ML17352A884
| ML17352A884 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 10/28/1994 |
| From: | FLORIDA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML17352A883 | List: |
| References | |
| NUDOCS 9411140078 | |
| Download: ML17352A884 (57) | |
Text
FLORIDA POWER AND LIGHT COMPANY TUIGG<rY POINT UNIT 3 SIMULATOR CERTIFICATION UPDATE NUMBER 1 TABLE OF CONTENTS
1.0 INTRODUCTION
2.0 SIMULATOR TESTS COMPLETED (1991 - 1994) 2.1 1991 TESTS 2.2 1992 TESTS 2.3 1993 TESTS 2.4 1994 TESTS 3.0 CERTIFICATION TEST CHANGES (1995 - 1998) 3.1 CHANGES 32 DELETIONS 3.3 ADDITIONS 4.0 FOUR YEAR TEST PLAN (1995 - 1998) 5.0 OUTSTANDING DISCREPANCIES APPENDIX A TEST ABSTRACTS A.1 TESTS ADDED A.2 TESTS CHANGED 9411140078 941028 PDR ADOCK 05000250
.P PDR
SIMULATOR CERTIFICATION UPDATE NUMBER 1
1.0 INTRODUCTION
In accordance with the requirements of 10 CFR 55.45 (bX5Xii) and (bX5Xvi), this report is submitted for the Turkey Point Unit 3 Simulator. Turkey Point Unit 4 training will be performed using the Unit 3 Training Simulator. Changes to the test plan as described in this update do not change the basis for the Unit 4 certification. The differences between Unit 3 and Unit 4 were reviewed by the Simulator Configuration Review Board (SCRB)'nd it was determined that none will have a negative effect on operator training or examinations.
This report presents the information required to be submitted on a four year cycle per the referenced sections of 10 CFR 55.45. Included herein are the certification tests completed during the previous four years, outstanding Simulator Discrepancy Reports (DRs), and the schedule for completion of these DRs.
E xperience gained during initial certification testing and the testing during the first four year cycle, changes to the plant, and the availability of plant data have resulted in some changes to the plan. The changes made to the test plan for the next four year cycle are presented along with a discussion of the basis for the changes. The changes in the plan are organized into the following three categories: Changes, Deletions, and Additions.
Eight tests have been changed, ten tests have been deleted, and eleven tests have been added. Changes to the Turkey Point Simulator test plan have been reviewed and approved by the SCRB.
The certification test plan is presented for each of the next four years and the abstracts for the tests that have been added or changed are included in Appendix A.
'he Turkey Point SCRB was established by administrative procedure O-ADM-305, Simulator Configuration Management. The SCRB provides overall control and direction of changes to the Simulator. The SCRB also reviews and approves the Certification test program and test results. Membership on the SCRB is selected per the guidelines of the Institute of Nuclear Operations, "Simulator Configuration Management System," INPO 87-016, August 1987.
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SIMULATOR CERTIFICATION UPDATE NUMBER 1 2.0 SIMULATORTESTS COMPLETED (1991 - 1994) 2.1 1991 TESTS TEST 0 TEST DESCRIPTION COMPLETION DATE MFW-002 Loss of Normal Feedwater 06-27-91 MFW-007 Equivalent TMI-2 Scenario 11-14-91 MGG-002 Loss of 4KV Bus 3A 05-09-91 MGG-003 Loss of 4IOt'us 3B 05-09-91 MGG-004 Loss of All AC 05-24-91 MMP-001 Loss of Vital AC Bus 3P06 05-24-91 MMP-002 Loss of Vital AC Bus 3P07 05-24-91 MMP-003 Loss of Vital AC Bus 3P08 05-24-91 MMP-004 Loss of Vital AC Bus 3P09 05-30-91 MMP-005 Loss of DC Bus 3A (3D01) 05-30-91 MMP-006 Loss of DC Bus 3B (3D23) 05-30-91 MMP-007 Loss of DC Bus 4A (4D01) 05-30-91 MMP-008 Loss of DC Bus 4B (4D23) 05-24-91 MRC-006 Loss of a Single RCP With Power Below P-8 06-28-91 MRC-007 Stuck Open Spray Valve 06-28-91 MSP-001 Bus Stripping and Load Sequencing Tests 11-14-91 NPE-002 Plant Startup Cold Shutdown to Hot Standby 12-17-91 NPE-003 Plant Startup &om-Hot Standby to Rated Power 12-17-91 SUR-001 Initial Criticality after Refueling, OP-0204.3 11-27-91 ~
SUR-002 Nuclear Design Check Tests During Startup Sequence after Refueling, OP-0204.5 12-17-91 SUR-026 Engineered Safeguards Integrated Test, 3-OSP-203 12-17-91 SUR-030 Full Length RCC - Periodic Exercise, OP-1604.1 06-25-91 SUR-031 Inducing Xenon OsciQations to Produce Various Incore Axial Offsets, OP-12304.8 06-25-91 SUR-032 Normal Operation of Incore Moveable Detector System and Power Distribution Surveillance, OP-12404.1 11-14-91 Page 3
0 0
SIMJLATOR CERTIFICATION UPDATE NUMBER 1 1991 TESTS (CONTINUED)
ANNUALTESTS TEST ¹ TEST DESCRIPTION COMPLETION DATE MFW-003 Loss of Normal and Emergency Feedwater 12-18-91 MRC-002 Large Break LOCA Inside Containment With Loss Of OQsite Power 06-28-91 MRC-004 PORV Failure (Open) Without High Pressure Injection 12-17-91 MRC-005 Loss of Forced Reactor Coolant Flow 07-01-91 MRX-001 Spurious Rod Position Indication Resulting in Maximum Rate Runback To 70% Power and Maximum Rate Return to Full Power 06-25-91 MRX-009 Manual Reactor Trip from 100% Power 06-25-91 MSG-001 Main Steam Line Break Inside Containment 12-18-91 MSG-003 Simultaneous Closure of All MSIVs 11-14-91 MTU-001 Turbine Trip Which Does Not Cause Automatic Reactor Trip 06-27-91 SST-001 Steady State 45% Power Heat Balance 11-27-91 SST-002 Steady State 75% Power Heat Balance 11-27-91 SST-003 Steady State 100% Power Heat Balance 12-18-91
'SST-004 100% Power 60 min Null Transient 11-14-91 Page 4
SIMJLATOR CERTIFICATION UPDATE NUMBER 1 1991 TESTS (CONTINUED)
As a result of major plant changes that were incorporated into the Simulator in 1991, the following tests were also run., MGG-005 and MGG-006 are new. The balance of the tests were part of the original certification test plan.
