IR 05000333/1982022
| ML20028B609 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 11/16/1982 |
| From: | Baunack W, Doerflein L, Kister H, Linville J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20028B601 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 50-333-82-22, NUDOCS 8212030078 | |
| Download: ML20028B609 (15) | |
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DCS NUMBERS 50333-820909 U.S. NUCLEAR REGULATORY COMMISSION 50333-821007 50333-821020 Region I.
82-22
Report No.
50-333 Docket No.
C DPR-59 Priority Category
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License No.
Power Authority of the State of New York Licensee:
P. O. Box 41 Lycoming, New York 13093-Facility Name:
J. A. FitzPatrick Nuclear Power Station Inspection at:
Scriba, New York Inspection conducted: October 1 31, 1982 bM- [
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Inspectors:
. Linvil'.e V5Fnior Re ent Inspector
/dat6 signed
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n/is/r v
~ t Inspector date signed f
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. Doerflin, Resi
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W.'H. Baunack, Project Inspector datd signed
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////4/rt Approved by: R. B. KisteF, Chief, Reactor Projects d(te s'igned
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Section 1C Inspection Summary:
Inspection on October 1-31,1982 (Report No. 50-333/82-22)
Areas Inspected:
Routine and reactive inspection during day and backshift hours by two resident inspectors and one regional base inspector (148 hours0.00171 days <br />0.0411 hours <br />2.44709e-4 weeks <br />5.6314e-5 months <br />) of licensee action on previous inspection findings; licensee event report review; operational safety verification; surveillance observations; maintenance observations; calibration; review of plant operations; followup on a plant trip; review of TMI Action Plan Category B requirements, and followup on Safety Relief Valve problems.
Results:
No violations were observed in nine of ten areas inspected. One violation was observed in one area; Failure to follow procedure (detail paragraph 5).
8212030078 821118 PDR ADOCK 05000333 G
PDR Region I Form 12 (Rev. April 77)
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DETAILS
1.
Persons Contacted R. Baker, Technical Services Superintendent V. Childs, Senior Resident Engineer
- R. Converse, Superintendent of Power M. Cosgrove, Site Quality Assurance Engineer M. Curling, Training Superintendent W. Fernandez, Maintenance Superintendent
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- J. Flaherty, Assistant Instrument and Control Superintendent
- H. Keith, Instrument and Control Superintendent
- R. Liseno, Acting Operations Superintendent-A. McKeen, Assistant to Radiological & Environmental Services Superintendent
- C. McNeill, Resident Manager J. Moyle, Manager-Operational Evaluation E. Mulcahey, Radiological & Environmental Services Superintendent D. Simpson, Training Coordinator T. Teifke, Security & Safety Superintendent V. Walz, Senior Plant Engineer The inspectors also interviewed other licensee personnel during this inspection including shift supervisors, administrative, operators,
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health physics, security, instrument and control, maintenance and contractor personnel.
- Denotes those present at the exit interview.
2.
Licensee Action on Previous Inspection Findings (Closed) VIOLATION (333/82-15-11): The inspector reviewed Quality Control (QC) checklists which have been prepared for initial use during the maintenance outage scheduled to begin in late October on repair of Residual Heat Removal pump discharge check valves, removal and replacement of Control Rod Drives, and replacement of Reactor Recirculation Pump seals. The inspector also reviewed new checklists under development for weld repair of a High Pressure Coolant Injection System drain line elbow and seal gasket replacement of the feedwater containment isolation non return valves.. These new QC checklists appear to be more detailed and specifically applicable to the job than previous checklists.
(Closed) VIOLATION (333/82-10-02): The inspector reviewed Operations Department Standing Order (0DS0) 19, Procedure for Control of Non-Security Related Keys Issued to the Operations Department, Revision 0, dated September 30, 1982. This procedure formalizes the administrative controls over keys to High Radiation Areas required by Technical Specification 6.ll(A)2.
(Closed)
INSPECTOR FOLLOWUP ITEM (333/82-15-14): The inspector reviewed ODSO 18, Equipment Status Control, Revision 0, dated September 23, 1982.
This procedure establishes administrative controls to ensure that equipment which is deficient or out of service can be easily identified by tags or stickers whether or not protective tagging is necessary.
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(Closed)
INSPECTORF5LLOWUPITEM(333/82-15-04): The licensee has decided to document all Average Power Range Monitor calibrations as x required by RAP 7.3.1.rather than revise the procedure to require documentation of only those APRM calibrations required by Technical s.
