ML20117C766

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Forwards Four Yr Rept for CNS Unit One Simulator Per 10CFR55.45 b.5.ii & b.5.vi
ML20117C766
Person / Time
Site: Catawba Duke Energy icon.png
Issue date: 08/21/1996
From: Miller W
DUKE POWER CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9608280036
Download: ML20117C766 (14)
v • d • e


Text

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DukeIbuer Compans (803)831-3000 Cataicba Training Center 4850 ConcordRoad '

fork, SC 29745 DUKEPOWER August 21,1996 U.S. Nuclear Regulatory Commission Washington D.C. 20555 Attention: Document Control Desk

SUBJECT:

Catawba Nuclear Station Simulator Four Year Report This report is submitted as the four year report for the Catawba Nuclear Station Unit One Simulator in accordance with 10CFR55.45 b.5.ii and b.5.vi. The required information is included in the form of attachments to this letter as follows:

ATTACHMENT 1.0 UNCORRECTED PERFORMANCE TEST FAILURES ATTACHMENT

2.0 DESCRIPTION

OF PERFORMANCE TESTS COMPLETED (1992 -1995)

ATTACHMENT 3.0 NEXT FOUR YEAR TEST SCHEDULE (1996 - 1999)

ATTACHMENT 4.0 SIMULATOR VS PLANT DIFFERENCES ATTACHMENT 5.0 ACRONYM LIST If there are any questions concerning this report, please feel free to call Glenn Spurlin (803)-831-3117.

$$$$hO W. H. Miller' Training Manager Catawba Nuclear Station j

cc:

W.R. McCollum G.R. Peterson 9608280036 960821 PDR ADOCK 05000413 R

PDR ama on -wes =,

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ATTACHMENT 1.0 Uncorrected Performance Test Failures 1

The Catawba Nuclear Station Unit One Simulator has no unresolved performance test failures. There are three outstanding work requests (request for correction of the problem) associated with transient testing from 1995. These work requests were generated to track resolution of differences between tranr:ent reference data supplied by our Duke Power Safety Analysis Group and by sirruatet performance. These differences are a result of replacement steam generator modeling on the simulator which was completed in 1995 due to replacement of our Unit 1 Steam Generators. New j

data for our transients was generated and compared to simulator response. The outstanding work requests are:

AFW-045 Written to track differences in cooldown effect of Auxiliary Feedwater (CA) on the system. This is a generic issue on all transients where CA is used.

Target date for completion is 12/31/96.

RCS-028 Tracks resolution of differences in Core Exit thermocouple and Reactor Coolant system pressure response on a Small Break LOCA with inadequate Core Cooling (Transient 15). Target date for completion is 12/31/96.

RCS-029 Tracks resolution of differences in Tavg and Reactor Power during a loss of all feedwater with an ATWS (Transient 14). Simulator parameters do not decrease as quickly as reference data. Target date for completion is 12/31/96.

The direction of each of the parameters tracked above is correct. The magnitude of the change is not significant enough to detract from training.

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ATTACHMENT 2.0.

Description of Tests Completed - 1992 Transients:

Comoletion Date

  1. 1 SGTR 12/09/92
  1. 2 LOCA 12/15/92
  1. 3 Sm Brk LOCA 12/09/92
  1. 4 Loss Of Offsite Power 12/14/92
  1. 5 NCP Trip 09/24/92
  1. 6 Loss of CF 11/30/92
  1. 7 Loss of FDW 12/09/92
  1. 8 Dropped Rod BOL 09/24/92
  1. 9 Load Rejection 12/14/92
  1. 10 Spray Valve Failure 11/30/92
  1. 11 Rx Trip 12/09/92
  1. 12 Steam Line Break 12/09/92
  1. 13 Feed Line Break 12/09/92
  1. 14 Loss CF with ATWS 12/14/92
  1. 15 Small Break LOCA 04/05/93
  1. 16 SGTR, Failed PORV 12/14/92