TEST ¹ TEST DESCRIPTION COMPLETION DATE MFW-008 'oss of Feedwater / ATWS 12-17-91 MGG-005 'oss of 4KV Bus 3C 05-09-91 MGG-006 Loss of 4Iot'us 3D 2 05-09-91 MRC-001 'team Generator Tube Rupture 12-17-91 MRC-003 'mall Break LOCA Inside Containment 05-24-91 MRC-008 'oss of B and C Reactor Coolant Pumps at 100% Power 06-28-91 RTT-001 2 Simulator Real Time Test 05-09-91 RTT-002 2 Simulator Real Time Validation Test 05-09-91 SUR-003 2 EDG 8 Hour Load and Load Rejection Test, OP-4304.3 05-09-91 SUR-009 'eactor Protection Test, 3-OSP-049.1 11-14-91 SUR-021 2 Standby Steam Generator Feedwater Pumps
/ Cranking Diesels Test, O-OSP-074.4 05-09-91
'TD bypass loop removal / EAGLE-21 installation
'mergency Power System Enhancement (EPSE) project Page 5
TUIGGV POINT UNIT 3 SIMULATOR CERTIFICATION UPDATE NUMBER 1 2.2 1992 TESTS TEST 0 TEST DESCRIPTION COMPLETION DATE MCV-001 Uncontrolled Maximum Rate Boron Dilution 07-16-92 MCV-002 Charging System Failures 06-10-92 MCV-004 Letdown and Volume Control Tank System Operations and Malfunctions 07-16-92 MFW-004 Feedwater Line Break Inside Containment 07-10-92 MFW-006 Failure of Steam Generator Level Channel Providing Input to the Feedwater Controller 07-16-92 MRC-003 Small Break LOCA Inside Containment 07-10-92 MSG-006 Closure of a Single MSIV At Several Different Power Levels 06-10-92 MTU-003 Turbine Lube Oil System (Bearings) 12-01-92 MTU-004 Turbine Gland Seal System 06-10-92 MTU-005 Turbine Turning Gear Operation 12-01-92 MTU-009 Turbine Lube Oil Control gi; Auto-Stop Oil 06-10-92 MTU-010 Turbine Lube Oil Pump S Motor 05-08-92 MTU-011 Failure of Turbine Control Valve Spring 07-16-92 NPE-005 Plant Shutdown &om Rated Power to Hot Standby 06-19-92 NPE-006 Cooldown &om Hot Standby to Cold Shutdown 07-10-92 SUR-004 Component Cooling Water Pumps Low Header Pressure Start Test, 3-OSP-030.5 05-08-92 SUR-007 CVCS Boric Acid Transfer Flow Test, 3-OSP-046.2 03-19-92 SUR-008 Boric Acid. Transfer Pump 3B Transfer and Control Switch Test, 3-OSP-046.5 03-19-92 SUR-009 Reactor Protection System Logic Test, 3-OSP-049.1 05-08-92 SUR-012 Emergency Containment Filter Fans Operating Test, 3-OSP-056.1 03-19-92 SUR-015 Intermediate Range Nuclear Instrumentation Analog Channel Operational Test, 3-OSP-059.2 03-19-92 SUR-017 Power Range Nuclear Instrumentation Analog Channel Operational Test, 3-OSP-059.4 05-08-92 SUR-020 Main Steam Isolation Valve Closure Test 07-10-92 0 Page 6
0 SIMULATOR CERTIFICATION UPDATE NUMBER 1 1992 TESTS (CONTINUED)
TEST ¹ TEST DESCRIPTION COMPLETION DATE MFW-003 Loss of Normal and Emergency Feedwater 12-14-92 MRC-002 Large Break LOCA Inside Containment With Loss Of OHsite Power 07-10-92 MRC-004 PORV Failure (Open) Without High Pressure Injection 07-16-92 MRC-005 Loss of Forced Reactor Coolant Flow 06-10-92 MRX-001 Spurious Rod Position Indication Resulting in Maximum Rate Runback To 70% Power and Maximum Rate Return To Full Power MRX-009 Manual Reactor Trip from 100% Power MSG-001 Main Steam Line Break Inside Containment MSG-003 Simultaneous Closure of All MSIV's MTU-001 Turbine Trip Which Does Not Cause Automatic Reactor Trip 05-08-92 SST-001 Steady State 45% Power Heat Balance 12-15-92 SST-002 Steady State 75% Power Heat Balance 12-01-92 SST-003 Steady State 100% Power Heat Balance 07-17-92 SST-004 100% Power 60 min Null Transient 06-10-92 Page 7
SIMULATOR CERTIFICATION UPDATE NUMBER 1 1992 TESTS (CONTINUED)
As a result of protection system hardware changes and major Simulator upgrades incorporated in 1992, the following tests were also run.
TEST 4 TEST DESCRIPTION COMPLETION DATE MRX-002 Loss of Protection System Channel 12-16-92
'SS-003 Loss of RHR While in Cold Shutdown 07-10-92
'SS-004 Loss of Inventory during a Shutdown and Partial Draindown Condition 07-10-92 NPE-001 Plant Fill and Vent Rom a Partial Draindown to a Solid Pressurizer 07-17-92
'nstallation of EAGLE-21 hardware in the Simulator 2
Simulator draindown model upgrade Page 8
TUI~POINT UNIT 3 SIMULATOR CERTIFICATION UPDATE NUINBER 1 2.3 1993 TESTS TEST 4 TEST DESCRIPTION COMPLETION DATE MCS-001 Component Cooling Water Operations and Malfunctions Up To and Including Total Loss of CCW 11-18-93 MFW;001 Loss of Vacuum Tests, Including Loss of Condenser Level Control 12-06-93 MGG-001 Generator Trip 08-27-93 MRC-001 Steam Generator Tube Rupture 08-10-93 MRC-008 Loss of B and C Reactor Coolant Pumps at 100% Power 11-18-93 MRX-003 Nuclear Instrumentation Failure During Startup 11-18-93 MRX-006 Dropped Control Rod 12-06-93 MRX-007 Dropped With Inability to Drive Control Rods 11-18-93 MSG-004 Tamsmitter Failure Resulting In Maximum Atmospheric Dump Demand 12-06-93 MSG-005 Failure of Reference Temperature to Steam Dumps 11-18-93 MTU-002 Turbine Trip from 100% Power 11-18-93 NPE-001 Plant Fill and Vent from a Partial Drain Down to a Solid Pressurizer 11-15;93 SUR-003 EDG 8 Hour Load and Load Rejection Test, OP<304,3 11-18-93 SUR-005 Reactor Coolant System Leak Rate Calculations, 3-OSP-041.1 08-27-93 SUR-010 RHR MOVs/System Pressure Interlock Test, 3-OSP-050.7 08-27-93 SUR-011 RHR MOVs 750, 751, 862, 863, Interlock Test, 3-OSP-050.8 08-27-93 SUR-021 Standby Steam Generator Feedwater Pumps
/Cranking Diesels Test, O-OSP-074.4 08-27-93 SUR-022 Auxiliary Feedwater Train 1 Operability Verification, 3-OSP-075.1 08-27-93 SUR-024 Main Turbine Valves Operability Test, 3-OSP-089 11-18-93 Page 9
~ '
TURKEY POINT UNIT 3 SIMULATOR CERTIFICATION UPDATE NUMBER 1 1993 TESTS (CONTINUED)
ANNUALTESTS TEST ¹ TEST DESCRIPTION COMPLETION DATE MFW-003 Loss of Normal and Emergency Feedwater 08-10-93 MRC-002 Large Break LOCA Inside Containment With Loss Of OQsite Power 12-06-93 MRC-004 PORU Failure (Open) Without High Pressure Injection 08-10-93 MRC-005 Loss of Forced Reactor Coolant Flow 11-18-93 MRX-001 Spurious Rod Position Indication Resulting in Maximum Rate Runback To 70% Power and Maximum Rate Return To Full Power MRX-009 Manual Reactor Trip from 100% Power MSG-001 Main Steam Line Break Inside Containment MSG-003 Simultaneous Closure of AllMSIVs MTU-001 Turbine Trip Which Does Not Cause Automatic Reactor Trip 11-18-93 SST-001 Steady State 45% Power'Heat Balance 11-15-93 SST-002 Steady State 75% Power Heat Balance 11-15-93 SST-003 Steady State 100% Power Heat Balance 11-15-93 SST-004 100% Power 60 min Null Transient 08-27-93 Page 10