Specifications.
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(Closed)
INSPECTORFOLLOWUPITEM(333/82-08-05): Through discussions with licensee personnel, the inspector detemined that the flow
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oscillations caused by the-Recirculation System motor generator set were caused by improper adjustment of the electrical brake on the scoop tube. The license'e' adjusted the brake to correct the problem.
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(Closed) VIOLATION-(333/82-10-03): The inspector reviewed the followhg licensee procedures to detemine that the licensee has implemented
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adequate administrative controls to ensure that fire doors remain in the proper position:
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FireProtecdiodPanelAnnunciatorResponseProcedureforSupervised
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Fire Doors, Revision 0, dated September 29, 1982.
F-ST-40, Daily Surveillance and Instrument Check, Revision 7, dated
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September 16, 1932, for unsupervised fire doors with automatic hold open and release mechanisms.
F-ST-76A, Fire Protection System Weekly Checks, Revision 2, dated
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August 26, 1982, for nomally closed and locked fire doors.
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3.
Licensee Event Report (LER) Review The inspector reviewed LER's to verify that the details of the events were clearly reported. The inspector detemined that reporting require-ments 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 fom was complete and generic applicability to other plants was not in_ question.
LER's 82-44*, 82-46, 82-47*, and 82-48 were revieted.
- LER's selected for onsite followup.
I LER 82-44 reported the failure of the motor bperator on' Low Pressure
Coolant Injection Valve 10 MOV 25A.
Inspector coverage of this event is covered in, paragraph 6 of inspection report 50-333/82-lV. *
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LER 82-47 reported a reactor trip when the "D" steam line inboard Main Steam Isolation Valve disc separated from the valve stem.
Inspector
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coverage of this e' vent is covered in paragraph 9 of this inspection repora.
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4.
Operational Safety Verification
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Control Room Observations
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J Daily, the inspectors verified selected plant parameters and equipment availability to ensure compliance with limiting conditions for operation of the plant Technical Specifications. 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. The inspector observed shift turnovers biweekly to ensure proper control room and shift manning. The inspectors directly observed the operations listed below to ensure
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adherence to approved procedures:
Routine Power Operation
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Reactor Startup
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Issuance of RWP's and Work Request / Event / Deficiency foms l
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No violati,ns were observed, b..
Shift Logs and Operating Records t
Selected shift' logs and operating records were reviewed to obtain
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infomation on plant problems and operations, detect changes and trends in perfomance, detect possible conflicts with Technical Specifications or regulatory requirements, detemine that records are being maintained and reviewed as required, and assess the effectiveness of the connunicaticies provided by the logs.
No violations were observed.
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/c.
Plant Tours
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During the inspection period, the inspectors made observations and conducted tours of the plant. During the plant tours, the inspectors
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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. The inspectors also checked fire protection, housekeeping / cleanliness, radiation protection, and physical security conditions to ensure compliance with plant procedures and regulatory
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reqtd rements.
No violations were observed.
d.
Tagout Verification
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The inspectors verified that the following safety-related protective
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tagout records (PTR's) were proper by observing the positions of breakers, switches and/or valves.
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PTR 820964 on the "A" Standby Gas Treatment System
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PTR 821001 on the Containment High Range Gamma Radiation Monitors
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PTR 821005 on Main Steam Isolation Valve 29-A0V-80D
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PTR 820963 on the Reactor Water Sample Isolation Valves 02-A0V-39
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and 40 PTR 821019 on "B" Emergency Service Water System
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PTR 821073 on "A" Emergency Service Water System
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During the verification of PTR 820964, the inspector noted that the position indicators for tagged manual valves SGT 3A and 3B indicated the valves were open when required to be shut. The operators involved in hanging the tags stated that they had used direction of movement of the valve handwheels as verification the valves were shut, not the position indicators, and that they had not changed the position of the valves. Since these valves are reqJired to be open during nomal operation, the licensee reviewed the valve installation and detennined that the valve operators were reverse acting, such that the handwheel is turned in the counterclockwise direction to shut the valve, and that the position indicators for the valves were in fact correct. The licensee shut valves SGT 3A and 3B prior to allowing work to commence on the Standby Gas Treatment system.
In addition, the licensee made an entry in the Operations Department Night Orders to infom all operators of the correct operation of these valves.
