ML20086N317
| ML20086N317 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 07/19/1995 |
| From: | WISCONSIN ELECTRIC POWER CO. |
| To: | |
| Shared Package | |
| ML20086N103 | List: |
| References | |
| NUDOCS 9507250171 | |
| Download: ML20086N317 (13) | |
Text
,- 4 l
l
.. l WISCONSIW ELECTRIC POWER COMPANY POINT BEACH NUCLEAR PLANT f
SIMULATOR FOUR-YEAR REPORT Contents 1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . 1 2 Simulator Information . . . . . . . . . . . . . . . . . . . .- 2 3 Completed Certification Tests . . . . . . . . . . . . . . . . 3 4 Certification Test Failure . . .'. . . . . . .. . . . . . . 8 5 Simulator Modifications . . . . . . . . . . . . . . . . . . . 8 6 Exceptions to ANSI /ANS 3.5 as endorsed by Reg Guide 1.149 . . 8 7 1996-1999 Certification Test Schedule . . . . . .. . . . . . 9 1 Introduction This Four-Year Report is provided on the fourth anniversary of initial certification in accordance with 10 CFR 55.45(b) (5) (ii) .
The report describes Simulator Certification Tests performed from 1992 to 1995, discusses test failures from 1992 to 1995 and :
provides a schedule of tests to be performed in the next four year '
period.
t PBNP Simulation Facility Four-Year Report Page 1 of 13 i
l 9507250171 950719 PDR ADOCK 05000266 P PDR
ra 2 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 [10 CFR 55.45 (b) (5) (ii) ]
k t
t i
I PBNP Simulation Facility Four-Year Report Page 2 of 13 l
l
3 Completed Certification Tests The annual and quadrennial certification tests were completed as scheduled in the initial certification report except as noted.
The test results were presented to and approved by the Simulator Review Committee each year.
3.1 Test Frequency Increase The certification report listed these tests for quadrennial performance. These tests should be and were performed annually.
14.6.1.4 Steady State Drift Test, 100% Power, BOL 14.6.2.1 100% Power Steady State Performance Test 14.6.2.2 75% Power Steady State Performance Test 14.6.2.3 28% Power Steady State Performance Test 3.2 Test Frequency Decrease The certification report listed this test for annual performance. This testing was performed in 1992 and 1993 but will also be performed on a quadrennial basis.
14.4.4 Simulator Operating Limits Test 3.3 Test Scope certification tests, which are performed on generic equipment such as an instrument bus failure, will be rotated each test interval. For example an instrument bus failure test will be rotated in each test interval to verify proper response to the failure of a different instrument bus. Many certification tests were performed on Unit 2 simulator as well. In addition to the certification test procedure requirements, a verification of recoverability from the failure and operator intervention to mitigate the failure is performed where appropriate.
l PBNP Simulation Facility Four-Year Report Page 3 of 13 l
l l
e- - _ -- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _
', .. l l
3.4 Completed 1992 PBNP Certification Tests Test # Title 14.3.3 Duty Cycle Test i 14.4.4 Simulator Operating Limits Test '
14.6.1.4 Steady State Drift Test, 100% Power, BOL 14.6.2.1 100% Power Steady State Performance Test 14.6.2.2 75% Power Steady State Performance Test 14.6.2.3 28% Power Steady State Performance Test' '
14.6.3.4 NSSS Mass Balance -i 14.6.5.1 Manual Reactor Trip 14.6.5.2 Simultaneous Trip of Both Main Feedwater Pumps 14.6.5.3 Simultaneous Closure of All Main Steam Isolation Valves 14.6.5.4 Simultaneous Trip of.All Reactor Coolant Pumps 14.6.5.5 Trip of Any Single Reactor Coolant Pump 14.6.5.6 Turbine Trip Below P-9 14.6.5.7 Maximum Rate Power Ramp 100% to 75% to 100% .
