ML20210B512
| ML20210B512 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 07/15/1999 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20210B503 | List: |
| References | |
| NUDOCS 9907230164 | |
| Download: ML20210B512 (18) | |
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1 WISCONSIN ELECTRIC POWER COMPANY POINT BEACH NUCLEAR PLANT SIMULATOR FOUR-YEAR REPORT Contents j
1.0 I n t rod uction..................................... 2 2.0 Simulator Information............................... 2 3.0 Completed Certification Tests........................... 3 4.0 Certification Test Failures............................. 4 5.0 Simulator Certification Test Program Review 5
6.0 Simulator Modifications.............................. 6 7.0 Exceptions to ANSI /ANS 3.5 as endorsed by Reg Guide 1.149......... 6 i
1 8.0 Other Simulator Certification Issues....................... 7 9.0 Certification Test Schedule 8
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t 9907230164 990715 PDR ADOCK 05000266 R
PDR PBNP Simulator Four-Year Report Page 1 of 18 I
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l 1.0 Introduction This report is provided every four years on the anniversary of initial certification in accordance with 10 CFR 55.45. The report describes Simulator Certification Tests (SCT) perforrned from 1996 to 1999, discusses test failures from 1996 to 1999, and provides a schedule of tests to be performed over the next four year period.
2.0 Simulator Information Owner Wisconsin Electric Power Company Simulator Vendor Westinghouse Electric Corporation Reference Plant Point Beach Nuclear Plant Unit 1, Docket No. 50-266 Point Beach Nuclear Plant Unit 2, Docket No. 50-301 Type Two Loop Pressurized Water Reactor Rating 1518.5 MWT(each unit)
Certification Date July 22,1991 Type of Report Four Year PBNP Simulator Four-Year Report Page 2 of 18
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4 3.0 Completed Certification Tests The annual and quadrennial certification tests were completed as scheduled in the 1995 Four-Year Report, except as noted in Sections 5.1 and 9.1.
3.1 Test Number Change The previous report listed the test numbers as all starting with "14." The number "14" indicated that the test was a part of the initial Simulator Acceptance Testing Program and was deleted as it provides ro value.
3.2 Test Deletion The certification tests listed below have been deleted due to revisions in plant procedures that now incorporate these evolutions in SCT 6.6.3, Hot Shutdown to Cold Shutdown and SCT 6.6.6, Cold Shutdown to Hot Shutdown.
SCT 6.6.3.1 Plant Cooldown, OP-5A Part C (CVC)
SCT 6.6.6.1 Plant Heatup, OP-5A Part A (CVC) 3.3 Test Title Change The title for the certification test listed below has been changed to conform to the plant procedure title.
SCT 6.6.6 Cold Shutdown to Iow Power Has been changed to:
SCT 6.6.6 Cold Shutdown to Hot Shutdown The plant procedure identifiers that had been a part of the certification test titles have been removed as they served no useful purpose. Procedural references are now contained within the main body of certification tests.
PBNP Simulator Four-Year Report Page 3 of 18
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4.0 Certification Test Failures Two tests failed during the four year test period. The simulator performance has since been corrected.
SCT 3.3 Duty Cycle Test (1996)
This test failure was the result of the computers not meeting the average spare time acceptance criteria. Simulator discrepancy report (SDR) 96-0077 was written to track and correct this discrepancy. The discrepancy was resolved and the test has since been performed satisfactorily.
SCT 6.8.12.1 lead Reference Channel Fails (1996)
This test failure was a result of the malfunction not producing the desired effect. SDR 96-0128 was written to track and correct this discrepancy. The discrepancy was resolved and the test has since been performed satisfactorily.
4.1 Certification Test Discrepancies During the four years of certification test performance there were fourteen valid problems identified on the Unit 1 Simulator. This resulted in the following simulator discrepancy reports being written to track corrective actions and subsequent acceptance testing:
SDR Title 95-0159 Plant Computer Shows Tcold Greater Than Thot by 8 Degrees 96-0002 Unable to Operate RCP Breaker During Fill & Vent 96-0021 Irr420 Scaled Incorrectly on the Plant Computer 96-0039 Rod Insertion Limits are Incorrect 96-0077 Computers Do Not Meet Average Spare Time Acceptance Criteria.