TURKEY POINT UNIT 3 SIMULATOR CERTIFICATION UPDATE NUMBER 1 O 2.4 1994 TESTS TEST 0 TEST DESCRIPTION COMPLETION DATE MCN-001 Containment Spray System Operations and Malfunctions 08-09-94 MCS-002 Intake Cooling Water System Operations and Malfunctions 03-24-94 MCS-003 Turbine Plant Cooling Water Operation and Malfunctions 03-24-94 MCS-004 Instrument Air System Operation and Malfunctions 03-24-94 MCV-003 Charging Line Break Outside Containment 07-08-94 MCV-005 Non-Regenerative Heat Exchanger Tube Leak 03-24-94 MFW-005 Main Feedwater Line Break Outside Containment 05-17-94 MFW-008 Loss of Normal Feedwater/ATWS 08-05-94 MRX-002 Loss of Protection System Channel 06-06-94 MRX-004 Stuck Control Rod 07-08-94 MRX-005 Uncoupled Control Rod Test 07-08-94 MRX-008 Fuel Cladding Failure Resulting in High Reactor Coolant Activity 08-05-94 MSG-002 Main Steam Line Break Outside Containment 07-08-94 MSS-001 Small Leak in Safety Injection Piping Outside Containment 05-17-94 MSS-002 Accumulator Operations and Malfunctions 08-05-94 MSS-003 Loss of RHR While in Cold Shutdown 05-17-94 MSS-004 Loss of Inventory During A Shutdown and Partial Draindown Condition 05-17-94 MTU-006 Hydrogen Seal Oil 06-06-94 MTU-008 Hydrogen Cooling 06-06-94 NPE-004 Reactor Trip Followed By Recovery to Rated Power 08-09-94 RTT-001 Simulator Real Time Test 05-17-94 RTT-002 Simulator Real Time Test Validation Test 05-17-94 SUR-014 Source Range Nuclear Instrumentation Analog Channel Operational Test, 3-OSP-059.1 03-24-94 SUR-016 Intermediate Range NIS Setpoint Verification, 3-OSP-059.3 08-05-94 Page 11
0 TURKEY'OINTUNIT 3 SIMJLATOR CERTIFICATION UPDATE NUMBER 1 1994 TESTS (CONTINUED)
SUR-018 Power Range Nuclear Instrumentation Shift Checks and Daily Calibration, 3-OSP-059.5 03-24-94 SUR-019 Process Radiation Monitoring Operability Test, 3-OSP-067.1 08-05-94 SUR-029 Operational Test of MOV-535, 536, and PORV 455C,456, OP-1300.2 03-24-94 ANNUALTESTS TEST ¹ TEST DESCRIPTION COMPLETION DATE MFW-003 Loss of Normal and Emergency Feedwater 05-17-94 MRC-002 Large Break LOCA Inside Containment With Loss Of Offsite Power 05-17-94 MRC-004 PORV Failure (Open) Without High Pressure Injection 05-17-94 MRC-005 Loss of Forced Reactor Coolant Flow 04-29-94 MRX-001 Spurious Rod Position Indication Resulting in Maximum Rate Runback To 70% Power and Maximum Rate Return To Full Power 03-24-94 MRX-009 Manual Reactor Trip from 100% Power 03-24-94 MSG-001 Main Steam Line Break Inside Containment 05-17-94 MSG-003 Simultaneous Closure of All MSIVs 04-29-94 MTU-001 Turbine Trip Which Does Not Cause Automatic Reactor Trip 06-06-94 SST-001 Steady State 45% Power Heat Balance 08-05-94 SST-002 Steady State 75% Power Heat Balance 08-05-94 SST-003 Steady State 100% Power Heat Balance 08-05-94 SST-004 100% Power 60 min Null Transient 03-24-94 Page 12
'ZUIGGrY POINT UNIT 3 SIMULATOR CERTIFICATION UPDATE NUMBER 1 3.0 CERTIFICATION TEST CHANGES (1995 - 1998)
Experience gained during initial certification testing and during the first four year cycle has resulted in several changes to the Turkey Point Simulator Certification Test plan. Furthermore, the availability of plant data has resulted in additions to the plan. These changes in the plan are organized into the following three categories: Changes, Deletions, and Additions.
Changes have been made to eliminate duplication, eliminate test runs that provide little or no incremental information, and to modify tests to reQect changes in the plant.
Deletions have been made as a result of changes in plant configuration and, to eliminate duplication, Five surveillance test procedures, not deemed. sufBciently valuable to repeat for the next four year cycle, were deleted to allow substitution of more important surveillance procedures.
Additions have been made as a result of plant changes, to make use of plant transient data from the Emergency Response Data Acquisition Display System (ERDADS), and to add new surveillance procedures.
3.1 CHANGES MCN-001, Containment Spray System Operations and Malfunctions.
This test will be generalized to include all of the containment systems that provide for mitigation of events that cause pressurization of the containment, i.e., both the spray system and the containment emergency coolers. The overall behavior of the containment pressure and temperature response is monitored in other certification tests such as the large and small break LOCAs, and the inside containment MSLB.
This test will focus on the response of the individual systems to malfunctions. The test will be renamed "Containment Emergency Systems Operations and Malfunctions."
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TUNG~ POINT UNIT 3 SI1VKLATOR CERTIFICATION UPDATE NUMBER 1 MCS-001, Component Cooling Water Operations and 1Vbdfunctions.
This test comprised two runs: Run 1, Loss of Intake Cooling Water (ICW) to the Component Cooling Water (CCW) Heat Exchangers, and Run 2, Total Loss of CCW Flow. Run 1, Loss of ICW to the CCW Heat Exchangers is performed in MCS-002, Intake Cooling Water System Operations and Malfunctions, and willbe deleted from MCS-001. The Total Loss of CCW Flow will continue to be performed in MCS-001.
MCS-003, Turbine Plant Cooling Water System Operations and Malfunctions.
This test comprised two runs; Run 1, Loss of Intake Cooling Water (ICW) to the Turbine Plant Cooling Water (TPCW) Heat Exchangers, and Run 2, Total Loss of TPCW. Run 1, Loss of ICW to the TPCW Heat Exchangers is performed in MCS-002, Intake Cooling Water System Operations and Malfunctions, and willbe deleted from MCS-003. The Total Loss of TPCW will continue to be performed in MCS-003.
MCV-004, Letdown and Volume Control Tank System Operations and Malfunctions.
This test comprised a total of five diQerent malfunctions, Run 1, Loss of CCW to the Non-Regenerative Heat Exchangers is performed in MCS-001, Component Cooling Water Operations and Malfunctions and willbe deleted from MCV-004.
The remaining four malfunctions (PCV-145 failed open, PCV-145 failed shut, LCV-115A failed to the divert position, and CV-204 failed shut) will continue to be tested as before.
MFW-002, Loss of Normal Feedwater.
Previous certification testing examined two cases, the first with the Auxiliary Feedwater (AFW) Qow controllers set at 135 gpm and the second with the controllers set at 300 gpm. The two tests were designed to examine the sensitivity of the response to the magnitude of the AFW flow rate. Experience has shown that the response of the two cases is largely the same with no additional information being gained by performing the second run. Therefore, the 300 gpm case willbe Page 14
TURKlH'OINTUNIT 3 SI1VKILATORCERTIFICATION UPDATE NUMBER 1 deleted. Also, the setting for the AFW How controller demand willbe changed to the current plant setting of 130 gpm.
MRX-001, Spurious Rod Position Indication Resulting in Maximum Rate Runback to 70% Power and Maximum Rate Return to Full Power.
The turbine runback due to dropped mds was removed during the Unit 3 Cycle 14 refueling outage. The test was changed to remove the dropped rod runback &om the scenario and the test title was changed to "Maximum Rate Power Ramp (100%
down to 75% and back up to 100%)". The power reduction and ascension willbe controHed manually by the test team.
MSG-001, Main Steam Line Break Inside Containment.