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Radioactive Waste Systems Controls The inspector witnessed selected portions of a liquid radioactive release to verify that the required release approvals were obtained,
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the required samples were taken and analyzed, the radioactive waste system was operated in accordance with approved procedures, and the release ~ control instrumentation was operable and in use.
The inspector observed the release of Batch 4596, B Laundry Drain Tank, on October 25, 1982.
The inspector: observed the surveys of radioactive waste shipment number 10-82-064L on October 6,1982, and shipment number WASH 82-55 on October 12,1982. The inspectors also reviewed the shipment records and observed that the shipments were properly labelled.
No violations were observed.
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f.
Emergency System Operability The inspectors verified operability of the following systems by ensuring that each accessible valve in the primary flow path was in the correct position, by confiming that power supplies and breakers were properly aligned for components that must activate upon an initiation signal, and by visual inspection of the major components for leakage and other conditions which might prevent fulfillment of their functional requirements.
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Reactor Core Isolation Cooling
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Low Pressure Coolant Injection.
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The inspectors also verified the operability of the following system by performing a complete walkdown of the accessible portions of the system. During the system verification, the inspectors-confimed that the licensee's system lineup procedures matched plant drawings and the as-built configuration; verified that valves were in the proper position, had power available and were locked (sealed) as required; verified that system instrumentation was properly valved in; and verified that there are no obvious deficiencies which might degrade system perfomance such as inoperable hangers or supports.
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No violations were observed.
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Surveillauce 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 perfomed by qualified personnel, limiting conditions for operation were met, and the system was correctly restored following the testing:
F-ST-34A, Reactor Building Ventilation, Drywell Isolation Valve.
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Atmosphere Control Valve, Drywell Sump Valve, and TIP Withdrawal l
Logic Functional and Simulated Automatic Actuation Test, Revision 8, dated March 2,1982, perfomed on October 5,1982.
F-ST-24D, RCIC Subsystem Automatic Isolation Logic System Functional
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Test, Revision 8. dated May 19, 1982, perfomed on October 19, 1982.
F-ST-22C, ADS Logic System Functional Test, Revision 5, dated May 19,
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1982, performed on October 20, 1982.
F-ST-8D, ESW Pump Flow Rate Test, Revision 9, dated January 27, 1982,
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perfomed on October 26, 1982.
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F-ST-3D, Core Spray MOV Velve Operability Test, Revision 4, dated
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May 19,1982, performed on October 26, 1982.
F-ST-3C, Core Spray Pump Operability Test and Keep Full Level Switch
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Functional Test, Revision 7, dated May 19, 1982, performed on October 26, 1982.
F-ST-9D, EDG Inoperative Test / Loss of 115 KV Reserve Power / Loss of
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Station Battery, Revision 4, dated June 9,1982, perfomed on October 26, 1982.
F-ST-2R, RHR Service Water Pump and MOV Operability Test, Revision 6,
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dated May 19, 1982, performed on October 26, 1982.
The inspector also witnessed all aspects of the following surveillance test to verify that the surveillance procedure conformed to technical specification requirements and had been properly approved, limiting conditions for operation for removing equipment from service were met, testing was perfomed by qualified personnel, test results met technical specification requirements, the surveillance test documentation was reviewed and equipment was properly restored to service following the test.
F-ST-2A, RHR Pump Flow Rate Test, Revision 8, dated May 19, 1982,
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perfomed on October 26, 1982.
During the perfomance of F-ST-24D, Reactor Core Isolation Cooling (RCIC)
Subsystem Automatic Isolation Logic System Functional Test, the inspector noted that for step VII.C.2, which tests the automatic switchover of the RCIC pump suction from the Condensate Storage Tank (CST) to the Suppression Pool on low CST level, the technician did not simulate the icw CST level at level switches 13-LS-76A and 13-LS-77B as required.
Instead, he had lifted the leads at terminals BB66 and BB68 in panel 9-30 which de-energized primary relays 13-AK-60,13-AK-61,13-AK-62 and 13-AK-63, the same relays controlled by the CST level switches. When the inspector questioned this, the technician and shift supervisor stated that this was considered acceptable since the logic system functional test verifies the proper logic and system response once the primary relays de-energize and since the circuit from the sensors (level switches) to the primary relays is checked monthly during the
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instrument functional test. Upon completion of the surveillance test, the
inspector reviewed the completed data sheet and noted that step VII.C.2 5:as l
signed off as having simulated the CST low level at level switches 13-LS-/6A l
and 13-LS-778. The inspector informed the licensee that the completed data i
sheet did not reflect what was actually done and that no administrative controls were used during the testing to ensure the proper lifting and restoration of leads.