14.6.5.8 LOCA with Loss of Offsite Power 14.6.5.9 Maximum Unisolable Main Steam Line Break 14.6.5.10 Pressurizer PORV Stuck Open Without High Head SI 14.6.5.12 Loss of Unit 1 Red Instrument Bus LER 91-005-00 14.6.5.13 Unit 1 Load Rejection Transient LER 90-010-01 14.6.5.14 Unit 2 Loss of Condensate Flow LER 90-005-00 '
14.6.6.6 Plant Start-Up Cold to Hot Standby 14.6.6.7 Nuclear Startup from Hot Standby to Rated Power 14.6.8.3.2 Loss of Component Cooling Water System :
14.6.8.5.6 Hotwell level control failure 14.6.8.8.3 Drifting Rod Group 14.6.8.13.7 Loss of 120 Volt AC Instrument Bus 14.6.8.29.3 Steam Generator Tube Rupture 14.6.8.29.6 RCS Cold Leg Temperature Transmitter Failure 14.6.8.29.8 Pressurizer Safety Valve Failure 14.6.8.33.1 Fuel Element Failure 14.6.8.37.2 Main Steam Line Break Outside Containment 14.6.8.37.3 Steam Generator Safety Valve Failure 14.6.8.39.3 SI Pump Failure i 14.7.1 Loss of All AC Power (Station Blackout) l 14.7.2 Loss of All Feedwater l 14.7.6 ATWS Initiated from a Loss of Main Feedwater )
14.10.4 SEP-1, Degraded RHR System Capability PBNP Simulation Facility Four-Year Report Page 4 of 13
~, -. _ , _.
4 3.5 Completed 1993 PBNP Certification Tests Test # Title 14.3.3 Duty Cycle Test 14.4.4 Simulator Operating Limits Test 14.6.1.4 Steady State Drift Test, 100% Power, BOL 14.6.2.1 100% Power Steady State Performance Test 14.6.2.2 75% Power Steady State Performance Test 14.6.2.3 28% Power Steady State Performance Test 14.6.3.5 BOP Mass Balance 14.6.5.1 Manual R.eactor Trip 14.6.5.2 Simultaneous Trip of Both Main Feedwater Pumps 14.6.5.3 Simultaneous Closure of All Main Steam Isolation Valves 14.6.5.4 Simultaneous Trip of All Reactor Coolant Pumps 14.6.5.5 Trip of Any Single Reactor Coolant Pump 14.6.5.6 Turbine Trip Below P-9 14.6.5.7 Maximum Rate Power Ramp 100% to 75% to 100%
14.6.5.8 LOCA with Loss of Offsite Power 14.6.5.9 Maximum Unisolable Main Steam Line Break 14.6.5.10 Pressurizor PORV Stuck Open Without High Head SI 14.6.6.1 Normal Pohnr to Low Power Operations, OP-3A 14.6.6.8 Secondary Dystems Startup, OP-13A 14.6.8.3.3 Thermal Bat'rier Heat Exchanger Leak 14.6.8.8.2 Dropped Rol 14.6.8.11.2 Diesel Genarator Inadvertent Trip 14.6.8.12.3 Inadvertent Turbine Trip 14.6.8.23.4 Power Rango Channel Summing and Level Amp Failure 14.6.8.23.5 Power Rango Detector Failure 14.6.8.25.12 Pressurizer Pressure Controller Failure 14.6.8.26.1 Reactor Tr3p / Bypass Breaker Failure 14.6.8.29.9 Pressurizer PORV Failure 14.6.8.30.3 RHR Pump Fails to Start on SI 14.6.8.41.1 CCW To Service Water Leak 14.10.1 Main Turbine Stop and Governor Valve Test, TS-3 Sect 1-4 14.10.5 OI-23, Stealn Dump Valves Modulating and Trip Test and Atmospheric Steam Dump Valve Test PBNP Simulation Facility Four-Year Report Page 5 of 13
e 3.6' Completed 1994 PBNP Certification Tests
~