The initial certification testing and annual testing included two Main Steam Line Break (MSLB) cases. In the first case the RCPs were tripped based on the RCP trip criteria &om the Emergency Operating Procedures (EOP). The second case was performed with the RCPs running as a sensitivity study to provide additional information for the comparison of the Simulator with the Best Estimate RETRAN model. The case with RCPs running is atypical of the expected sequence during a MSLB and provides little additional information to justify continuing its performance as an annual test, Therefore, the "RCPs On" case will be deleted.
(See Section 3,3 for an additional Hot Zero Power (HZP) MSLB test that will be included for the next four year cycle,)
MSG-003, Simultaneous Closure of All MSVPs.
The initial certification test plan called for two runs to be performed. The first examined the response with rod control in automatic, and the second with rod control in manual. Previously, these runs were significantly different because the rods in manual case would quickly trip whereas the rods in automatic case would survive for some time. During the Unit 3 Cycle 13 refueling outage a reverse power turbine trip was installed in the plant. This results in a turbine trip /
reactor trip after approximately 60 seconds making both runs very similar.
Therefore, the "Rods in Manual" case will be deleted.
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TURIG'rY POINT UNIT 3 SIMJLATOR CERTIFICATION UPDATE NUMBER 1 3.2 DELETIONS MCV-003, Charging Line Break Outside Containment.
This scenario is performed in MCV-002 Charging System Failures and therefore willbe deleted.
MRC-008, Loss of B and C RCPs at 10¹ power.
This test was compared to a plant trip that occurred on Unit 4 on 4/9/90.
Numerous plant changes have occurred since 1990 including removal of the RTD bypass loops and changes to the protection and process control systems. These changes make the comparison between the present Simulator and the 4/9/90 event inappropriate.
MSG-004, Transmitter Failure Resulting in Maximum Atmospheric Dump Demand.
One atmospheric dump valve has a capacity of only 3.3% of full steam Qow. This results in a change in steam load similar to normal operating changes. The transient is relatively mild and does not provide unique information. Therefore, this test willbe deleted, MTU-004, Turbine Gland Seal System.
This test checks the normal operation of the gland sealing system. These checks are performed during NPE-003, Plant Startup Rom Hot Standby to Rated Power.
Therefore, this test willbe deleted, MTU-010, Turbine Lube Oil Pump and Motor.
This test was mis-titled in the original certification test plan. The test is actually a test of the Steam Generator Feed Pumps (SGFP) Lube Oil system. This is a minor support system which has no indications in the control room and no role in training. Continuing to test this system provides little value in terms of training and this test willbe deleted.
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TURKFrY POINT UNIT 3 SIMULATOR CERTIFICATION UPDATE NUMBER 1 SUR-007, CVCS Boric Acid Transfer Flow Test, 3-OSP-046.2 This test verifies that the boric acid pumps can deliver Qow through the normal and emergency Qowpaths. These Qowpaths are verified during other certification tests such as the Plant Shutdown &om Rated Power to Hot Standby (NPE-005),
the cooldown from Hot Standby to Cold Shutdown (NPE-006), and during operator training scenarios such as the ATWS event. Since initial certification testing and testing during the first four year cycle have demonstrated the ability to perform this surveillance procedure and the substantive elements of the surveillance are examined via other tests, this test willbe deleted and surveillance test SUR-033 substituted. (See Section 3.3 - Additions)
SUR-008, Boric Acid Transfer Pump Transfer and Control Switch Test, 3-OSP-046.5 This test checks the operation of the boric acid pump &om both the local control station and the control room. Certification testing during the first four year cycle has demonstrated the ability to perform this surveillance. Continuing to test this surveillance provides little value in terms of training and therefore this test willbe .
deleted and surveillance test SUR-034 substituted. (See Section 3.3 - Additions)
SUR-012, Emergency Containment Filter Fan Operating Test, 3-OSP-056.1.
Since there are limited filter fan indications inside the control room and initial certification testing and additional testing during the first four year cycle have demonstrated the ability to perform this surveillance, this test will be deleted and surveillance test SUR-035 substituted. (See Section 3.3 - Additions)
SUR-018, Power Range NIS Shift Checks and Daily Calibrations, 3-OSP-059.5.
This test is performed during several other tests including the annual steady state tests (SST-001, SST-002, and SST-003). Therefore, a separate test is not necessary. This test willbe deleted and surveillance test SUR-036 substituted.
(See Section 3.3 - Additions)
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TURKEZ'OINTUNIT 3 SIMULATOR CERTIFICATION UPDATE NUMBER 1 SUR-029, Operational Test of MOV-536, 636, and PORV-465C, 466, OP-1300.2.
This test performs a stroke test on the PORV block valves and seat leakage test on the PORVs. Since initial certification testing and additional testing during the first four year cycle have demonstrated the ability to perform this surveillance, this test willbe deleted and surveillance test SUR-037 substituted. (See Section 3.3 - Additions) 3.3 ADDITIONS MGG-005, Loss of 4KV Bus 3C.
This test was added in 1991 following the Emergency Power System Enhancement project. This test verifies that the load centers, motor control centers and 480v loads powered from the 3C bus respond correctly to the loss of the 3C 4kv bus.
MGG-006, Loss of 4KV bus 3D.
This test was added in 1991 following the Emergency Power System Enhancement (EPSE) project. The 3D 4kv safety related bus was installed in 1991 during the EPSE project. This test verifies that the 3H load center, 3D motor control center and 480v loads powered &om the 3D bus respond correctly to the loss of the 3D 4kv bus.
MRC-009, Fast Load Reduction, 3-ONOP-100..
This test will simulate high RCP seal leakoK requiring a fast load reduction using 3-ONOP-100. The Simulator results will be compared to plant data &om the Unit 3 fast load reduction that occurred on 4-27-92.
MRX-010, Spurious High Containment Pressure Safety Injection.
This test will simulate a spurious high containment pressure SI &om 28% power.
The Simulator results will be compared to plant data &om the Unit 4 high containment pressure SI that occurred on 3-26-92.
Page 18
TURKEY POINT UNIT 3 SIMULATOR CERTIFICATION UPDATE NUMBER 1 MRX-011, Loss of C 4KV Bus Reactor Trip.
This test will simulate a loss of the C 4kv bus from 10¹ power. The Simulator results will be compared to plant data &om the Unit 4 loss of C 4kv bus reactor trip that ocnnred on 9-23-94.
MSG-007, Main Steam Line Break with Reduced Shutdown Margin.
This test will examine a hypothetical Main Steam Line Break (MSLB) &om Hot Standby with a shutdown margin that will result in a return to power during the cooldown. This will examine a MSLB event scenario with slightly different characteristics than the standard test (MSG-001).
SUR-033, Accident Monitoring Instrumentation Channel Checks, 3-OSP-204.
This test covers the surveillance of the accident instrumentation including core exit thermocouples, reactor vessel level indicators, subcooled margin monitors, and containment radiation, pressure and level indicators.
SUR-034, Safeguard Relay Rack Train A,B, Periodic Test, OP-4004.2.
This test covers the operator surveillance of the safeguards logic.
SUR-035, Containment Isolation Racks QR50 and QR51 Periodic Test, OP-4004.4.
This test covers the operator surveillance of the containment pressure channels.
SUR-036, Component Cooling Water System Flow Balance, 3-OSP-030.9.
This test verifies that all safety related components cooled by CCW receive the minimum required Qow with the CCW system aligned in its most limiting accident configuration.
SUR-037, Determination of Quadrant Power Tilt Ratio, 3-OSP-059.10.
This test covers the determination of the Quadrant Power Tilt Ratio (QPTR).
