Redundant components were operable.-
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These activities did not violate the limiting conditions for operation.
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Required administrative approvals and tagouts were obtained prior to
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initiating the work.
Approved procedures were used or the activity was within the " skills of
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the trade."
Appropriate radiological controls were properly implemented.
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Ignition / fire prevention controls were properly implemented.
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Equipment was properly tested prior to returning it to service.
During this inspection period, the following activities were observed:
WR 76/12931 on the Fire Protection Mater System Air Compressor 76-AC-11.
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WR 13/20992 on Condensate Storage Tank Level switch 13-LS-77B.
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WR 93/20925 on "C" Emergency Diesel Generator Air Pressure gage.
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WR 93/14879 on "A" and "C" Emergency Diesel Generator Space Heaters.
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WR 11/19188 on the Standby Liquid Control System Explosive Valves.
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No violations were observed.
7.
Licensee Event Followup a.
At about 3:00 a.m. on August 4,1982, the licensee informed the inspector that a shutdown was in progress because drywell unidentified leakage was approaching the 5 gallon per minute (gpm) limit of Technical Specification 3.6.D.1.
When the inspector arrived on site at 3:30 a.m. the leakage was 4.4 gpm and reactor power had been reduced to sixty percent. In an effort to stop the leakage the licensee alectrically backseated, by physically holdinc in the contactor until maximum limitorque motor current was reached, several inside containment isolation valves in the core spray, reactor core isolation cooling, and high pressure coolant injection systems.
In addition, the licensee backseated the discharge an1 discharge bypass valves in the B reactor recirculation loop. When the latter two valves were backseated, the drywell unidentified leakage decreased to its previous rate of about 1 gpm. To demonstrate continued operability of the backseated valves and to confirm the source of the leakage, the licensee stroked closed or partially closed each of the valves which had been backseated. When the leak did not reappear, the licensee began a normal power ascension back to full power.
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At 2:45 p(SBLC) System inoperable for approximately three hours tothe licen
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August 30,1982, b.
Control replace both explosive valves and their primers. The explosive valves were inadvertently " fired" during the performance of surveillance test F-ST-26I, Reactor Water Cleanup (RWCU) Isolation Logic System Functional Test, Revision 3, dated May 1982. When testing the SBLC system initiation isolation feature of RWCU, the surveillance procedure
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requires that the fuses for the four primers be removed. the SBLC pump
"A" breaker be opened, the SBLC control switch placed in "A" pump run position, verification that there is a RWCU isolation signal, returning the SBLC control switch to stop, shutting the SBLC pump "A" breaker, and then repeating the sequence for the "B" pump. 'owever, when removing the primer fuses, the operator removed the wrong fuses. He was then mislead by the fact that one of the fuses he did remove provided power to the primer continuity indicator lights.
The primers, which receive their power downstream of the SBLC pump breakers, are set up so that all four primers will fire when the SBLC control switch is taken to the "A" or "B" pump run position. Consequently,
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during the surveillance testing when the control switch was placed in the
"A" pump run position, the "B" explosive valve was fired. The "A"
explosive valve was similarly " fired" when the "B" pump run position was
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tested. No SBLC injection occurred as the pump breakers were open when the respective pump was selected on the control switch.
Following restoration from F-ST-26I, it was noted that primer continuity meters were reading approximately 2.5 and 3.5 milliamps (ma) for the A and B e plosive valve primers, respectively. The primer continuity circuit is set up to indicate a loss of continuity)to the primers if current drops less than 3 ma (normal value > 5 ma. The fact that the
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"B" continuity circuit did not provide a loss of continuity indication even though the explosive valve had fired,.is under review by the licensee.
The inspector will-review the results of this review in a subsequent inspection.
(333/82-19-06)
8.