l 96-0098 Hotwell Level /remp Increase 96-0102 Safety Injection Pump Motor Amps Too High 96-0110 Steam Generator. Narrow Range I2 vel Swell is Too High 96-0128 Malf EHC5 Doesn't Work - Load Reference Failure
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96-0162 Pressurizer level Setpoint Calculation i
96-0164 Unable to Drain During Fill and Vent 96-0165 U1 Turbine Stop Valve Test /Imse 20 MW at 370 MWE 96-0166 Audio Count Rate Not Functioning (SFR#1203) Accumulator Fill Valve Allows Too Much Flow i
i PBNP Simulator Four-Year Report Page 4 of 18
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- 5.0 Simulator Certification Test Program Review An internal review of the simulator certification test program was conducted during the second half of 1998. This review identified discrepancies and weaknesses with regard L
'to the certification testing program. These identified issues and corrective actions are described below.
i 5.1 Identified Issues:
l Eight cenification tests, listed in the 1991 report, were omitted from the 1995 report. The l
test number, title, date it should have been scheduled, and the date it was performed are listed below:
Test Title Date it should have ' Date been scheduled performed SCT 6.6.6.2 Removing RHR System from Operation 1995 1998 l
i SCT 6.6.6.3 Reactor Startup 1995 1998 SCr 6.8.5.3 less of Condenser Vacuum 1995 1998 i
SCT 6.8.8.8 Uncoupled Rod 1995 1996
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SCT 6.8.11.1 Diesel Generator Failure to Start 1995-1996
- SCT 6.8.12.1 Imad Reference Channel Fails 1995 1996 i
SCT 6.8.25.8 Steam Dump lead Reject Controller Failure 1997 1998 SCr 6.8.29.1 R'ID Bypass Line Failure 1997 1998 t
l In addition to the list above, a number of cenification tests had been modified without adequate review and approval, and some completed tests were not being reviewed in a timely manner.
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5.2 Corrective Actions:
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Simulator Guidelines were developed and issued. The guidelines placed simulator support personnel duties, responsibilities, and procedures under formal revision control. In addition to describing the current organization, the guidelines contain specific procedures which dictate tr.sk performance requirements.
The simulator certification tests are being reviewed, revised, and issued under formal revision control. Final test approval rests with the Simulator Review Committee chainnan.
l-PBNP Simulator Four-Year Report -
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Once issued, all tests will be run to ensure proper simulator performance. Discrepancies will be captured and corrected through the simulator discrepancy report process. The action plan specifies a goal of having all certification tests revised and successfully run on the simulator prior to December 31,1999. As of June 1,1999, thirty-seven out of eighty-nine certification tests have been revised and successfully run on the simulator.
6.0 Simulator Modifications The Simulator Review Committee establishes the priority for implementing modifications.
Numerous modifications have occurred, including some significant ones listed below:
I Turbine Driven Auxiliary Feedwater Pump Overspeed Trip Valve Replacement.
e Steam Generator Instrument Ixvel Tap Changes.
Completion of Emergency Diesel Generator Installation and Diesel Generator Governor Valve Replacement.
Chemical and Volume Control Makeup Water and Boric Acid Totalizer Controller Replacement.
Low Pressure Turbine Rotor Replacement.
Eighteen Month Core Load Installation (Udt 2).
Instmment Bus Transfer Switch Installation.
Altemate Shutdown Electrical Bus (B08 and B09) Installation.
Radiation Monitoring System Control Terminal (RMS CT) Installation.
7.0 Exceptions To ANSI /ANS 3.5-1985 As Endorsed By Reg Guide 1.149 7.1 Simulator Background Sounds The simulator background sounds, while not required, are mentioned as a consideration in ANSI /ANS 3.5-1985, Section 3.2.3. The original equipment is not functional and Wisconsin Electric has determined that the repair of this equipment is not a priority at this time. Background sounds will be considered if the simulator is converted to a PC platform.
PBNP Simulator Four-Year Report Page 6 of 18
7.2 RMS CT Modification The RMS CT modification was installed ahead of the plant. Installation has not been completed in the plant, presenting a possible conflict with ANSI /ANS 3.5-1985, Section 5.3. This is a difference that has existed for approximately three years. Current plans are to maintain the simulator in this configuration as the RMS CT is currently scheduled to be installed in the plant during the third quaner of 1999.
7.3 G01 Diesel Generator Control Section of Control Panel CO2 The G01 Diesel Generator Control section of control panel CO2 leads the plant. Installation j
has not been completed in the plant, pre.senting a possible conflict with ANSI /ANS 3.5-1985, Section 3.2.2. This is a difference that has existed for approximately three years.
I Current plans are to maintain the simulator in this configuration as the G01 Diesel Generator Control modification is currently scheduled to be installed in the plant during the refueling outage on Unit 2 in 2000.