Page 19
TUIGGZ'OINTUNIT 3 SIMULATOR CERTIFICATION UPDATE NUMBER 1 4.0 FOUR YEAR.TEST PLAN (1995-1998)
Per the requirements of Regulatory Guide 1.149, the Simulator Certification test program willbe conducted in its entirety on a four year cycle. All of the ANSVANS 3.6 Appendix B tests will be performed annually. Approximately 25% of the remaining tests in the Certification program will be performed each year.
Table 4-1 presents the ANSVANS 3.6 tests that will be performed annually. Tables 4-2 through 4-6 present the test plan for each of the next four years. The tests planned each year represent a cross section of the various types of tests. As in the previous four year cycle, tests may be added to meet new or special requirements.
The 1996-1998 Turkey Point Simulator test plan has been reviewed and approved by the Simulator Configuration Review Board (SCRB).
Page 20
TUIGGlY POINT UNIT 3 SI1UKLATOR CERTIFICATION UPDATE MiMBER 1 Table 4-1 Annual Tests MFW-003 Loss of Normal and Emergency Feedwater MRC-002 : Large Break LOCA Inside Containment With Loss Of Offsite Power
'RC-004 PORV Failure (Open) Without High Pressure Injection MRC-005 Loss of Forced Reactor Coolant Flow MRX-001 Maximum Rate Power Ramp (100% To 75% and back to 100%%uo)
MRX-009 Manual Reactor Trip from 100% Power MSG-001 Main Steam Line Break Inside Containment MSG-003 Simultaneous Closure of All MSIVs MTU-001 Turbine Trip Which Does Not Cause Automatic Reactor Trip SST-001 Steady State 50% Power Heat Balance SST-002 Steady State 75% Power Heat Balance SST-003 Steady State 100% Power Heat Balance SST-004 100% Power 60 Minute Null Transient Page 21
TUI~POINT UNIT 3 SIMULATOR CERTIFICATION UPDATE NUMBER 1 Table 4-2 1995 Test Plan MFW-002 Loss of ¹rmal Feedwater MFW-007 Equivalent TMI-2 Scenario MGG-002 Loss of 4KV Bus 3A MGG-003 Loss of 4KV Bus 3B MGG-004 Loss of All AC MGG-005 Loss of 4KV Bus 3C MGG-006 Loss of 4IOt'us 3D MMP-001 Loss of Vital AC Bus 3P06 MMP-002 Loss of Vital AC Bus 3P07 MMP-003 Loss of Vital AC Bus 3P08 MMP-004 Loss of Vital AC Bus 3P09 mrtP-005 Loss of DC Bus 3A (3D01)
Loss of DC Bus 3B (3D23)
MMP-007 Loss of DC Bus 4B (4D01)
MMP-008 Loss of DC Bus 4A (4D23)
MRC-006 Loss of a Single Reactor Coolant Pump With Power Below P-8 MRC-007 Stuck Open Spray Valve MSG-007 Main Steam Line Break from Hot Standby with Reduced Shutdown Margin.
MSP-001 Bus Stripping and Load Sequencing Tests NPE-002 Plant Startup Cold Shutdown to Hot Standby NPE-003 Plant Startup &om Hot Standby to Rated Power SUR-026 Engineered Safeguards Integrated Test, 3-OSP-203.1 & 3-OSP-203.2 SUR-030 Full Length RCC - Periodic Exercise, OP-1604.1 SUR-031 Inducing Xenon Oscillations to Produce Various Incore Axial Offsets, 0-OP-059.3 Page 22
0 TUIGGrY POINT UNIT 3 SIMULATOR CERTIFICATION UPDATE NUMBER 1 Table 4-3 1996 Test Plan MCV-001 Uncontrolled Maximum Rate Boron Dilution MCV-002 Charging System Failures MCV-004 Chemical Volume Control System Operations and Malfunctions MFW-004 Feedwater Line Break Inside Containment MFW-006 Failure of Steam Generator Level Channel Providing Input to the Feedwater Controller MRC-003 Small Break LOCA Inside Containment MRC-009 Fast Load Reduction, 3-ONOP-100 MRX-010 Spurious High Containment Pressure SI MSG-006 Closure of a Single MSIV At Several Different Power Levels MTU-003 Turbine Lube Oil System (Bearings)
MTU-005 Turbine Turning Gear Operation MTU-009 Turbine Lube Oil Control & Auto-Stop Oil MTU-011 Failure of Turbine Control Valve Spring NPE-005 Plant Shutdown &om Rated Power to Hot Standby NPE-006 Cooldown &om Hot Standby to Cold Shutdown SUR-004 Component Cooling Water Pumps Low Header Pressure Start Test, 3-OSP-030.5 SUR-009 Reactor Protection System Logic Test, 3-OSP-049.1 SUR-015 Intermediate Range Nuclear Instrumentation Analog Channel Operational Test, 3-OSP-059.2 SUR-017 Power Range Nuclear Instrumentation Analog Channel Operational Test, 3-OSP-059.4 SUR-020 Main Steam Isolation Valve Closure Test, 3-OSP-072 SUR-035 Containment Isolation Racks QR50 and QR51 Periodic Test, OP-4004.4 SUR-036 Component Cooling Water System Flow Balance, 3-OSP-030.9 SUR-037 Determination of Quadrant Power Tilt Ratio, 3-OSP-059.10 Page 23
TURKIC Y POINT UNIT 3 SIMULATOR CERTIFICATION UPDATE NUMBER 1 Table 4-4 1997 Test Plan MCS-001 Component Cooling Water Operations and Malfunctions MFW-001 Loss of Vacuum Tests, Including Loss of Condenser Level Control MGG-001 Generator Trip MRC-001 Steam Generator Tube Rupture MRX-003 Nuclear Instrumentation Failure During Startup MRX-006 Dropped Control Rod MRX-007 Dropped With Inability to Drive Control Rods MRX-011 Loss of C 4KV Bus Reactor Trip MSG-005 Failure of Reference Temperature to Steam Dumps MTU-002 Turbine Trip &om 100% Power NPE-001 Plant Fill and Vent &om a Partial Drain Down to a Solid Pressurizer RTT-001 Simulator Real Time Test RTT-002 Simulator Real Time Test Validation Test SUR-003 EDG 8 Hour Load Test and Load Rejection Test, 3-OSP-023.2 SUR-005 Reactor Coolant System Leak Rate Calculations, 3-OSP-041.1 SUR-010 RHR MOVs/System Pressure Interlock Test, 3-OSP-050.7 SUR-011 RHR MOVs 750, 751, 862, 863, Interlock Test, 3-OSP-050.8 SUR-021 Standby Steam Generator Feedwater Pumps/Craning Diesels Test, O-OSP-074.4 SUR-022 Auxiliary Feedwater Train 1 Operability Verification, 3-OSP-075.1 SUR-024 Main Turbine Valves Operability Test, 3-OSP-089 SUR-032 Normal Operation of Incore Moveable Detector System and Power Distribution Surveillance, OP-12404.1 SUR-033 Accident Monitoring Instrumentation Channel Checks, 3-OSP-204 SUR-034 Safeguard Relay Rack Train A,B, Periodic Test, OP-4004.2 Page 24
TUI~ POINT UNIT 3 SIMULATOR CERTIFICATION UPDATE NUMBER 1 Table 4-5 1998 Test Plan MCN-001 Containment Emergency Systems Operation and Malfunctions MCS-002 Intake Cooling Water System Operations and Ma1lfunctions MCS-003 Turbine Plant Cooling Water Operation and Malfunctions MCS-004 Instrument Air System Operation and Malfunctions MCV-005 Non-Regenerative Heat Exchanger Tube Leak MFW-005 Main Feedwater Line Break Outside Containment MFW-008 Loss of Normal Feedwater/ATWS MRX-002 Loss of Protection System Channel MRX-004 Stuck Control Rod MRX-005 Uncoupled Control Rod Test MRX-008 Fuel Cladding Failure Resulting in High Reactor Coolant Activity MSG-002 Main Steam Line Break Outside Containment MSS-001 Small Leak in Safety Injection Piping Outside Containment MSS-002 Accumulator Operations and Malfunctions MSS-003 Loss of RHR While in Cold Shutdown MSS-004 Loss of Inventory During Partial Draindown MTU-006 Hydrogen Seal Oil MTU-008 Hydrogen Cooling NPE-004 Reactor Trip Followed By Recovery to Rated Power SUR-001 Initial Criticality aRer Refueling, 3-OSP-040.6 SUR-002 Nuclear Design Check Tests During Startup Sequence after Refueling, 3-OSP-040.5 SUR-014 Source Range Nuclear Instrumentation Analog Channel Operational Test, 3-OSP-059.1 SUR-016 Intermediate Range NIS Setpoint Verification, 3-OSP-059.3 SUR-019 Process Radiation Monitoring Operability Test, 3-OSP-067.1 Page 25
TUI~ POINT UNIT 3 SIMULATOR CERTIFICATION UPDATE NUMBER 1 5.0 OUTSTANDING DISCREPANCIES In general all Simulator Discrepancy Reports (DRs) are addressed within one calendar year after they are written. The schedule may be extended beyond one year in special circumstances such as the need for certain equipment or very low priority DRs. The SCRB will review and approve the extension of the schedule for any DR that will not be completed in one year.