Test # Title 14.3.3 Duty Cycle Test 14.6.1.4 Steady State Drift Test, 100% Power, BOL 14.6.2.1 100% Power Steady State Performance Test 14.6.2.2 75% F0wer Steady State Performance Test i 14.6.2.3 28% Power Steady Sta,te Performance Test i 14.6.3.2 75% Power Heat Balance 14.6.5.1 Manual Reactor Trip 14.6.5.2 Simultaneous Trip of Both Main Feedwater Pumps 14.6.5.3 Simultaneous Closure of All Main Steam Isolation Valves 14.6.5.4 T multaneous Trip of All Reactor Coolant Pumps 14.6.5.5 Trip of Any Single Reactor Coolant Pump 14.6.5.6 Turbine Trip-Below P-9 14.6.5.7 Maximum Rate Power Ramp 100% to 75% to 100%
14.6.5.8 LOCA with Loss of Offsite Power i 14.6.5.9 Maximum Unisolable Main Steam Line Break ,
14.6.5.10 Pressurizer PORV Stuck Open Without High Head SI l' 14.6.8.2.1 Compressed Air System Header Break 14.6.8.5.1 Main Feedpump Discharge Line Break ;
14.6.8.5.2 Feedline Break Inside Containment i 14.6.8.5.4 Main Feedwater Pump Trip I 14.6.8.8.1 Stuck Rod 14.6.8.8.4 Improper Bank Overlap i 14.6.8.9.2 Letdown Line Leak Outside Containment i 14.6.8.13.3 Loss of 4160 Volt Bus i 14.6.8.16.3 Generator Trip 14.6.8.25.3 Tref Program Failure 14.6.8.25.15 Tavg Bistable Failur~e ,
14.6.8.37.5 Stuck Open Condenser Dump Valve 14.6.8.41.2 Service Water Pump Failure 14.8.1 IT-290: AFW System Check Valves and Flow Indications 14.8.2 ORT 6: Containment Spray Service Test '
14.10.2 EOP-02: Natural Circulation Cooldown ,
i f
i 1
l P
i PBNP Simulation Facility Four-Year Report Page 6 of 13
. 1 4 .
j i
3 7 Completed 1995 PBNP Certification Tests
~
Test # Title 14.3.3 Duty Cycle Test ]
14.6.1.4 Steady State Drift Test, 100% Power, BOL i 14.6.2.1 100% Power Steady State Performance Test !
14.6.2.2 75% Power Steady State Performance Test i 14.6.2.3 28% Power Steady State Performance Test l 14.6.3.1 100% Power Heat Balance ;
14.6.3.3 28% Power Heat Balance 14.6.4.1.3 Rod Worth Test 14.6.4.1.5 Isothermal Temperature Coefficient 14.6.5.1 Manual Reactor Trip 14.6.5.2 Simultaneous Trip of Both Main Feedwater Pumps ,
14.6.5.3 Simultaneous Closure of All Main Steam Isolation Valves 14.6.5.4 Simultaneous Trip of All Reactor Coolant Pumps 14.6.5.5 Trip of Any Single Reactor Coolant Pump '
14.6.5.6 Turbine Trip Below P-9 14.6.5.7 Maximum Rate Power Ramp 100% to 75% to 100% "
14.6.5.8 LOCA with Loss of Offsite Power 14.6.5.9 Maximum Unisolable Main Steam Line Break 14.6.5.10 Pressurizer PORV Stuck Open Without High Head SI 14.6.6.2 Reactor Shutdown - OP-3B i 14.6.6.3 Hot Shutdown to Cold Shutdown - OP-3C 14.6.6.3.1 Plant Cooldown - OP-5A Part C (CVC) ;
14.6.6.3.2 Placing RHR in Operation, OP-7A 14.6.8.2.2 Instrument Air Compressor Trip 14.6.8.5.5 Feedwater Flow Transmitter Failure i 14.6.8.8.5 Logic Cabinet Urgent Failure l 14.6.8.11.2 Diesel Failure Inadvertent Trip 1 14.6.8.13.6 Loss of 125 Volt DC Buss i 14.6.8.25.2 Back Up Heater Reduced Capacity 14.6.8.29.2 DBA LOCA 14.6.8.30.2 RHR Pump Trip 14.6.8.37.1 Main Steam Line Break Inside Containment 14.10.3 Reactor Coolant Pump Operation i
l l
PBNP Simulation Facility Four-Year Report Page 7 of 13
]
1
. l
+
4 Certification Test Failure .
At'the current time, there are no uncorrected certification (i.e performance) test failures. During the four year test period, the following test was dispositioned as a failure. The simulator performance has been corrected.