8.0 Other Simulator Certification Issues 8.1 Unit 2 Simulator Certification The initial Simulator Certification Report, submitted July 1991, stated that " Unit 2 software will be formally tested as part of a long-term project to certify the Unit 2 portion of the PBNP simulator." Wisconsin Electric has determined that Unit 2 Sirnulator cenification is not economically viable at this time.
Acceptance testing demonstrated that Unit 2 hardware and software did not negatively impact Unit 1 performance. The software to support Unit 2 operation is essentially a copy of Unit 1. As such, it does not contain certain Unit 2 specific features nor model Unit 2 specific response. The execution of model software for both units is under the control of a single executive system running on two mainframe computers connected via shared memory. This system executes all model software for both units.
Wisconsin Electric has concluded that conducting selected training sessions on Unit 2, such as reactor startups and occasional equipment malfunctions, does provide a benefit to the operating crews and initial license class candidates. While response may not exactly mimic Unit 2 plant response, the training is very valuable in reinforcing the mirror image layout and addressing other human factors issues, such as communications between unit operators during a casualty response, diagnosis of multiple unit equipment failures, etc.
While the Unit 2 simulator will occasionally be used during training cycles,it will not be the unit of focus during performance of NRC required JPMs and operating examinations.
PBNP Simulator Four-Year Report Page 7 of 18
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l H.2 Incore Flux Mapping System The Incore Flux Mapping System will no longer be included as part of the simulator modeled system. The panel will be left in place for visual simulation only per Simulator Review Committee decision. It should be noted that crew members are not tasked with operating this system.
8.3 Plant Modifications Plant modification reviews are performed via hard copy review of the full modification package rather than the method described in the initial Simulator Certification Report.
8.4 Simulator Control Functions The initial simulator certification report stated that all simulator control functions would be tested. Verification of simulator control functions is limited to those that are required to perform training and certification testing.
j 9.0 Certification Test Schedule 9.1 Schedule Changes One test originally scheduled for 1996 was performed in 1998:
SCT 4.4 Simulator Operating Limits Three of the tests originally scheduled for 1997 were performed in 1996:
SCT 6.6.1 Normal Power to Low Power Operations SCT 10.1 Main Turbine Stop and Governor Valve Test l
l SCT 10.5 Steam Dump Valves Modulating and Trip Tcst, and Atmospheric Steam Dump Valve Test l
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i PBNP Simulator Four-Year Report Page 8 of 18 l
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9.2 1996 PBNP Certification Tests (as performed)
Test ~
Title
.SCr3.3 Duty Cycle Test S CT 6.1.4 Steady State Dr.fr Test,100% Power, BOL S CT 6.2.1 100% Power Steady State Performance Test S CT 6.2.2 75% Power Steady State Performance Test S CT 6.2.3 28% Power Steady State Performance Test S 'T 6.3.4 NSSS Mass Balance S T 6.5.1 Manual Reactor Trip SCT 6.5.2 Simultaneous Trip of Both Main Feedwater Pumps SCT 6.5.3 Simultaneous Closure of All Main Steam Isolation Valves SCT 6.5.4 Simultaneous Trip of All Reactor Coolant Pumps SCT 6.5.5 Trip of Any Single Reactor Coolant Pump SCT 6.5.6 Turbine Trip Below P-9 SCT 6.5.7 Maximum Rate Power Ramp 100% to 75% to 100%
SCT 6.5.8 LOCA With less of Offsite Power SCT 6.5.9 Maximum Unisolable Main Steam Line Break SCT 6.5.10 PZR PORY Stuck Open Without High Head SI
- SCT 6.6.1 Normal Power to Low Power Operations SCT 6.6.2 Reactor Shutdown SCT 6.6.3 Hot Shutdown to Cold Shutdown SCT 6.6.3.1 Plant Cooldown, OP-5A, Part C (CVC)
SCT 6.6.3.2 Placing Residual Heat Removal System in Operation SCT 6.6.6 Plant Start-Up Cold to Hot Standby SCT 6.6.7 -