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Malfunctions i

NDX-4 ND Pump Discharge Rupture 11/20/92 EPL-2 Loss of Vital Channel 04/06/93 KCX-2 KC Relief Valve Failure 11/2092 IRX-10 Stuck Rod 11/20/92 CHX-1 Variable NC Activity 09/24/92 EGB-2 Main generator Breaker Trip 11/20/92 ILE-1 PZR Level Master Failure 11/20/92 NCX-2 PZR PORV Failure 11/20/92 CFX-7 Main Feedline Break 04/05/93 Others Steady State Normal Ops Test 04/05/93 (Four Power Levels including the one hour test at 100% operations.)

Simulator real Time Test 12/29/92

ATTACHMENT 2.0 (cont.)

Description of Tests Completed - 1993 Transients:

Comotetion Date

  1. 1 SGTR 03/24/94
  1. 2 LOCA 03/24/94
  1. 3 Sm Brk LOCA 03/24/94
  1. 4 Loss Of Offsite Power 03/24/94
  1. 5 NCP Trip 03/24/94
  1. 6 Loss of CF 03/24/94
  1. 7 Loss of FDW 03/24/94
  1. 8 Dropped Rod BOL 03/29/94
  1. 9 Load Rejection 11/30/93
  1. 10 Spray Valve Failure 11/30/93
  1. 11 Rx Trip 11/30/93
  1. 12 Steam Line Break 11/30/93
  1. 13 Feed Line Break 12/02/93
  1. 14 Loss CF with ATWS 12/03/93
  1. 15 Small Break LOCA 12/13/93
  1. 16 SGTR, Failed PORV 12/09/93 Malfunctions SGX-1 SGTR 04/12/93 NCX-7 NC Leak 03/31/94 EPX-2 Loss of Switchyard 11/03/93 DGX-1 DG Failure to Start 04/12/93 RNX-3 RN Pump Failure 04/12/93 CFX-1 Loss of FWPT Vacuum 03/31/94 IRX-6 Dropped Rod 11/03/93 IRX-9 Rods Fail to Move 11/03/93 IRX-3 Uncontrolled Rod Motion 04/12/93 SMX-7 Steam Line Break inside Containment 11/03/93 Others Steady State Normal Ops Test 03/31/94 (Four Power Levels including the one hour test at 100% operations.)

Simulator real Time Test 01/09/94

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i ATTACHMENT 2.0 (cont.)

Description of Tests Completed - 1994 1

. Transients:

Comoletion Date j.

  1. 1

?SGTR 02/02/95 i

  1. 2 LOCA~

02/02/95 j

'#3 Sm Brk LOCA 02/02/95

  1. 4 Loss Of Offsite Power 02/02/95
  1. 5 NCP Trip 02/02/95
  1. 6 Loss of CF 02/02/95
  1. 7 Loss of FDW 03/16/95
  1. 8 Dropped Rod BOL 02/02/95 j
  1. 9 Load Rejection 02/02/95 i
  1. 10 Spray Valve Failure 02/20/95 j
  1. 11 Rx Trip

.02/02/95

  1. 12 Steam Line Break 02/06/95
  1. 13 Feed Line Break 02/06/95 l
  1. 14 Loss CF with ATWS 02/02/95
  1. 15 Small Break LOCA 03/16/95
  1. 16 SGTR, Failed PORV 02/06/95 i

I Malfunctions

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NCX-8 LOCA 01/04/95 VlX-1 Loss of VI 01/04/95 1

l EPX-8 Loss of 4160 VAC Bus 01/04/95 NCP-1 NCP Trip 12/01/94 IRX-13

. Improper Overlap on Nls 12/01/94 EHC-1 Inadvertent Turbine Trip 12/01/94 ENB-11 Power Range Detector Failure 12/01/94 IPE-1 Pzr Master Pressure Controller Failure 12/01/94 IPX-1 Auto Rx Trip Failure 01/04/95 i

Others Steady State Normal Ops Test 02/06/95.