14.6.5.6 Turbine Trip Below P-9 B2.2(6) Main turbine trip (maximum power level which does not result in immediate reactor trip)
The reason the test failed was a reactor trip occurred within two minutes of the turbine trip. The current training simulator software was tested during the spring 1995 simulator outage when the test was performed successfully. Reactor power was reduced to less than 2% without operator action.
Reactor and secondary systems were stable in automatic control after 10 minutes.
5 Simulator Modifications The Simulator Review Committee esta'blishes the priority for modifications. A large number of modifications have occurred which range from annunciator setpoint changes to addition of Emergency Diesel Generators. Whenever modifications are made to the simulator both the Unit 1 and Unit 2 simulation are tested.
Discussed below are two major modifications.
Two additional Emergency Diesel Generators are being installed in the reference plant. The Committee requested simulator installation prior to plant implementation so that operator training would occur prior to plant operations with the new equipment.
The simulator was modified and tested just after delivery for midloop operations. Failures can be entered to fail instrumentation, PER pumps or introduce leaks. The simulator provides feedback expected from the plant for operator intervention.
6 Exceptions to ANSI /ANS 3.5 as endorsed by Reg Guide 1.149 None PBNP Simulation Facility Four-Year Report Page 8 of 13
7 1996-1999 Certification Test Schedule The' new certification test schedule is identical to the schedule provided in the original certification submittal except as noted l below.
The two core physics tests were deleted from the quadrennial
[ tests since the entire core physics are tested whenever the software is upgraded to the current fuel cycle. j 14.6.4.1.3 Rod Worth Test I 14.6.4.1.5 Isothermal Temperature Coefficient I The certification report listed these tests for quadrennial performance which should be performed annually.
14.6.1.4 Steady State Drift Test, 100% Power, BOL 14.6.2.1 100% Power Steady State Performance Test i 14.6.2.2 75% Power Steady State Performance Test 14.6.2.3 28% Power Steady State Performance Test The certification report listed this test for annual performance but will be performed quadrennial.
14.4.4 Simulator Operating Limits Test These transient tests were performed in the initial four year test period. Due to changes in plant configuration these will be deleted.
14.6.5.13 Unit #1 Load Rejection Transient LER 90-010-01 14.6.5.14 Unit 2 Loss of Condensate Flow LER 90-005-00 The test performance date for 14.6.6.8 is 1996 instead of 1997 since it is performed as part of procedue 14.6.6.7, " Nuclear Startup from Hot Standby to Rated Power" which is performed in 1996.
14.6.6.8 Secondary Systems Startup, OP-13A PBNP Simulation Facility Four-Year Report Page 9 of 13
- 7.1 1996 PBNP Certification Test Schedule Tes't # Title 14.3.3 Duty Cycle Test ,
14.4.4 Simulator Operating Limits Test
~
14.6.1.4 Steady State Drift Test, 100% Power, BOL 14.6.2.1 100% Power Steady State Performance Test 14.6.2.2 75% Power Steady Ftate Performance Test 14.6.2.3 28% Power Steady State Performance Test 14.6.3.4 NSSS Mass Balance 14.6.5.1 Manual Reactor Trip 14.6.5.2 Simultaneous Trip of Both Main Feedwater Pumps 14.6.5.3 Simultaneous Closure of All Main Steam Isolation Valves .