In addition, the inspector pointed out that section 1.F.6 of the Technical Specifications defines a logic system functional test as a test of relays and contacts of a logic circuit from sensor to activated device. The inspector infomed the licenses that failure to perform F-ST-24D as written is a violation of Technical Specification 6.8.
(50-333/82-22-01)
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Maintenance Observations The inspectors observed portions of various safety-related maintenance activities to determine that redundant components were operable, these activities did not violate the limiting conditions for operation, required administrative approvals and tagouts were obtained prior to initiating the work, approved procedures were used or the activity was within the " skills of the trade," appropriate radiological controls were properly implemented, ignition / fire prevention controls were properly implemented, and equipment was properly tested prior to returning it to service.
During this inspection period, the following activities were observed:
WR 01-125/18235 on replacement of the "A" Standby Gas Treatment
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System charcoal filters.
WR 29/18796 on the repair of Main Steam Isolation Valve 29-A0V-86D.
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WR 29/16380 on the repair of Main Steam Isolation Valve 29/A0V-80D.
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WR 29/14976 on the installation of the inboard Main Steam Isolation
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Valve 90% limit switches.
WR 27/14977 on the pre-operational testing of the "A" Containment
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High Range Ganna Radiation Monitor.
WR 02-2-16027 on functional testing and repair of four Grinnell
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hydraulic snubbers.
No violations were observed.
Calibration a.
The inspector reviewed the 1982 functional test and calibration data sheets for the instrument channels listed below from the indicated Technical Specification (TS) Table to verify that the frequency of calibration specified in the Technical Specifications was met.
TS Table 4.1-1, " Reactor Protection System Instrument Functional
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Tests." Instrument Channels reviewed included Average Power Range Monitor (APRM) High Flux, APRM Inoperative, APRM Downscale, APRM Flow Bias, APRM High Flow in Startup or Refuel, High Reactor Pressure, Reactor Low Water Level, High Water Level in Scram Discharge Instrument Volume, Main Steam Line High Radiation, Main Steam Line Isolation Valve Closure, and Reactor Pressure Pennissive.
TS Table 4.1-2, " Reactor Protection System Instrument Calibration."
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Instrument Channels reviewed included APRM Flow Bias Signal, High Reactor Pressure, Reactor low Water Level, High Water Level in Scram Discharge Instrument Volume, Main Steam Line Isolation Valve
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Closure, Main Steam Line High Radiation, and Reactor Pressure Permissive.
TS Table 4.2-1, " Minimum Test and Calibration Frequency for
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Primary Containment Isolation System." Instrument Channels reviewed included Reactor Low-Low Water Level, Main Steam High Flow, Main Steam Low Pressure, and Reactor Water Cleanup High Temperature.
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TS Table 4.2-2, " Minimum Test and Calibration Frequency for Core
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and Containment Cooling Systems."
Instrument Channels reviewed included Reactor Water Level, Drywell Pressure, Automatic Depressurization System-Low Pressure Coolant Injection or Core Spray Pump Discharge Pressure Interlocks, Core Spray Sparger Differential Pressure, High Pressure Coolant Injection (HPCI)
and Reactor Core Isolation Cooling (RCIC) Steam Line High Flow, and HPCI Suction Source Levels.
No violations were observed.
b.
The inspector reviewed th.. completed test documentation from a. above on a sampling basis to vuify that the test documentation was complete, the acceptance criteria was met, the proper revision was used, and the individuals performing the calibrations were qualified.
The following calibration procedures and associated data were selected for this review:
F-ISP-3, Reactor High/ Low Water Level Instrument Functional Test /
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Calibration, Revision 9, dated February 3, 1982.
Data was reviewed for tests perfomed April 2,1982, May 24,1982 and July 19, 1982.
F-ISP-5, Reactor High Pressure Instrument Functional Test /
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Calibration, Revision 5, dated April 21, 1981.
Data was reviewed for tests perfomed February 22, 1982, May 19, 1982 and August 10, 1982.