1 (Four Power Levels including the one hour test at 100% operations.)

Simulator real Time Test 12/29/94

ATTACHMENT 2.0 (cont.)

Description of Tests Completed - 1995 Transients:

Comoletion Date

  1. 1 SGTR 01/25/96
  1. 2 LOCA 03/13/96
  1. 3 Sm Brk LOCA 03/13/96
  1. 4.

Loss Of Offsite Power 03/13/96

  1. 5 NCP Trip 12/04/96
  1. 6 Loss of CF 03/13/96
  1. 7 Loss of FDW 03/07/96 i
  1. 8 Dropped Rod BOL and EOL 01/25/96
  1. 9 Load Rejection 03/07/96
  1. 10 Prz PORV Failure 01/09/96
  1. 11 Rx Trip 01/29/96
  1. 12 Steam Line Break 01/27/96
  1. 13 Feed Line Break 03/06/96
  1. 14 Loss CF with ATWS 03/12/96 i
  1. 15 Small Break LOCA 03/12/96
  1. 16 SGTR, Failed PORV:

This transient was deleted and will be replaced with a Hot Leg LOCA in 1996. The SGTR, failed PORV provided no useful data for simulator fidelity.

Malfunctions NCX-3 Pzr Safety Failure 11/01/95 EPX-5 Loss of 6900 V Switchgear 11/01/95 MTX-3 Loss of Cond Vacuum 11/01/95 NDX-1 ND Pump Failure 11/01/95 MLT-2 Loss of All Feedwater-ran Transient 6 for this test.

EPL-4 Loss of Protective System Channel 01/02/96 IRE-1 Process instrument and Control Failure 11/01/95 NIX-4 ECCS Header Valve Fail to Open on ECCS 11/01/95 IRX-2 Uncontrolled Rod Withdrawal 12/05/95 Others Steady State Normal Ops Test 01/04/96 **

(Four Power Levels including the one hour test at 100% operations.)

Simulator real Time Test Test not run.

    • The 100% test was run satisfactorily against predicted data. Other tests were not run due to S/G replacement project. No new data will be available for steady state tests until after Unit 1 startup.

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ATTACHMENT 3.0 Annual Simulator Testing Schedule for 1996 Transients:

Transient #1 S/G Tube Leak Transient #2 Large Break LOCA

. Transient #3 Small Break LOCA Transient #4 Loss of Offsite Power Transient #5 Single NC Pump Trip Transient #6 Loss of Main Feedwater Transient #7 Loss of All Feedwater Transient #8 Dropped Rod (EOL and BOL)

Transient #9 Load Rejection Transient #10 PORV Failure Transient #11 Reactor Trip Transient #12 Steam Line Break Transient #13 Feedwater Line Break Transient #14 Loss of CF ATWS Transient #15 Small Break LOCA ICC Transient #16 Hot Leg LOCA Malfunctions:

NDX4 ND Pump Discharge Rupture EPL2 Loss of Vital Channel KCX2 KC Relief Valve Failure

-IRX10 Stuck Rod i

l CHX1 Variable NC Activity i

EGB2 Main Generator Breaker Trip ILE1.

PZR Level Master Failure NCX2 PZR PORV Failure CFX7 Main Feedline Break i

Others:

Steady State / Normal Ops Test -

(Four Power levels, including the one hour at 100% normal operations)

Simulator Real Time Test

ATTACHMENT 3.0 (cont 3 Annual Simulator Testing Schedule For 1997 Transients:

Transient #1 S/G Tube Leak Transient #2 Large Break LOCA Transient #3 Small Break LOCA Transient #4 Loss of Offsite Power Transient #5 Single NC Pump Trip Transient #6 Loss of Main Feedwater Transient #7 Loss of All Feedwater l

Transient #8 Dropped Rod (EOL and BOL)

Transient #9 Load Rejection Transient #10 PORV Failure Transient #11 Reactor Trip -

Transient #12 Steam Line Break Transient #13 Feedwater Line Break Transient #14 Loss of CF ATWS 1