14.6.5.4 Simultaneous Trip of All Reactor Coolant Pumps 14.6.5.5 Trip of Any Single Reactor Coolant Pump 14.6.5.6 Turbine Trip Below P-9 14.6.5.7 Maximum Rate Power Ramp 100% to 75% to 100% ,
14.6.5.8 LOCA with Loss of Offsite Power !
14.6.5.9 Maximum Unisolable Main Steam Line Break 14.6.5.10 Pressurizer PORV Stuck Open Without High Head SI 14.6.6.6 Plant Start-Up Cold to Hot Standby 14.6.6.7 Nuclear Startup from Hot Standby to Rated Power 14.6.6.8 Seconde y Systems Startup, OP-13A ,
14.6.8.3.2 Loss o; Component Cooling Water System 14.6.8.5.6 Hotwell level control failure 14.6.8.8.3 Drifting Rod Group .
14.6.8.13.7 Loss of 120 Volt AC Instrument Bus 14.6.8.29.3 Steam Generator Tube Rupture 14.6.8.29.6 RCS Cold Leg Temperature Transmitter Failure i 14.6.8.29.8 Pressurizer Safety Valve Failure 14.6.8.33.1 Fuel Element Failure 14.6.8.37.2 Main Steam Line Break Outside Containment 14,6.8.37.3 Steam Generator Safety Valve Failure 14.6.8.39.3 SI Pump Failure ,
14.7.1 Loss of All AC Power (Station Blackout) '
14.7.2 Loss of All Feedwate'r 14.7.6 ATWS Initiated from a Loss of Main Feedwater 14.10.4 AOP-9C, Degraded RHR System Capability I
PBNP Simulation Facility Four-Year Report Page 10 of 13
V, .
+
7.2 1997 PBNP Certification Test Schedule
~
Test / Title 14.3.3 Duty Cycle Test 14.6.1.4 Steady State Drift Test, 100% Power, BOL
~
14.6.2.1 100% Power Steady State Performance Test 14.6.2.2 75% Power Steady State Performance Test 14.6.2.3 28% Power Steady State Performance Test 14.6.3.5 BOP Mass Balance 14.6.5.1 Manual Reactor Trip 14.6.5.2 Simultaneous Trip of Both Main Feedwater Pumps 14.6.5.3 Simultaneous Closure of All Main Steam Isolation Valves 14.6.5.4 Simultaneous Trip of All Reactor Coolant Pumps 14.6.5.5 Trip of Any Single Reactor Coolant Pump 14.6.5.6 Turbine Trip Below P-9 14.6.5.7 Maximum Rate Power Ramp 100% to 75% to 100%
14.6.5.8 LOCA with Loss of Offsite Power 14.6.5.9 Maximum Unisolable Main Steam Line Break 14.6.5.10 Pressurizer PORV Stuck Open Without High Head SI 14.6.6.1 Normal Power to Low Power Operations, OP-3A 14.6.8.3.3 Thermal Barrier Heat Exchanger Leak 14.6.8.8.2 Dropped Rod 14.6.8.11.2 Diesel Generator Inadvertent Trip 14.6.8.12.3 Inadvertent Turbine Trip 14.6.8.23.4 Power Range Channel Summing and Level Amp Failure 14.6.8.23.5 Power Range Detector Failure 14.6.8.25.12 Pressurizer Pressure Controller Failure 14.6.8.26.1 Reactor Trip / Bypass Breaker Failure 14.6.8.29.9 Pressurizer PORV Failure 14.6.8.30.3 RHR Pump Fails to Start on SI 14.6.8.41.1 CCW To Service Water Leak 14.10.1 Main Turbine Stop and Governor Valve Test, TS-3 Sect 1-4 14.10.5 OI-13, Steam Dump Valves Modulating and Trip Test and Atmospheric Steam Dump Valve Test l
l l
PBNP Simulation Facility Four-Year Report Page 11 of 13 1
1
~'
+
7 . ,3 1998 PBNP Certification Test Schedule
