ML12011A169
| ML12011A169 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 12/14/2011 |
| From: | Energy Northwest |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| GO2-11-201 | |
| Download: ML12011A169 (347) | |
Text
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Chapter 15 ACCIDENT ANALYSES
TABLE OF CONTENTS
Section Page LDCN-08-035 15-i 15.0 GENERAL.................................................................................15.0-1 15.0.1 ANALYTICAL OBJECTIVE........................................................15.0-3 15.0.2 ANALYTICAL CATEGORIES.....................................................15.0-3 15.0.2.1 Single Loop Operation (SLO).....................................................15.0-4 15.0.3 EVENT E VALUATION..............................................................15.0-5 15.0.3.1 Identification of Causes and Frequency Classification........................15.0-5 15.0.3.1.1 Unacceptable Results for Incidents of Mode rate Frequency [Anticipated (Expected) Oper ational Transients]............................15.0-6 15.0.3.1.2 Unacceptable Results for Infrequent Incidents
[Abnormal (Unexpected) Opera tional Transients]...........................15.0-6 15.0.3.1.3 Unacceptable Results for Limiting Faults [Design-Basis (Postulated) A ccidents]...........................................................15.0-6 15.0.3.2 Sequence of Even ts and Systems Operation.....................................15.0-7 15.0.3.2.1 Single Failures or Operator Errors.............................................
15.0-8 15.0.3.2.1.1 General...........................................................................15.0-8 15.0.3.2.1.2 Initiating Event Analysis.......................................................15.0-8 15.0.3.2.1.3 Single Active Comp onent Failure or Single Operator Error Analysis...................................................................15.
0-9 15.0.3.3 Core and System Performance....................................................15.0-9 15.0.3.3.1 Mathematical Model..............................................................15.0-10 15.0.3.3.2 Input Parameters and Initial Conditions for Analyzed Events.............
15.0-11 15.0.3.3.3 Considerati on of Uncertainties..................................................15.0-11 15.0.3.3.3.1 Core Flow Uncertainty Analysis.............................................15.
0-12 15.0.3.3.4 Results...............................................................................
15.0-13 15.0.3.4 Barrier Performance.................................................................15.0-14 15.0.3.5 Radiol ogical Consequences........................................................15.
0-14 15.
0.4 REFERENCES
.........................................................................
15.0-14 15.1 DECREASE IN REACTOR COOLANT TEMPERATURE.....................15.1-1 15.1.1 LOSS OF FEED WATER HEATING..............................................15.1-1 15.1.1.1 Identification of Causes and Frequency Classification........................15.1-1 15.1.1.1.1 Identification of Causes..........................................................15.
1-1 15.1.1.1.2 Frequency Classification.........................................................15.1-1 15.1.1.2 Sequence of Even ts and Systems Operation.....................................15.1-1 15.1.1.2.1 The Effect of Single Failures and Operat or Errors..........................15.1-2 C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 Chapter 15 ACCIDENT ANALYSES
TABLE OF CONTENTS (Continued)
Section Page LDCN-10-029 15-ii 15.1.1.3 Core and Syst em Performa nce ....................................................
15.1-2 15.1.1.3.1 Mathematical Model .............................................................. 15.1-2 15.1.1.3.2 Input Parameters and Initia l Conditions ....................................... 15.1-2 15.1.1.3.3 Results ............................................................................... 15.1-3 15.1.1.3.4 Considerations of Uncertainties ................................................. 15.1-3 15.1.1.4 Barrier Performance ................................................................. 15.1-3 15.1.1.5 Radiological Consequences ........................................................ 15.1-3 15.1.2 FEEDWATER CONTROLLER FAILURE - MAXIMUM DEMAND ...... 15.1-3 15.1.2.1 Identification of Causes and Frequency Classification ........................ 15.1-3 15.1.2.1.1 Identification of Causes .......................................................... 15.1-3 15.1.2.1.2 Frequency Classification ......................................................... 15.1-3 15.1.2.2 Sequence of Events and Systems Operation ..................................... 15.1-4 15.1.2.2.1 Sequence of Even ts and Systems Operation - Single Loop Operation ... 15.1-4 15.1.2.2.2 The Effect of Single Fa ilures and Operator Errors .......................... 15.1-5 15.1.2.3 Core and Syst em Performa nce ....................................................
15.1-5 15.1.2.3.1 Mathematical Model .............................................................. 15.1-5 15.1.2.3.2 Input Parameters and Initia l Conditions ....................................... 15.1-6 15.1.2.3.3 Results ............................................................................... 15.1-6 15.1.2.3.4 Consideration of Uncertainties .................................................. 15.1-7 15.1.2.4 Barrier Performance ................................................................. 15.1-7 15.1.2.5 Radiological Consequences ........................................................ 15.1-7 15.1.3 PRESSURE REGULATO R FAILURE - OPEN ................................. 15.1-7 15.1.3.1 Identification of Causes and Frequency Classification ........................ 15.1-7 15.1.3.1.1 Identification of Causes .......................................................... 15.1-7 15.1.3.1.2 Frequency Classification ......................................................... 15.1-7 15.1.3.2 Sequence of Events and Systems Operation ..................................... 15.1-8 15.1.3.2.1 Sequence of Events ............................................................... 15.1-8 15.1.3.2.2 Systems Operation................................................................. 15.1-8 15.1.3.2.3 The Effect of Single Fa ilures and Operator Errors .......................... 15.1-8 15.1.3.3 Core and Syst em Performa nce ....................................................