Transient #15 Small Break LOCA ICC Transient #16 Hot Leg LOCA Malfunctions:

SGX1 SGTR NCX7 NC Leak j

EPX2 Loss of Switchyard DGX1 D/G Failure to Start RNX3 RN Pump Failure CFX1 Loss of FWPT Vacuum IRX6 Dropped Rod IRX9 Rods Fail to Move IRX3 Uncontrolled Rod Motion SMX7 Steam Line Break Inside Containment Others:

Steady State / Normal Ops Test (Four Power levels, including the one hour at 100% normal operations)

Simulator Real Time Test

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ATTACHMENT 3.0 (cont.)

Annual Simulator Testing Schedule for 1998 l

Transients:

l Transient #1 S/G Tube Leak Transient #2 Large Break LOCA Transient #3 Small Break LOCA Transient #4 Loss of Offsite Power Transient #5 Single NC Pump Trip Transient #6 Loss of Main Feedwater i

Transient #7 Loss of All Feedwater 1

Transient #8 Dropped Rod (EOL and BOL)

Transient #9 Load Rejection-Transient #10 Porv Failure Transient #11 Reactor Trip -

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Transient #12 Steam Line Break Transient #13 Feedwater Line Break Transient #14 Loss of CF ATWS Transient #15 Small Break LOCA ICC Transient #16 Hot Leg LOCA Malfunctions:

NCX8 LOCA VlX1 Loss of VI EPX8 Loss of 4160 VAC Bus NCP1 NCP Trip IRX13 Improper Overlap EHC1 Inadvertent Turbine Trip ENB11 Power Range Det. Failure IPE1 PZR Master Pressure Controller Failure IPX1 Auto Reactor Trip Failure Others:

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Steady State / Normal Ops Test (Four Power levels, including the one hour at 100% normal operations)

Simulator Real Time Test I

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ATTACHMENT 3.0 (cont.)

Annual Simulator Testing Schedule for 1999

' Transients:

Transient #1 S/G Tube Leak Transient #2 Large Break LOCA Transient #3 Small Break LOCA Transient #4 Loss of Offsite Power Transient #5 Single NC Pump Trip Transient #6 Loss of Main Feedwater Transient #7 Loss of All Feedwater Transient #8 Dropped Rod (EOL and BOL)

Transient #9 Load Rejection Transient #10 PORV Failure Transient #11 Reactor Trip Transient #12 Steam Line Break Transient #13 Feedwater Line Break Transient #14 Loss of CF A1WS Transient #15-Small Break LOCA ICC Transiant #16 -

Hot Leg LOCA Malfunctions:

i NCX3 PZR Safety Failure EPX5 Loss of 6900 Switchgear j

MTX3' Loss of Cond. Vacuum NDX1

'ND Pump Failure MLT2 Loss of All Feedwater EPL4 Loss of Protective System Channel j

IRE 1 Process Instrument and Control Failure NIX 4 ECCS Header Valves Fail to Open on ECCS Activation IRX2 Uncontrolled Rod Withdrawal

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Others:

Steady State / Normal Ops Test (Four Power levels, including the one hour at 100% normal operations)

Simulator Real Time Test i

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ATTACHMENT 4.0 Simulator vs Plant Differences The following items are changes made to the simulator since the last four year certification. These items do not reflect true plant conditions in either size, layout, environmental surroundings, or actual operation. Each change has been evaluated and determined to be beneficial to training. No negative impacts have been caused by the differences, i

.1. A new D/G simulator has been added to the simulator. The panels for the D/G are the same size, design and layout as the plant. A difference does exists in the environment surrounding the panels. The simulator panels do not reflect the true temperature, noise level and ventilation effects experienced at the local panels.

2.' A small panel has been added to allow operation and indication of the Unit 2 Nuclear Service Water pumps. This panel does not reflect the size or location of the actual Unit 2 control boards in the plant. This panel was installed at the request of Operations to give the operators quick information on the operation of these pumps. Before installation of the panels, operators were required to call the booth for status of the pumps.