Nuclear Startup From Hot Standby to Rated Power SCT 6.6.8 Secondary Systems Startup SCT 6.8.3.2 less of Component Cooling Water System SCT 6.8.5.6 Hotwelllevel Control Failure SCT 6.8.8.3 Drifting Rod Group SCT 6.8.8.8 UncoupLd Rod SCT 6.8.11.1 Diesel Generator Failure to Start SCr 6.8.12.1 Load Reference Channel Fails SCT 6.8.13.7 Imss of 120 Volt AC Instrument Bus SCT 6.8.29.3 Steam Generator Tube Rupture SCT 6.8.29.6 RCS Cold Irg Temperature Transmitter Failure SCT 6.8.29.8 Pressurizer Safety Valve Failure SCT 6.8.33.1 Fuel Element Failure SCT 6.8.37.2 Main Steam Line Break Outside Containment SCT 6.8.37.3 Steam Generator Safety Valve Failure SCT 6.8.39.3 SI Pump Failure PBNP Simulator Four-Year Report Page 9 of 18
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1996 PBNP Certification Tests (as performed) (cont) l Test Title SCT 7.1 less of All AC Power (Station Blackout)
SCT 7.2 Loss of All Feedwater SCT 7.6 ATWS Initiated From a Imss of Main Feedwater SCT 10.1 Main Turbine Stop and Governor Valve Test SCT 10.4 Degraded RHR System Capability l
SCT 10.5 Steam Dump Valves Modulating & Trip Test and Atmospheric Steam Dump Valve Test l
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l PBNP Simulator Four-Year Report Page 10 of 18
9.3 1997 PBNP Certification Tests (as performed)
Test Title SCT 3.3 Duty Cycle Test SCT 4.4 Simulator Operating Limits Test SCT 6.1.4 Steady State Drift Test,100% Power, BOL SCT 6.2.1 100% Power Steady State Performance Test SCT 6.2.2 75% Power Steady State Performance Test SCT 6.2.3 28% Power Steady State Performance Test SCT 6.3.5 BOP Mass Balance SCT 6.5.1 Manual Reactor Trip SCT 6.5.2 Simultaneous Trip of Both Main Feedwater Pumps SCT 6.5.3 Simultaneous Closure of All Main Steam Isolation Valves SCT 6.5.4 Simultaneous Trip of All Reactor Coolant Pumps SCT 6.5.5 Trip of Any Single Reactor Coolant Pump SCT 6.5.6 Turbine Trip Below P-9 SCT 6.5.7 Maximum Rate Power Ramp 100% to 75% to 100%
SCT 6.5.8 LOCA With Loss of Offsite Power SCT 6.5.9 Maximum Unisolable Main Steam Line Break SCT 6.5.10 Pressurizer PORV Stuck Open Without High Head SI l
SCT 6.8.3.3 Thermal Barrier Heat Exchanger Leak SCT 6.8.8.2 Dropped Rod SCT 6.8.11.2 Diesel GeneratorInadvertent Trip SCT 6.8.12.3 Inadvertent Turbine Trip SCT 6.8.23.4 Power Range Channel Summing and Level Amp Failure SCT 6.8.23.5 Power Range Detector Failure SCT 6.8.25.12 Pressurizer Pressure Controller Failure SCT 6.8.26.1 Reactor Trip / Bypass Breaker Failure SCT 6.8.29.9 Pmssurizer PORV Failure SCT 6.8.30.3 RHR Pump Fails to Stan on SI SCT 6.8.41.1 CCW to Service Waterleak l
PBNP Simulator Four-Year Report Page 11 of 18
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e 9.4 1998 PBNP Certification Tests (as performed)
Test Title SCF 3.3 Duty Cycle Test SCT 4.4 Simulator Operating Limits SCT 6.1.4 Steady State Drift Test,100% Power, BOL SCT 6.2.1 100% Power Steady State Performance Test SCr 6.2.2 75% Power Steady State Performance Test SCT 6.2.3 28% Power Steady State Performance Test
- SCT 6.3.2 75% Power Heat Balance SCr6.5.1 Manual ReactorTrip SCT 6.5.2 Simultaneous Trip of Both Main Feedwater Pumps SCT 6.5.3 Simultaneous Closure of All Main Steam Isolation Valves SCT 6.5.4 Simultaneous Trip of All Reactor Coolant Pumps SCT 6.5.5 Trip of Any Single Reactor Coolant Pump SCF 6.5.6 Turbine Trip Below P-9 SCT 6.5.7 Maximum Rate Power Ramp 100% to 75% to 100%
SCT 6.5.8 LOCA With Loss of Offsite Power SCT 6.5.9 Maximum Unisolable Main Steam Line Break SCT 6.5.10 Pn:ssurizer PORV Stuck Open Without High Head SI SCr6.6.6.2-Removing RHR System From Service SCF6.6.6.3 Reactor Startup SCT 6.8.2.1 Compressed Air System Heeder Break SCT 6.8.5.1 Main Feedpump Discharge Line Break SCF 6.8.5.2 Feedline Break Inside Containment SCT 6.8.5.3 IAss of Condenser Vacuum SCT 6.8.5.4 Main Feedwater Pump Trip SCT 6.8.8.1 Stuck Rod SCT6.8.8.4 Improper Bank Overlap SCT 6.8.9.2 Letdown Line 1.cak Outside Containment SCT 6.8.13.3 loss of 4160 Volt Bus SCT 6.8.16.3 Generator Trip SCr 6.8.25.3 Tref Program Failure SCF 6.8.25.8 _
Steam Dump Load Reject Control Failure SCT 6.8.25.15 Tavg Bistable Failure SCF 6.8.29.1 RTD Bypass Line Failure SCT 6.8.37.5 Stuck Open Condenser Dump Valve SCF 6.8.41.2 Service Water Pump Failure SCT 8.1 AFW System Check Valves and Flow Indications SCT 8.2 ORT 6: Containment Spray Service Test SCT 10.2 Natural Circulation Cooldown PBNP Simulator Four-Year Report Page 12 of 18
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' 9.5 1999 PBNP Certification Test Schedule Test Title SCT 3..