- Tes't # Title l 14.3.3 Duty Cycle Test 1 14.6.1.4 Steady State Drift Test, 100% Power, BOL 14.6.2.1 100% Power Steady State Performance Test 1,4.6.2.2 75% Power Steady State Performance Test 14.6.2.3 28% Power Steady State Performance Test 14.6.3.2 75% Power Heat Balance r 14.6.5.1 Manual Reactor Trip 14.6.5.2 Simultaneous Trip of Both Main Feedwater Pumps 14.6.5.3 Simultaneous Closure of All Main Steam Isolation Valves 14.6.5.4 Simultaneous Trip of All Reactor Coolant Pumps ,
14.6.5.5 Trip of Any Single Reactor Coolant Pump ';
14.6.5.6 Turbine Trip Below P-9 14.6.5.7 Maximum Rate Power Ramp 100% to 75% to 100%
14.6.5.8 LOCA with Loss of Offsite Power i I
14.6.5.9 Maximum Unisolable Main Steam Line Break 14.6.5.10 Pressurizer PORV Stuck Open Without High Head SI ;
~
14.6.8.2.1 Compressed Air System Header Break 14.6.8.5.1 Main Feedpump Discharge Line Break ,
14.6.8.5.2 Feedline Break Inside Containment 14.6.8.5.4 Main Feedwater Pump Trip 14.6.8.8.1 Stuck Rod 14.6.8.8.4 Improper Bank Overlap i 14.6.8.9.2 Letdown Line Leak Outside Containment 14.6.8.13.3 Loss of 4160 Volt Bus 14.6.8.16.3 Generator Trip 14.6.8.25.3 Tref Program Failure [
14.6.8.25.15 Tavg Bistable Failure 14.6.8.37.5 Stuck Open Condenser Dump Valve .
14.6.8.41.2 Service Water Pump Failure 14.8.1 IT-290: AFW System Check Valves and Flow Indications -
14.8.2 ORT 6: Containment Spray Service Test >
14.10.2 EOP-02: Natural Circulation Cooldown i
l PBNP Simulation Facility Four-Year Report Page 12 of 13 i
~
a . 1 l
7.,4 1999 PBNP Certification Test schedule Tes't #' Title 14.3.3 Duty Cycle Test 14.6.1.4 Steady State Drift Test, 100% Power, BOL 14.6.2.1 100% Power Steady State Performance Test 14.6.2.2 75% Power Steady State Performance Test l 14.6.2.3 28% Power Steady State Performance Test i 100% Power Heat Balance 14.6.3.1 14.6.3.3 28% Power Heat Balance )
14.6.5.1 Manual Reactor Trip '
14.6.5.2 Simultaneous Trip of Both Main Feedwater Pumps l 14.6.5.3 Simultaneous Closure-of All Main Steam Isolation '
Valves 14.6.5.4 Simultaneous Trip of All Reactor Coolant Pumps .
14.6.5.5 Trip of Any Single Reactor Coolant Pump 14.6.5.6 Turbine Trip Below P-9 14.6.5.7 Maximum Rate Power Ramp 100% to 75% to 100% l 14.6.5.8 LOCA with Loss of Offsite Power 14.6.5.9 Maximum Unisolable Main Steam Line Break 14.6.5.10 Pressurizer PORV Stuck Open Without High Head SI 14.6.6.2 Reactor Shutdown - OP-3B 14.6.6.3 Hot Shutdown to Cold Shutdown - OP-3C 14.6.6.3.1 Plant Cooldown - OP-5A Part C (CVC) 14.6.6.3.2 Placing RHR in Operation, OP-7A ,
14.6.8.2.2 Instrument Air Compressor Trip 14.6.8.5.5 Feedwater Flow Transmitter Failure 14.6.8.8.5 Logic Cabinet Urgent Failure 14.6.8.11.2 Diesel Failure Inadvertent Trip 14.6.8.13.6 Loss of 125 Volt DC Buss 14.6.8.25.2 Back Up Heater Reduced Capacity 14.6.8.29.2 DBA LOCA ,
14.6.8.30.2 RHR Pump Trip 14.6.8.37.1 Main Steam Line Break Inside Containment 14.10.3 Reactor Coolant Pump, Operation l
P i
PBNP Simulation Facility Four-Year Report Page 13 of 13.
.