F-ISP-4-1, Drywell High Pressure (HPCI, LPCI, RHR, SBGT, EDG, Core
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Spray) Functional Test / Calibration, Revision 5, dated April 8,1981.
Data was reviewed for tests perfomed February 22, 1982, June 7, 1982 and August 3, 1982.
F-ISP-5-3, Reactor Low Pressure-LPCI Core Spray Permissive Instrument
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Functional Test / Calibration, Revision 6, dated October 22,1981.
Data was reviewed for tests perfomed March 22, 1982, April 20, 1982 and August 10, 1982.
F-ISP-49, Reactor Water Cleanup Area High Temperature Instrument
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Functional Test / Calibration, Revision 7, dated April 14, 1981.
Data was reviewed for tests performed February 20, 1982, June 21, 1982 and August 16, 1982.
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F-ISP-66-1, Scram Discharge Volume High Water Level Instrument
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Functional Test / Calibration, Revision 7, dated September 23, 1982.
Data was reviewed for tests performed March 17, 1982, May 11, 1982 and August 6, 1982.
F-ISP-70, Reactor High Pressure Permissive Instrument Functional
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Test / Calibration, Revision 6, dated June 8, 1981. Data was reviewed for tests performed April 12, 1982, June 7, 1982 and August 5,1982.
F-ISP-75, Condensate Storage Tank Low Water Level (HPCI), Revision 4,
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dated November 17, 1981. Data was reviewed for tests performed April 12, 1982, June 7, 1982 and August 5, 1982.
F-ST-lG, Main Steam Line High Radiation Functional Test,_ Revision 5,
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dated May 19,1982. Data was reviewed for tests performed June 20, 1982, August 30,1982 and October 24, 1982.
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F-ST-5B, APRM Instrument Functional Test (Run Mode), Revision 8, dated June 16,1982.
Data was reviewed for tests performed June 29, 1982, August 27, 1982 and October 1,1982.
F-IMP-2-3.5, Reactor Vessel Metal Temperature Instrumentation Test
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and/or Calibration, Revision 2, dated August 10, 1981. Data was reviewed for calibration of instrument 02-3-TR-90 performed February 10, 1981.
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F-IMP-ll.1, Standby Liquid Control System, Revision 2, dated June 11, 1981.
Data was reviewed for calibration of instrument ll-PI-65 performed February 9,1981.
F-IMP-23.3, High Pressure Coolant Injection Sy tem Flow Indication
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Instrument Test and/or Calibration, Revision 2, dated August 10, 1981. Data was reviewed for calitration of flow instrument 23-FI-108-1 performed January 23, 1981.
No violations were observed.
c.
The inspector examined the technical content of the procedures listed below to verify that a satisfactory calibration would result.
Included in this review was a determination that:
Procedures have been reviewed and approved as required by
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Technical Specifications.
Calibration will be to the required accuracy.
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As-found and as-left conditions will be recorded.
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Acceptance values for trip settings conform with the Technical
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Specification requirements and allowances are made for instrument drif _
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Calibration equipment used will be traceable.
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Return to service is controlled such that the instrument will
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be properly returned to service.
Calibration sheets are initialled to identify the instrument
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technician perfoming the calibration.
The following procedures were selected for the above review:
F-ISP-5, Reactor High Pressure Instrument Functional Test /
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Calibration, Revision 5, dated April 21, 1982.
F-ISP-3-2, Reactor Lo-Lo/Lo-Lo-Lo Water Level (HPCI, LPCI, RHR,
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ADS, Core Spray) Instrument Functional Test / Calibration, Revision 11, dated January 27, 1982.
F-ISP-46, Main Steam Line High Flow Instrument Functional Test /
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Calibration,' Revision 5, dated May 13, 1981.
F-ISP-60. Reactor Protection System Time Response Check, Revision 7,
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dated May 13, ~ 198,1.
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i-F-ISP-65-1, Main Steam Line Isolation Valve Closure (RPS) Instrument
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Calibration, Revision 2, dated June 2,1981.
i F-ISP-66-1, Scram Discharge Volume High Water Level Instrument
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Functional Test / Calibration, Revision 7, dated September 23, 1982.
l F-ISP-75, Condensate Storage Tank Low Water Level- (HPCI), Revision 4,
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~ dated November 17, 1981.
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F-IMP-2-3.5, Reactor Vessel Metal Temperature Instrumentation Test and/cr Calibration, Revision 2, dated August 10, 1981.