Duty Cycle Test SCr 6.1.4 Steady State Drift Test,100% Power, BOL SCr 6.2.1 100% Power Steady State Performance Test SCr 6.2.2 75% Power Steady State Performance Test l
SCT 6.2.3 28% Power Steady State Performance Test l
SCT 6.3.1 100% Power Heat Balance SCT 6.3.3 28% Power Heat Balance SCT 6.3.4 NSSS Mass Balance SCr 6.3.5 BOP Mass Balance SCT 6.5.1 Manual ReactorTrip l
SCT 6.5.2 Simultaneous Trip of Both Main Feedwater Pumps SCT 6.5.3 Simultaneous Closure of All Main Steam Isolation Valves SCT 6.5.4 Simultaneous Trip of All Reactor Coolant Pumps SCT 6.5.5 Trip of Any Single Reactor Coolant Pump SCT 6.5.6 Turbine Trip Below P-9 l
SCT 6.5.7 Maximum Rate Power Ramp 100% to 75% to 100%
SCT 6.5.8 LOCA with Loss of Offsite Power SCT 6.5.9 Maximum Unisolable Main Steam Line Break SCT 6.5.10 Pressurizer PORV Stuck Open Without High Head SI SCr 6.6.1 Normal Power to Low Power SCr 6.6.2 Reactor Shutdown l
SCT 6.6.3 Hot Shutdown to Cold Shutdown L
SCT 6.6.3.2 Placing RHR in Operation SCr 52.6
' Cold Shutdown to Hot Shutdown SCT 6.6.6.2 Removing RHR System From Service SCT 6.6.6.3 Reactor Startup SCr6.6.7 low Power to Normal Power Operations SCT 6.6.8
. Secondary Systems Startup l
SCT 6.8.2.2 Instrument Air Compressor Trip SCr6.8.3.2 Standby CCW Pump Start Failure SCr6.8.3.3 Thermal Barrier Heat Exchanger Leak SCT 6.8.5.2 Feedline BreakInside Containment SCT 6.8.5.5 Feedwater Flow Transmitter Failure SCT 6.8.5.6 Hotwelllevel Control Failure SCT 6.8.8.2 Dropped Rod SCT 6.8.8.3 Ratcheting Rod Group SCT 6.8.8.4 Improper Bank Overlap SCT 6.8.8.5 legic Cabinet Urgent Failure SCT 6.8.8.8 Uncoupled Rod SCT 6.8.11.1 Diesel Generator Failure to Start PBNP Siraulator Four-Year Report Page 13 of 18
i 1999 PBNP Certification Test Schedule (cont)
Test Title SCT 6.8.11.2 Diesel Failure Inadvenent Trip SCT 6.8.12.1 Imad Reference Channel Fails SCT 6.8.12.3 Inadvenent Turbine Trip SCT 6.8.13.3 Loss of 4160 Volt Bus SCT 6.8.13.6 loss of 125 Volt DC Bus SG 6.8.13.7 Loss of 120 Volt AC Bus SCT 6.8.16.3 Generator Trip SCT 6.8.23.4 PR Channel Summing & level Amp Failure SCT 6.8.23.5 Power Range Detector Failure SCT 6.8.25.-2 Back Up Heater Reduced Capacity SCT 6.8.25.3 Tref Program Failure SCT 6.8.25.12 Pressurizer Pressure Control Failure SCT 6.8.25.15 Tavg Bistable Failure SG 6.8.26.1 Reactor Trip / Bypass Breaker Failure SCT 6.8.29.1 RTD Bypass Line Failure SCT 6.8.29.2 DBA LOCA SCT 6.8.29.3 Steam GeneratorTube Rupture SCT 6.8.29.6 RCS Cold leg Temperature Transmitter Failure SCT 6.8.29.8 Pressurizer Safety Valve Failure SCT 6.8.29.9 Pressurizer PORV Failure SCT 6.8.30.2 RHR Pump Trip SCT 6.8.30.3 RHR Pump Fails to Start on Safety InjecCon Actuation SCT 6.8.33.1 Fuel Element Failure SG 6.8.37.1 Main Steam Line Break Inside Containment SCT 6.8.37.2 Main Steam Line Break Outside Containment
. SCT 6.8.37.3 Steam Generator Safety Valve Failure SCT 6.8.37.5 Stuck Open Condenser Dump Valve SCT 6.8.39.3 SafetyInjection Pump Failure SCT 6.8.41.1 CCW to Service WaterIrak SCT 7.1 Loss of All AC Power (Station Blackout)