5.1 1991 CERTIFICATION TEST DR'S 73 Certification test DRs were written and all have been completed.
5.2 1992 CERTIFICATION TEST DR'S 69 Certification test DRs were written and all have been completed.
5.3 1993 CERTIFICATION TEST DR'S 28 Certification test DRs were written and all have been completed.
5.4 1994 CERTIFICATION TEST DR'S 26 Certification test DRs were written and all but the following DRs have been completed.
TEST DR 8 TITLE DUE DATE MRX005 9400069 Secondary side oscillations after rod drop 5/16/95 MRX008 9400114 Investigate R-11/R-12 response 8/05/95 NPE004 9400115 RCP current drops when transferring to the auxiliary Transformer 8/08/95 SUR019 9400075 PRMS drawers do not display exact alarm setpoint 6/10/95 N/A 9400116 Physical Fidelity 8/09/95 Page 26
TUIGG~~Y POINT UNIT 3 SIMULATOR CERTIFICATION UPDATE NUMBER 1 APPENDIX A TEST ABSTRACTS A.1 TESTS ADDED
OINT SIMULATOR CERTIFICATION TEST PROCEDURE TITLE: LOSS OF 4KV BUS 3C NUMBER MGG~
ANS 3.5 REFERENCE SECllONS: J. T,RP) Loss of Electrical Power DESCRIPllON The purpose of this test ls to verify proper simulator response to a loss of the 3C 4kV bus or to a loss of an Individual load center or motor control center suppNed from the 3C 4kV bus. This bus suppNes load centers 3E, 3F. and 3G and these supply motor control centers 3B non-vital, 3C non-vital, 3G, 3H, 3B43, RB (which feeds motor control center RC). and F (which feeds motor control center 39. ANloods of 450 volts or greater in this train that hove some Impact ln the control foam; e.g.,
alarm, control, or Indication, wiN be verified to lose power when the appropifate 4kV bus, load center, or motor control center is energized. This test wm be performed by failing open the supply breaker to each motor control center, loads on that motor control center w8 be verified lost. and the power wI then be restored to the motor control center. After aN motor control centers have been tested. the supply breaker to each load center wm be fai7ed open, loads on that load center willbe verified lost, Including power to the motor control centers. and the power to the load center wiN then be restored. After aN load centers have been tested. power to the 3C 4kV bus wi7I be lost and loads verified, Including power to the toad centers. In order to make lt easier to start and stop loads, this test willbe conducted from hot standby.
OPllONS The N3 transformer feeders to the 3C 4kV feeders can be operated by a variety of mechanisms. These Include operating them from the control room console, operating them from the Instructor's facility, using the instructor's faci7ity to operate them locally, and placing in a malfunction that w8 cause them to open or prevent their operation. The load center feeders have sfmNar options. The motor control center feeders must be opened from the Instructor's faclrity.
INmAL CONDmONS FINAL CONDIllONS Hot standby. The 3C 4kV bus is supplied from the P3 transformer. Hot standby with the 3C 4kV bus ~nerglzed. The data coNeclion sheet for loss of 3C 4kV bus has been completed.
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INT SIMULATOR CERTIFICATION lEST PROCEDURE EKE: LOSS OF 4KV BUS 3D NUMBER: MGG~
ANS 8.5 REFERENCE SECllONS: 8.1.2(J) Loss of ElecMcal Power DESCRIPTION The purpose of this test ls to verify proper simulator response to a toss of the 3D 4kV bus. lhe 3D4kV bus can be supplied from either the 3A 4kV bus or the 3B4kV bus. The 4kV supply to the 3D bus willbe manually swapped from 3B to 3A and back to 3B In order to verify the proper operation of this function. Since the 3D 4kV bus ls part of the station blackout gBO) cross connection between Unit 3 and Unit 4, that function wi also be tested In several situations. lhe interlocks that prevent paraMing trains 3A and 3B via the 3D4kV bus and the 125 VDC transfer switch 3$ 75 wN be tested. lhe transfer switch provides control power forthe drcult breakers on the 3D 4kV bus. In order to make it easier to start and stop loads and swap busses, this test weal be conducted from hot standby.
OPllONS During normal operatIons, the 4kV bus 3D can be supplied from the 3A or 3B 4kV bu* lhe breakers being tested can be opened and closed by a vorfety of mechanisms. These indude operating them from the control room console, operating them from the instructor's facility, using the instructors facility to operate them locally, and placing In a malfunction that wm cause them to open or prevent their operation.
INlllALCONDITIONS RNAL CONDlllONS Hot standby. The 3D 4kV bus ls energized from the 3B 4kV bus. Hot standby. The data collection sheet for loss of 4kV bus 3D has been completed, the manual transfer of supply busses has been completed, the Interlocks that prevent parallelling the 3A and 3B trains via the 3D 4kV bus have been tested, transfer switch 3$ 75 has been tested, and the SBO function has been tested.
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P OINT SIMUIATOR CERTIFICATION TEST PROCEDURE TITLEr FAST LOAD REDUCllON, M)NOP-100 NUMBER MRC~
ANS 8.5 REFERENCE SECTIONSr 3.1.1(5) Load Changes 3.1.2@5 Process Instrumentatton, Alarms, and Control System Failures DESCRIPllON This test wilt simuiate excessive Ieakoff from the Number 1 seal on the 3C RCP requiring a fast reduction In power using 3C)NOP-100, Fast Load Reduction. lhis scenario has been chosen to compare the simuiator response to the Unit 3 fast toad reduction on Ap8 27, 1992 from the same cause. Actions wil be simutated to approximate those taken during the Unit 3 load reduction.
OPllONS None INITIALCONDITIONS FINAL CONDITIONS 87% power, Steady state PIant stabie at hot standby and the C RCP secured.
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OINT SIMULATOR CER11RCAllON TEST PROCEDURE llTLEr SPURIOUS HIGH CONTAINMENTPRESSURE SAFETY INJECllON NUMBER: MR)M10 ANS S.5 REFERENCE SECllONS: S.1.2(19) Reactor Tifp DESCRIPllON This test wfiisimuiate a reactor trip caused by a B train Safety Injecffon signaiInitfated durfng the perfonnance of the containment prelure channel test. ibis scenario has been chosen to compare the simulator response to the Unit 4 trfp on March 26, 1992 tram the same cause. Actions wfiibe simulated to approxfmate those taken foiiowfng the Unit 4 trfp.