F-IMP-ll.1, Standby Liquid Control System, Revision k., dated
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June 11, 1981.
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F-IMP-23.3, High Pressure Coolant Injection (HPCI) System Flow
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l Indication Instrument Test and/or Calibration, Revision 2, dated l
August 10, 1981.
No violations were observed.
d.
The inspector reviewed the licensee's system for calibration of components associated with safety-related systems or functions but not l
specified in the Technical Specifications as requiring calibration to l
detemine if a backlog of past due calibrations is developing.
In addition, the inspector selected ten installed components, listed below, used to make Technical Specification related measurements or used by a
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operator during nomal operations or post-accident to verify that they are included in the licensee's system and that reviewed and approved procedures are available to calibrate the component.
Components associated with safety-related systems or functions selected for review included: Jet Pump Loop A Flow 02-3-FI-92A, Standby Gas Treatment System Flow Indication 01-125-FI-100, Reactor Vessel metal temperature 02-3-TR-90, Residual Heat Removal Loop "A" Flow 10-FI-133A, Standby Liquid Control Pump Discharge Pressure 11-PI-65, "A" Core Spray Pump Discharge Pressure 14-PI-48A, High Pressure Coolant Injection Flow Indication 23-FI-108-1,
"A" Emergency Service Water Pump Discharge Pressure 46-PI-ll2A, Reactor Core Isolation Cooling Pump Discharge Pressure 13-PT-60, and "A" Emergency Diesel Generator tachometer 93RPMA.
No violations were observed.
8.
Review of Plant Operations Review and Audit On October 12 and 13, 1982, the inspectors attended Plant Operating Review Comittee (PORC) meetings 82-83 and 82-84. The inspectors verified that the Technical Specification membership and frequency requirements were met. The inspectors also reviewed the minutes from these PORC meetings and detemined that they accurately reflected the decisions and recommendations made by the PORC members in the meeting.
No violations were observed.
9.
Followup on a Plant Trip At 11:43 a.m. on October 7,1982, the reactor scramed from full power on high neutron flux as the result of a reactor pressure spike. The pressure spike occurred when the "D" steam line inboard Main Steam Isolation Valve (MSIV) main disc separated from the valve stem and blocked flow in the "D" steam line. Recent mechanical failures of these Rockwell-Edward Flite Flow Stop Valve MSIV's were the subject of IE Infomation Notice No. 81-28. All systems functioned properly and therewas no Emergency Core Cooling System Actuation following the scram. There was no release associated with this trip.
Upon disassembling the "D" steam line inboard MSIV, the licensee found that the stem disc antirotation pin had sheared. This allowed the stem disc to rotate and unscrew itself from the valve stem. This failure of the stem-to-stem disc threaded connection also causes the main disc to be separated from the valve stem since the main disc is supported and attached to the valve stem by the stem disc. The licensee replaced the stem and stem disc on the "D" steam line inboard MSIV and to prevent recurrence, installed two antirotation pins 900 apart in the stem disc. The licensee also disassembled the "D" steam line outboard MSIV and replaced its stem and stem disc. As with the inboard MSIV, two antirotation pins were installed in the outboard MSIV stem disc. Both the inboard and outboard "D" steam line MSIV's successfully passed a Type C leak rate test following the repairs.
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10. Review of TMI Action Plan Category B Requirements
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A review was conducted of the licensee's programs for upgrading of Reactor
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and Senior Reactor Operator Training and Qualifications (NUREG-0737, Item 1.A.2.1), and Training for Mitigating Core Damage (NUREG-0737, Item II.B.4).
The following records were reviewed:
Indoctrination and Training Procedure N.5, Licensed Operator
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Requalification, Revision 4, dated August 7, 1980.
Requalification Training Schedules, Cycles 81-1 through 81-9.
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Training Lecture Outlines for Requalification Training, Cycles
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81-1 through 81-9.
Selected Individual Training Files.
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Records of Simulator Training.
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Findinas:
The fonnal licensed operator requalification + raining program consists of one week of training out of each six weeks of work. This includes one week of simulator training (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in 1981), and the remainder classroom training (32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> / week) including examinations.
Records are maintained to show the training received by each individual.