SCT 7.2 Imss of All Feedwater SCT 7.6 ATWS Initiated From a Imss of MFW SCT 8.2 ORT-6: Containment Spray Service Test SCT 10.1 TS-3 Main Turbine Stop/ Governor Valve Test SG 10.3 Reactor Coolant Pump Operation SCT 10.4 Degraded RHR System Capability SCT 10.5 Steam Dump Valve & Atmospheric Dump Valve Test PBNP Simulator Four-Year Report Page 14 of 18 1
9.6 2000 PBNP Certification Test Schedule Test Title SCT 3.3 Duty Cycle Test SCr 4.4 Simulator Operating Limits Test SCT 6.1.4 Steady State Drift Test,100% Power, BOL SCT 6.2.1 100% Power Steady State Performance Test SCT 6.2.2 75% Power Steady State Performance Test SCT 6.2.3 28% Power Steady State Performance Test SCT 6.3.4 NSSS Mass Balance SCT 6.5.1 Manual Reactor Trip SCF 6.5.2 Simultaneous Trip of Both Main Feedwater Pumps SCT 6.5.3 Simultaneous Closure of All Main Steam Isolation Valves SCT 6.5.4 Simultaneous Trip of All Reactor Coolant Pumps SCF 6.5.5 Trip of Any Single Reactor Coolant Pump SCr 6.5.6 Turbine Trip Below P-9 SCT 6.5.7 Maximum Rate Power Ramp 100% to 75% to 10%
SCr 6.5.8 LOCA With Loss of Offsite Power SCF 6.5.9 Maximum Unisolable Main Steam Line Break SCT 6.5.10 Pmssurizer PORV Stuck Open Without High Head SI SCr 6.8.3.2 Loss of Component Cooling Water ?ystem SCr 6.8.5.6 HotwellI.evel Control Failure SCT 6.8.8.3 Ratcheting Rod Group SCr 6.8.12.1 Load Reference Channel Fails SCT 6.8.13.7 Imss of 120 Volt AC Instrument Bus SCT 6.8.29.3 Steam Generator Tube Rupture SCT 6.8.29.6 RCS Cold I.eg Temperature Transmitter Failure SCF 6.8.29.8 Pressurizer Safety Valve Failure SCT 6.8.33.1 Fuel Element Failure SCT 6.8.37.2 Main Steam Line Break Outside Containment SCT 6.8.37.3 Steam Generator Safety Valve Failure SCr 6.8.39.3 SI Pump Failure i
SCT 7.1 Imss of All AC Power (Station Blackout)
SCT 7.2 Loss of All Feedwater SCT 7.6 ATWS Initiated From a Loss of Main Feedwater SCT 10.4 Degraded RHR System Capability i
PBNP Simulator Four-Year Report Page 15 of 18
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9.7 2001 PBNP Certification Test Schedule Test Title SCT 3.3 Duty Cycle Test SCr 6.1.4 Steady State Drift Test,100% Power, BOL SCT 6.2.1 100% Power Steady State Performance Test SCr 6.2.2 '
75% Power Steady State Performance Test SCT 6.2.3 28% Power Steady State Performance Test SCT 6.3.5 BOP Mass Balance I
SCF 6.5.1 Manual Reactor Trip SCT 6.5.2 Simultaneous Trip of Both Main Feedwater Pumps SCT 6.5.3 Simultaneous Closure of All Main Steam Isolation Valves i
SCF 6.5.4 Simultaneous Trip of All Reactor Coolant Pumps SCT 6.5.5 Trip of Any Single Reactor Coolant Pump SCr 6.5.6 Turbine Trip Below P-9 SCr6.5.7 Maximum Rate Power Ramp 100% to 75% to 100%
SCT 6.5.8 LOCA With I.oss of Offsite Power
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SCT 6.5.9 Maximum Unisolable Main Steam Line Break SCT 6.5.10