OPllONS INIllALCONDITIONS FihfAL CONDIllOh5 28% power, Steady state Plant stabie at hot standby.
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OINT SIMULATOR CENlFICAllONTEST PROCEDURE llRE: LOSS of C 4KV BUS REACTOR IIP NUMBER MRX<11 ANS 8.5 REFERENCE SECIlONS: 8.1.2(8) Loss of Eiectrtcai Power 8.1.2(12) Control Rod Failures DESCRIPllON This test wisimulate a loss of the C 4KVbus with power at 100%, which wm result In a reactor trip. Thisscenarlo has been chosen to compare the simulator response to the Unit 4 trip on September 23, 1994 from the same cause. Due to degraded power horn the MG set to the IAC control rod power cabinet and the loss of the 4C 4RV bus which supplies backup power, the 12 rods supplied from the IAC power cabinet dropped into the core resuNng In an OTAT reactor trip. Actions wIIIbe simulated to approxfmate those taken following the Unit 4 Irfp.
OPllONS INllMLCONDlllONS FINAL CONDmONS 1%% power, EOL Steady state. Plant stable at hot standby.
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OINT SIMULATOR CERTIFICATION TEST PROCEDURE llILEr MAINSTEAM ONE BREAK NTH REDUCED SHUTDONNMARGIN NUMBER MSGR ANS J.5 REFERENCE SECllONS: 8.1.2L20) Mafn Steam Line as well as Main Feed One Breaks goth inskte and outside Containment)
DESCRIPllON This steam line break tronsfent wS examine the response of the simulator to amain steam line break inskfe containment from hot standby conditfons with a reduced shutdown margin. The shutdown margfn willbe reduced to a degree that wfllallow a return to power as a resuft of the moderator feedback during the cooktown transient. The test Is not Intended to followIn detail the emergency operoffng procedures covering this type of transient. However. operator actions to turn offthe reactor coolant pumps on Iow su&cooling margin and isolate the auxiliary feedwoter to the affected steam generator have been programmed into the scenarfo.
No other operator actions willbe taken during the coune of the event. Allcontrol and safety systems wSbe in automatic ond fully functional. A steam line break equivalent to the area of the flow reshfctor at the steam generator outlet is assumed to occur in the B steam line Inskte containment.
OPTIONS The simulator is capable of simulating steam Tine breaks of any stre at several locations Inside and outside containment on each of the steam Enes.
INmAL coNDmoNs FINAL coNDmoNs Hot Standby, Subcrfticof, EOL The test wS run for 15 minutes.
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OINT SIMUIATOR CERllFICATION TEST PROCEDURE NIE: ACCIDENTMONITORING INSlRUMENTAllONCHANNB. CHECKS, M)SP404 NUMBER: SUR-D38 ANS 8.5 REFERENCE SECllONSr 8.1.1 (ID) Operator Conducted Survellkrnce on Safety-Related Equipment or Systems DESCRIPllON This certification test wI demonstrate the ability of the slmukrtor to support the operator conducted surveNance procedure 3~P-2N, Acckfent Monitoring Instrumentation Channel Checks. With no malfunctions present. the ability to successfully perform this survesance willbe verNed.
OPllONS This test can be performed at any time in core fife.
INITIAI.CONDITIONS FINAI. CONDITIONS 100% power. steady state, MOL SurveNance complete.
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INT SIMULATOR CER77FICATION TEST PROCEDURE TITLE: SAFEGUARD REIAYRACX TRAINA,B, PERIODIC TEST, OPADLR2 NUMBERs SUR~
ANS 3.5 REFERENCE SECTIONSs 3.1.1 (10) Operator Conducted Surveiiksnce on Safety-Related Equipment or Systems DESCRIPTION This certification test willdemonstrate the aMity of the simulator to support the operator conducted surveIance procedure OPAN42. Safeguard Relay Rack Train A,B, Periodic Test. With no malfunctions present, the ability to successfully perform this survei7!ance wi7I be verified.
OP17ONS This test can be performed at any power level and at time in core life.
INITIALCONDmONS RNAL CONDmONS IR% power, steady state. MOL Surveillance complete.
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INT SIMUlATOR CERllFICAlloNTEST PROCEDURE llILE: CONTAINMENTISOIA11ON RACKS CÃ50 AND QR51 PERIODIC TEST, OPMD04.4 NUMBER SUR-N5 ANS 8.5 REFERENCE SEClloNS: 8.1.1 (IO) Operator Conducted Surveiikrnce on Safety-Retated Equipment or Systems DESCRIPlloN This certÃcation QR50 and QR51 ~c test wIIIdemonstrate the abiiity of the simulator to support the operator conducted surveillance procedure OM0044, Contairment isoiatlon Racks Test. With no maifunctions present, the abffity to successfuiiy perform this surveillance wI be verified.
OPTIONS lhis test can be performed at any power level and at time in core kfe.
INmAL CoNDmoNS FINAL coNDmoNs 100% power, steady state, MOL Surveiilance complete.
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INTSIMULATOR CERTIFICATION TEST PROCEDURE llREi COMPONENT COOLING WATER SySTEM FLOW BALANCE, 3&SP~.9 e
NUMBER: SURES ANS 3,5 REFERENCE SECllONS: 3.1.1 (10) Operator Conducted Surveiikrnce on Safety-Related Equipment or Systems DESCRIPllOH lhls certification test wm demonstrate the abilityof the simulator to support the surveillance procedure 3~F0.9. Component Cooling Water System How Bakince.
This test verities that all safety rekrted components cooled by CCW receive the minimum required flow with the CCW system aligned In its most ilmNng acckient conjuration. With no malfunctions present, the abiTity to successfully perfom this surveillance wIIIbe ventied.
OPTIONS lhfs test can be performed at time ln core life.
INITIALCONDmONS FINAL CONDITIONS Cokt shutdown Survelance complete.
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OINT SIMULATOR CERTIFICATION TEST PROCEDURE lllLEr DETERMINATIONOF ClUADRANTPOWER TILTRAllO, M)SPM9. 10 NUMBER: 5UR-N7 ANS J.5 REFERENCE SECTIONSr 8.1.1 (10) Operator Conducted Surveillance on Safety-Related Equipment or Systems test DESCRIPllON certification willdemonstrat lhls the abilityof the simulator to support the operator conductedsurveNance procedure 3-OSMSR 10, Determination of Quadrant Power lilfRatio. With no malfunctions present, the ability to successfully perform this surveillance willbe verified.
OPIIONS TMs test can be performed at time in core life.
INITIALCONDITIONS RNAL CONDmONS MOL IR% power, steady state SurveIance complete.
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TUI~rY'OINTUNIT 3 SIMUILATORCERTIFICATION UPDATE NUMBER 1 APPENDIX A TEST ABSTRACTS A.2 TESTS CHANGED
OINT SIMUIATOR CERllFICAllON lEST PROCEDURE m1Ei CONTAINMENTEMERGENCY SYSTEMS OPERAllONS AND MAIFUNCllONS NUMBER: MCÃ-001 ANS 8.5 REFERENCE SECTIONSi 8. 14@8) Passive Malfunctions ln Engineered Safety Features Systems DESCRIPTION This test wfifexercise varfous malfunctions In the containment emergency systems. Proper system response to the malfunctions wl be verified. lhe containment emergency systems include the containment spray pumps, the emergency containment cooler fans, and the emergency containment filter fans.
OPllONS There are a wide vaifety of failures available in the containment emergency systems.