The licensed operator retraining program corisists of classroom and walkthrough training in the following categories:
Reactor Theory and Principles of Operation
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Features of Facility Design, Fuel Handling and Core Parameters
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Instrumentation and Centrol
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Thermodynamics, Heat Transfer, and Fluids
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General and Specific Operating Characteristics
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Safety and Emergency Systems
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Standard Procedures
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Emergency Procedures
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Administrative Procedures, Conditions and Limitations
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Radiation Control and Safety, Radioactive Materials Handling, Disposal
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Recognizing and Mitigating Consequences of Severe Core Damage
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Waste Handling
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Self Monitoring
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First Aid
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Fire Protection and Spill Procedure
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Refueling Interlocks and Procedures
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E0P Walkthroughs
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QA, QC, 10 CFR and Administrative Procedures
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Technical Specifications
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The categories of themodynamics, heat transfer and fluids, and recognizing and mitigating consequences of severe core damage consist of approximately 57 hours6.597222e-4 days <br />0.0158 hours <br />9.424603e-5 weeks <br />2.16885e-5 months <br /> of classroom work and appear to contain all the training required by NUREG-0737.
In addition, related subjectgportions of which deal directly with mitigating core damage are taught. These subjects include features of facility design, instrumentation and control, general and specific operating characteristics, safety and emergency systems, emergency procedures, and emergency operating procedure walkthroughs. No specific number of hours can be assigned to the related subjects which are devoted to mitigating core damage. However, the total retraining associated with this subject is in excess of the 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> which is considered adequate for mitigating core damage training.
All personnel in the operating chain and shift technical advisors have i
l received the training for mitigating core damage through their participation in the retraining program with the exception of the Resident Manager.
A discussion with the Resident Manager, and a review of documentation, shows that the Resident Manager received equivalent training while participating in a 12 week General Electric SR0 Certification course.
In addition, the Resident Manager passed the Facility Training Program examination related to mitigating core damage after completing a Self Study Program.
To ensure that replacement personnel in non-licensed positions in the operating chain receive the required mitigating core damage training, the licensee has included the required training in procedure RPT-ll, Training for Professional and Supervisory Personnel.
While reviewing lesson plans and records relating to NUREG-0737 Retraining Requirements, in the areas of heat transfer, fluid flow, themodynamics, and mitigating core damage, certain matters which require further evaluation were identified. These included: available records make it difficult to detennine that all required training is given to each person required to be i
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trained, there appear to be deviations from facility administrative training procedure requirements, individuals training records appear to contain errors,
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some program requirements to be conducted on a two year cycle are repeated at the end of one year, and approximately 14% of alloted training time is devoted to self study. The licensec stated that a detailed audit of the training program would be conducted by November 30, 1982 to evaluate these preliminary findings. This item is unresolved.
(50-333/82-22-02)
No violations were observed.
11. Followup on Safety Relief Valve Problems During the October 7 to 17, 1982 cutage, the licensee repaired the inoperable themocouple on D Safety Relief Valve (SRV). Following the reactor startup on October 17, 1982, the licensee noted that the tailpiece temperature of the D SRV increased to approximately 3000F, indicating pilot valve seat leakage.
This situation is similar to one on the F SRV prior to the shutdown which was reviewed in Inspection Report 82-15 and for which the licensee declared the F SRV inoperable. The licensee replaced the topworks of the F SRV during the October outage. Since little or no industry data exists to actually correlate relief valve operability with pilot valve seat leakage, the licensee has not declared the D SRV inoperable. The licensee has, however, submitted a proposed change to the Technical Specifications (TS) to establish requirements for operation with SRV tailpiece temperatures greater than 2500F or more than
40 F above the steady state value. This proposed change is similar to TS Amendment 70 issued for the high tailpiece temperature on F SRV prior to the October outage and includes the requirements for reporting, perfoming an
engineering evaluation to justify continued operation and testing the valve in the as-found condition during the next cold shutdown of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more.
Continued operation with the high tailpiece temperature on D SRV is considered unresolved pending approval of this change to the Technical Specifications.
(50-333/82-22-03)
12. Unresolved Items Unresolved items are matters about which more infomation is required in order to ascertain whether they are acceptable items, violations or deviations. The unresolved items identified during this inspection are discussed in paragraphs 10 and 11.
13.
Exit Interview At periodic intervals during the course of this inspection, meetings were held with senior facility management to discuss inspection scope and findings.
On October 29, 1982, the inspector met with licensee representatives (denoted in paragraph 1) and summarized the scope and findings of the inspection as they are described in this report.