~ Pressurizer PORV Stuck Open Without High Head SI SCT 6.6.1 Normal Power to Low Power Operations SCT 6.6.2 Reactor Shutdown SCT 6.6.3 Hot Shutdown to Cold Shutdown SCr6.6.3.2 Placing RHR in Operation SCT 6.6.6 Cold Shutdown to Hot Standby SCT 6.6.6.2 Removing RHR System From Operation SCT 6.6.6.3 Reactor Startup SCT 6.6.7 Low Power to Normal Power Operations SCT 6.6.8 Secondary Systems Startup SCT 6.8.3.3 Thermal Barrier Heat Exchanger Leak SCF 6.8.8.2 Dropped Rod SCF 6.8.11.2 Diesel GeneratorInadvertent Trip SCT 6.8.23.4 Power Range Channel Summing and level Amp Failure SCT 6.8.23.5 Power Range Detector Failure SCr 6.8.25.12 Pressurizer Pressure Controller Failure SCF 6.8.26.1 Reactor Trip / Bypass Breaker Failure SCF 6.8.29.1 RTD Bypass Line Failure SCr 6.8.29.9 Pressurizer PORV Failure SCT 6.8.30.3 RHR Pump Fails to Start on SI SCT 6.8.41.1 CCW to Service Waterleak SCT 10.1 Main Turbine Stop and Governor Valve Test PBN.P Simulator Four-Year Report Page 16 of 18
l 9.8 2002 PBNP Certification Test Schedule Test Title SCT 3.3 Duty Cycle Test SCT 6.1.4 Steady State Drift Test,100% Power, BOL SCT 6.2.1 100% Power Steady State Performance Test SCT 6.2.2 75% Power Steady State Performance Test SCT 6.2.3 28% Power Steady State Performance Test SCT 6.3.2 75% Power Heat Balance SCr6.5.1 Manual Reactor Trip SCT 6.5.2 Simultaneous Trip of Both Main Feedwater Pumps SCr6.5.3 Simultaneous Closure of All Main Steam Isolation Valves SCT 6.5.4 Simultaneous Trip of All Reactor Coolant Pumps SCT 6.5.5 Trip of Any Single Reactor Coolant Pump SCT 6.5.6 Turbine Trip Below P-9 SCT 6.5.7 Maximum Rate Power Ramp 100% to 75% to 100%
SCT 6.5.8 LOCA With Loss of Offsite Power l
SCT 6.5.9 Maximum Unisolable Main Steam Line Break l
SCT 6.5.10 Pressurizer PORV Stuck Open Without High Head SI l
SCT 6.8.2.1 Compressed Air System Header Break SCT 6.8.5.1 Main Feedpump Discharge Line Break SCT 6.8.5.2 Feedline Break Inside Containment SCT 6.8.5.3 Loss of Condenser Vacuum SCT 6.8.5.4 Main Feedwater Pump Trip SG 6.8.8.1 Stuck Rod SCT 6.8.8.4 Improper Bank Overlap SCT 6.8.9.2 Ixtdown Line Izak Outside Containment SCT 6.8.13.3 Less of 4160 Volt Bus SCF 6.8.16.3 Generator Trip SCT 6.8.25.3 Tref Program Failure SCT 6.8.25.8 Steam Dump Imad Rejection Control Failure SCT 6.8.25.15 Tavg Bistable Failure SCT 6.8.37.5 Stuck Open Condenser Dump Valve SCr 6.8.41.2 Service Water Pump Failure SCT 8.1 AFW System Check Valves and Flow Indications SCT 8.2 ORT 6: Containment Spray Service Test SCT 10.2 Natural Circulation Cooldown PBNP Simulator Four-Year Report Page 17 of 18
9.9 2003 PBNP Certification Test Schedule f
Test Title SCr 3.3 Duty Cycle Test SCT 6.1.4 Steady State Drift Test,100% Power, BOL SCF 6.2.1 100% Power Steady State Performance Test SCT 6.2.2 75% Power Steady State Performance Test j
SCT 6.2.3 28% Power Steady State Performance Test SCT 6.3.1
- 100% Power Heat Balance SCT 6.3.3 28% Power Heat Balance SCT 6.5.1 Manual Reactor Trip SCT 6.5.2 Simultaneous Trip of Both Main Feedwater Pumps SCT 6.5.3 Simultaneous Closure of All Main Steam Isolation Valves SCr6.5.4 Simultaneous Trip of All Reactor Coolant Pumps SCT 6.5.5 Trip of Any Single Reactor Coolant Pump SCT 6.5.6 Turbine Trip Below P-9 SCT 6.5.7 Maximum Rate Power Ramp 100% to 75% to 100%