INITIALCONDITIONS FINAL CONDmONS MOL stead'y state at 100% power Each run willcontfnue until the proper response has been verifie.
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INTSIMUIATOR CERllFICAllONTEST PROCEDURE TITLEs COMPONENT COOLING WATER OPERAllONS AND MALFUNCllONS NUMBER MCS40l ANS 8.5 REFERENCE SECllONS: 8.1.2 (8) Loss of Component Cooling System DESCRIPllON This test weal verify the simulators response to a malfunction of the Component Cooling Water system. AilCCW pumps willbe tripped resuNng In a total loss of CCW cooling.
OPllONS lhere are several different means to cause a loss of CCW.
INITIALCONDITIONS FINAL CONDITIONS Steady state 100% power. The test wI run for 20 minutes after the loss of CCW.
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INT SIMULATOR CENlFICAllONTEST PROCEDURE TTTLEr TllRBINE PlANT COOLING WATER SySTEM OPERAllONS AND MAIFUNCTTONS NUMBER. MCS-ON ANS 3.5 REFERENCE SECTTONS: 8.1.2g) Loss of Service Water or Cooling fo Individual Components DESCRIPllON This test willverify the simulators response to a malfunction of the Turbine Plant Cooing Water system. AIITPCW pumps wi7I be tripped resulting in a total loss of TPCW cooling.
OPllONS There are several different means to cause a loss of Turbine Plant Cooling Water.
INIllALCONDITIONS FINAL CONDlllONS MOL steady state at 100% power The test willbe stopped 30 minutes after the Initiation of the event.
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OINT SIMULATOR CERTIRCATION lEST PROCEDURE TIILEr LETDONV AND VOLUME CONlROL TANK5 NSIEM OPERAllONS AND MALFUNCTIONS NUMBER MCV~
ANS 3.5 REFERENCE SECllONS: 3.1.2 (18) Failure of Reactor Cookrnt Pressure and Volume Control Systems DESCRIPllON The test checks the response of the Letdown and Volume Control Tank porttonsof the CVCS system. Various malfunctions whichaffect these systems wN be Initiated to verifyproper system response. A total of four different malfunction tests wN be run: (1) The letdown control valve PCV-145 wiN be failed open, g PCV-145 wN be fai7ed closed, (3) The VCT level control valve LCV1 15A wN be faBed to the chert position. and (4) The letdown Isolation valve CV 204 wN be failed closed.
OPllONS lhere are numerous malfunctions which can be run on the Letdown and Volume Control Tank systems. Representative malfunctions should be chosen to exercise as many parts of the systems as possible.
INlllALCONDITIONS FINAL CONDITIONS IRK power, normal letdown lineup. Terminate each run after system parametershave stabII!zedor trends are clearly evktent.
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INT SIMUlATOR CERllFICATION TEST PROCEDURE TITLE: LOSS OF NORMAL FEEDWATER NUMBER: MFW~
ANS 3.5 REFERENCE SECllONS: 3.1.2(f9 Loss of Normal Feedwater or Feedwater System Failure DESCRIPllON The purpose of this cerNffcaNon test Is to examine the simulator response to a loss of normal feedwater. This loss of normal feedwater transient willbe compared to a best estfmate analysis usfng the Turkey Point RHPAN model. As such, no operator octions wiN be taken during the course of the event and several assumptions have been mode to make the imufotor and the RETRAN models consistent. Since the RElRAN model does not Include charging and letdown models, these paths willbe Isolated In the simulator. The transient willbe inltfated by tripping open both feedwater pump motor breakers. lhe turbine runback that would normally result hom the tripping of these breakers is blocked. AMSAC willalso be blocked. AIIcontrol systems ore In automatic except the control rods.
OPllONS The main feedwater can be lost via a variety of mechanisms Including the fling closed of the IsolaNon or regulalfon valves, pump bearing failures, and motor breaker failures.
INmAL CONDITIONS FINAL CONDmONS Steady state at 100% power. BOL, EquiTbifum xenon The test wfffrun for 20 minutes. By that Nme the steam generator levels should be recovering steadily and the system approaching a stable hot standby condiNon.
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OINT SIMUIATOR CERTIFICAllONTEST PROCEDURE TITLE: MAXIMUMRATE POWER RAMP (100% to 75% and back to 100%)
NUMBER: MRX-001 ANS 8.5 REFERENCE SECllONS: B.22 PJ Maximum rate Power Ramp (100% down to 75% and back up to 100%)
DESCRIPTION TNs certilicatlon test wi7l evaluate the ablfity of the simulator to perform a rapid decrease In power from 100% power to 75% and return back to 100%. Although this Is an Appendix B test, due to the nature of this test manual actions have to be token. The off-normal operating procedure 3NOP-100, Fast Load Reduction wi be used to rapidly reduce power to approximately 75%. After the plant has stablRzed, a return to power at the maximum rate possible wIoccur. lhe test team wi8 pull control rods and di7ute the PCS while picking up load on the turbine. Ifrequired, extra letdown ortflces and charging pumps wN be used to dPute. During the return fo power average temperature wIIIbe closely matched with the reference temperature and the al Emits willbe obseived.
OPllONS INlllALCONDITIONS RNAL CONDITIONS Steady state, 100% power, MOL 100% power after recovery.
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INTSIMUIATOR CERTIFICATION TEST PROCEDURE llRE: MAINSlEAM LINE BR&LKINSIDE CONTAINMENT
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MSG-001 ANS 8.5 REFERENCE SECllONS: 3.1.2I20) Main Steam Line as well as Main Feed One Breaks (Both Inskfe and outside Containment)
B 22 (9) Maximum SIze Unlsolabie Main Steam Line Rupture DESCRIPllON This steam fne break transient wIIIbe compared to a best estimate analysis using the Turkey Point RETRAN model. As such, the test Is not intended to followin detai7 the emergency operating procedures covering this type of transient. However, operator actions to turn offthe reactor coolant pumps on low su&cooling margin and isolate the auxiliary feedwater to the affected steam generator have been programmed into the scenario. No other operator actions wIIIbe taken during the course of the event. Several assumptions have been mode In order to make the simulator and the REI PAN model consistent. Since the RETRAN model does not Include charging and letdown models or accumulatoo, these paths wi7I be isolated ln the simulator by the scenario. Allother control and safety systems will be in automatic and fully functional. A steam line break equivalent to the area of the flow restrictor at the steam generator outlet Is assumed to occur in the B steam line inside containment. Since this an ANS 3.5 Appendix B transient no operator actions willbe allowed after the transient starts. However, as mentioned above, the reactor coolant pumps willbe stopped when there is an Indication that Sl Is occurring with a low su&cooling margin. lhe aum7iary feedwater willalso be Isolated to the affected steam generator.
OPllONS The simulator is capable of simulating steam fine breaks of any dze at several locations Inside and outside contairvnent on each of the steam lines.
INITIALCONDmONS FINAL CONDlllONS 1006'ower steady state. EOL lhe fest wiIIrun for 10 minutes.
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lllLE: SIMULTANEOUS CLOSURE OF ALL MSIYS NUMBER: MSG~
ANS 8.5 REFERENCE SECllONS: B22P) Simultaneous Closure of AllMSIYs DESCRIPllON This test examines the simulator response to the simultaneous closure of all of the main steam Ene Isolation valves (MSIYs). AN control and protection systems wm be In automatic. Since this is an ANS 3.5 Appendor B test, no operator follow up actions wi8 be taken OPTIONS Any or all of the MSIV's can be closed by a variety of failure mechanisms. These include giving the valves a fail close signal or using an instructor override on the control board handswitches.
INmAL CONDmONS RNAL CONDIllONS 100% power steady state. BOL The transksnt Is analyzed for approxlmatet'y 10 minutes.
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