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3. The Loose Parts Monitor System has been replaced in the plant. This modification will not be done on the simulator. The existing simulator panel is face front only and provides no training benefit. This panel will be deleted from the simulator.

ATTACHMENT 5.0 Catawba System Designators System Designator System Nomenclature AS Auxiliary Steam BB Steam Generator Blowdown Recycle BW Steam Generator Wet Lay-Up CA Auxiliary Feedwater CF Main Feedwater CL CF Pump Turbine Condensate Seal CM Condensate CS Condensate Sarnpling Electrical Systems (3 letter)

EBA 230KV Switchyard EBE 230KV Switchyard Control EDA Control Rod Position Indication EEA EnvironmentalInstrumentation EEB MeteorologicalInstrumentation EFA Fire Detection EGB Generator Excitation EMA Emergency Safeguards Feature,1.47 Bypass EMB Annunciators, Status and Monitor Lights EMC Event Recorder EMD Loose Parts Monitoring ENA incore Instrumentation ENB Excore Instrumentation ENC Boron Dilution Mitigation EPA Unit Main Power EPB 6.9KV Unit Normal Auxiliary Power EPC 4.16KV Essential Auxiliary Power EPD 600VAC Unit Normal Auxiliary Power EPE 600VAC Essential Auxiliary Power EPF 240/120VAC Auxiliary Control Power EPG 120VAC Vital Instrument and Control Power EPJ 250VAC Auxiliary Power EPK 125VAC Auxiliary Control Power EPL 125VAC Vital Instrument and Control Power EPM 13.8KV Auxiliary Power EQB Diesel Generator Load Sequencer EOC Diesel Generator

ATTACHMENT 5.0 (cont 3 Catawba System Designators System Designator System Nomenclature EOD Standby Shutdown Facility Diesel Generator Mechanical Systems FD D/G Engine Fuel Oil GH Generation Hydrogen HA Bleed Steam to 'A' Heaters HB Bleed Steam to 'B' Heaters HC Bleed Steam to 'C' Heaters HD Bleed Steam to 'D' Heaters i

HE Bleed Steam to 'E' Heaters HF Bleed Steam to 'G' Heaters HM Moisture Separator Reheater Bleed Steam q

ICCM Inadequate Core Cooling Monitor IDE Steam Dump Control IFE Feedwater Control IKE Operator Aid Computer ILE Pressurizer Pressure & Level Control IPE Reactor Protection IRE Rod Control ISE ESF Actuation ITE Main Turbine Instrumentation & Control (l&C)

IWE Feedwater Pump Turbine I & C KC Component Cooling KD D/G Engine Cooling Water KF Spent Fuel Cooling KG Generator Stator Cooling Water KR Recirculated Cooling Water LD D/G Engine Lube Oil LF CF Pump Turbine Lube Oil LG Generator Seal Oil LH Main Generator Hydraulic Oil LP CF Pump Turbine Hydraulic Oil LT Main Turbine Lube Oil NC Reactor Coolant ND Residual Heat Removal NF Ice Condenser Refrigeration NI Safety injection

ATTACHMENT 5.0 (cont.)

Catawba System Designators

- System Desianator System Nomenclature NS Containment Spray NV Chemical and Volume Control NW

' Containment isolation Valve injection Water RC Condenser Circulating Water RL Conventional Low Pressure Service Water RN Nuclear Service Water SM Main Steam SP Main Steam Supply to CF Pump Turbines SV Main Steam to Atmosphere VC Control Area HVAC VE Annulus Ventilation VF Fuel Pool Ventilation i

VI Instrument Air

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VP Containment Purge VO Containment Air Release and Addition VS Station Air W

Containment Ventilation VX Containment Air Return & Hydrogen Reduction WG Waste Gas WL Liquid Waste Recycle

'YC Control Area Chilled Water l

ZM Main Vacuum i

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