SCT 6.5.8 LOCA With loss of Offsite Power SCr6.5.9 Maximum Unisolable Main Steam Line Break SCF 6.5.10 Pressurizer PORV Stuck Open Without High Head SI SCT 6.8.2.2 Instmment Air Compressor Trip SCT 6.8.5.5 Feedwater Flow Transmitter Failure SCT 6.8.8.5 Logic Cabinet Urgent Failure SCT 6.8.8.8 Uncoupled Rod SCT 6.8.11.1 Diesel Generator Failure to Start SCF 6.8.12.3 Inadvertent Turbine Trip SCT 6.8.13.6 Loss of 125 Volt DC Bus SCT 6.8.25.2 Back Up Heater Reduced Capacity SCT 6.8.29.2 DBA LOCA SCT 6.8.30.2 RHR Pump Trip l
SCT 6.8.37.1 Main Steam Line Break Inside Containment SCT 10.3 Reactor Coolant Pump Operation SCF 10.5 Steam Dump Valves Modulating and Trip Test and Atmospheric Steam Dump Valve Test PBNP Simulator Four-Year Report Page 18 of 18
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Holtec Center,555 Lincoln Drive West, Marlton, NJ 08053 HOLTEC Telephone (609) 797-0900 Fax (609) 797-0909 INTERNATIONAL July 16,1999 Carl Paperiello, Ph.D.
Director Office of Nuclear Material Safety and Safeguards U. S. Nuclear Regulatory Commission Washington, DC 20555-0001
Subject:
USNRC Docket No. 71-9261 Ill-STAR 100 Certificate of Compliance No. 71-9261 Renewal ofIIoltec Quality Assurance Program Approval
Reference:
IIoltec Project 5014
Dear Dr. Paperiello:
The purpose of this correspondence is to request renewal of the NRC's approval of IIoltec International's Quality Assurance Program in accordance with the provisions of 10 CFR 71.38. The current NRC approval of 11oltec's OA Program expires on August 31,1999. Enclosed is one uncontrolled copy of the current IIoltec Quality Assurance Manual for your review.
If you have any questions or require additional information, please contact us.
S'
- crely, Mark Soler Acting Quality Assurance Manager Cc:
Mr. E. William Brach (w/ encl.)
Document ID: 5014331
Enclosure:
IIoltec International Quality Assurance Manual, Revision 11, dated February 1,1999.
Approval:
Yu W
GT7Ai IcA A
- N.
hrian Gutilerman, PE K. P. Singh, Ph.D., PE Licensing Manager President and CEO 9907260022 990716 PDR ADOCK 071007s4 C
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Holtec Center,555 Linoin Drive West, Marlton, NJ 08053 HOLTEC Telephone (609) 797-0900 Fax (609) 797-0909 INTERNATIONAL Dr. Carl Paperiello U. S. Nuclear Regulatory Commission Document ID 5014331 Page 2 of 2 Client Distribution (w/o encl.h Recinien.lt Utility Mr. David Bland Southern Nuclear (if UG Chairman)
Mr. J. Nathan Leech Commonwealth Edison Mr. Bruce Patton Pacific Gas & Electric Co. - Diablo Canyon Dr. Max DeLong Private Fuel Storage, LLC Mr. Rodney Pickard American Electric Power Mr. Ken Phy New York Power Authority Mr. David Larkin Washington Public Power Supply System Mr. Eric Meils Wisconsin E'ectric Power Company Mr. Paul Plante Maine Yankee Atomic Power Company Mr. Stan Miller Vermont Yankee Corporation Mr. Jim Clark Southern California Edison - SONGS Mr. Ray Kellar Entergy Operations - Arkansas Nuclear One Mr. Joe Andrescavage GPUN - Oyster Creek Nuclear Power Station Mr. Ron Bowker IES Utilities Mr. William Swantz Nebraska Public Power District I
Mr. Mark G. Smith Pacific Gas & Electric - Ilumboldt Bay