Review of plant Operations l
Training i
Using the current site access list, the inspector selected a sample of
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twenty-eight contractor personnel, qualified in radiation protection, to verify that they had received the annual retraining in radiation protection as required by Indoctrination and Training Procedure ITP-3, " General Employee
Training," Revision 2, dated November 19, 1980. The inspector reviewed the
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attendance records for all radiation protection training conducted during 1981 and 1982 and determined that, as of August 27, 1982, sixteen of the twenty-eight individuals examined had not received radiation protection retraining during the past fifteen months. The inspector also noted that the licensee i
has not established administrative controls to ensure contractor personnel receive the required retraining. The inspector informed the licensee that
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this is a violation of the requirements of training procedure ITP-3, and Technical Specification 6.4.
(333/82-19-07)
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II.K.3.22, Automatic Switchover of Reactor Core Isolation Cooling System Suction i
By review of modification package F1-81-005, the inspector determined that this j
item is complete with the exception of the ur, qualified level ~ switches used.
These level switches will be replaced by qualified level switches in accordance
with the licensee's program for implementation of environmental qualification
of all electrical equipment.
11. Containment Vent and Purge Review While performing a survey to determine the extent to which the facility is operated with its large containment vent and purge system valves open, the inspector detemined that the licensee is not routinely exhausting the
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primary containment through the Standby Gas Treatment System (SBGT) filters as required by Environmental Technical Specification (ETS) 2.3.B.10.
It has been the licensee's normal practice since June 1,1979 to leave the 20 inch torus purge containment isolation valves open continuously during nomal operation to vent the nitrogen in-leakage from the containment instrument
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i nitrogen system and maintain the 1.7 psid required by Technical Specification 3.7.A.7.
Downstream of these valves the 6 inch bypass valve to the SBGT system l
is also normally left open. However, downstream of this valve are the normally i
closed inlet valves to the 2 SBGT trains, and three nomally closed suction valves, one from the containment purge system and two from the reactor building above and below the refuel floor. Since the reactor building is normally maintained at a negative pressure with respect to the containment and to the
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outside environment, the small routine leakage is probably back into the reactor building through the suction valves which are close to the 6 inch bypass valve rather than through the long run of piping, the closed SBGT train isolation valves, the HEPA and charcoal filters and out to the atmo-sphere through the stack. When the inspector informed the licensee that the routine purging of the primary containment was not through the SBGT system as required by ETS 2.3.B.10, the licensee immediately initiated operation of one
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SBGT train to comply with the requirement.
(333/82-19-08)
In addition to the ETS requirement, the licensee had comitted to ensuring i
that all releases from the containment are through the stack via the SBGT
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system in section 5.2.3.6 of the FSAR, in letter no. JPN-79-50 dated
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November 15, 1979 and in section 1.2.2 of.the 10 CFR 50 Appendix I submittal.
The inspector further noted that the valve lineup table of F-0P-37, Nitrogen
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Ventilation and Purge; Containment Vacuum Relief and Containment Differential Pressure system, Revision 9, dated March 18, 1982, temporary change dated l
April 9,1982, indicates that the 20 inch torus exhaust valves 27A0V117 and 27A0V118 are normally open and haw been since Revision 1, dated June 1979 except between March 18, 1982 and April 9, 1982 before Revision 9 was changed, i
while the licensee was attempting to operate with only the two inch bypass valves 27MOV117 and 27MOV123 open instead. This was apparently unsuccessful
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j because of the volume of nitrogen in-leakage from the containment instrnent nitrogen. Operation with 27A0V117 and 27A0V118 normally open is not consistent
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with Table 3.7-1 of Technical Specifications which indicates that the
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suppression chamber exhaust valves are normally closed.
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12. Review of Periodic and Special Reports Upon receipt, the inspector reviewed periodic and special reports. The review included the following:
Inclusion of information required by the NRC; test results and/or supporting information consistent with design predictions and perfomance specifications; planned corrective action for resolution of problems; reportability and validity of reported information.
The following periodic report was reviewed:
Operating Status Report for the month of July 1982, dated August 6,1982.
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No violations were observed.
13. Exit Interview At periodic intervals during the course of this inspection, meetings were held with senior facility manageme-t to discuss inspection scope and findings.
On September 1,1982, the inspector met with the licensee representatives (denoted in paragraph 1) and sumarized the scope and findings of the inspection as they are described in this report.