15.1-9 15.1.3.3.1 Mathematical Model .............................................................. 15.1-9 15.1.3.3.2 Input Parameters and Initia l Conditions ....................................... 15.1-9 15.1.3.3.3 Results ............................................................................... 15.1-9 15.1.3.3.4 Consideration of Uncertainties .................................................. 15.1-9 15.1.3.4 Barrier Performance
................................................................. 15.1-10 15.1.3.5 Radiological Consequences ........................................................ 15.1-10 C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 Chapter 15 ACCIDENT ANALYSES
TABLE OF CONTENTS (Continued)
Section Page LDCN-10-029 15-iii 15.1.4 INADVERTENT SAFETY/R ELIEF VALVE OPENING ..................... 15.1-10 15.1.4.1 Identification of Causes and Frequency Classification ........................ 15.1-10 15.1.4.1.1 Identificati on of Causes .......................................................... 15.1-10 15.1.4.1.2 Frequency Classification ......................................................... 15.1-10 15.1.4.2 Sequence of Events and Systems Operation ..................................... 15.1-11 15.1.4.2.1 Sequence of Events
............................................................... 15.1-11 15.1.4.2.2 Systems Operation................................................................. 15.1-11 15.1.4.2.3 The Effect of Single Fa ilures and Operator Errors .......................... 15.1-11 15.1.4.3 Core and Syst em Performanc e ....................................................
15.1-11 15.1.4.3.1 Mathematical Model
.............................................................. 15.1-11 15.1.4.3.2 Input Pa rameters and Initial Conditions
....................................... 15.1-11 15.1.4.3.3 Re sults ...............................................................................
15.1-12 15.1.4.4 Barrier Performance
................................................................. 15.1-12 15.1.4.5 Radiological Consequences ........................................................ 15.1-12 15.1.5 SPECTRUM OF STEAM PIPING FAILURES INSIDE AND OUTSIDE OF CONTAINMENT IN A PRESSURIZED WATER
REACTOR ..............................................................................
15.1-12 15.1.6 INADVERTENT RESIDUAL HEAT REMOVAL SHUTDOWN COOLING OPERATION ............................................................ 15.1-12 15.1.6.1 Identification of Causes and Frequency Classification ........................ 15.1-13 15.1.6.1.1 Identificati on of Causes .......................................................... 15.1-13 15.1.6.1.2 Frequency Classification ......................................................... 15.1-13 15.1.6.2 Sequence of Events and Systems Operation ..................................... 15.1-13 15.1.6.2.1 Sequence of Events
............................................................... 15.1-13 15.1.6.2.2 System Operation
.................................................................. 15.1-13 15.1.6.2.3 Effect of Single Failu res and Operator Action ...............................
15.1-14 15.1.6.3 Core and Syst em Performanc e ....................................................
15.1-14 15.1.6.4 Barrier Performance
................................................................. 15.1-14 15.1.6.5 Radiological Consequences ........................................................ 15.1-14 15.
1.7 REFERENCES
......................................................................... 15.1-14
15.2 INCREASE IN RE ACTOR PRESSURE ............................................. 15.2-1 15.2.1 PRESSURE REGULATOR FAILURE - CLOSED
.............................. 15.2-1 15.2.1.1 Identification of Causes and Frequency Classification ........................ 15.2-1 15.2.1.1.1 Identification of Causes .......................................................... 15.2-1 15.2.1.1.2 Frequency Classification ......................................................... 15.2-1 C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 Chapter 15
ACCIDENT ANALYSES
TABLE OF CONTENTS (Continued)
Section Page LDCN-06-057,07-011 15-iv 15.2.1.2 Sequence of Even ts and Systems Operation.....................................15.2-1 15.2.1.2.1 Sequen ce of Events...............................................................15.2-1 15.2.1.2.2 System s Operation.................................................................15.
2-2 15.2.1.2.3 The Effect of Single Failures and Operat or Errors..........................15.2-2 15.2.1.3 Core and System Performance....................................................15.2-2 15.2.1.3.1 Mathematical Model..............................................................15.2-2 15.2.1.3.2 Input Parameters and Initia l Conditions.......................................15.2-3 15.2.1.3.3 Results...............................................................................
15.2-3 15.2.1.3.4 Considerati on of Uncertainties..................................................15.2-3 15.2.1.4 Barrier Performance.................................................................15.2-3 15.2.1.5 Radiol ogical Consequences........................................................15.2-3 15.2.2 GENERATOR L OAD REJECTION...............................................15.2-4 15.2.2.1 Identification of Causes and Frequency Classification........................15.2-4 15.2.2.1.1 Identification of Causes..........................................................15.
2-4 15.2.2.1.2 Frequency Classification.........................................................15.2-4 15.2.2.1.2.1 Genera tor Load Rejection.....................................................15.2-4 15.2.2.1.2.2 Generator Load Rejection with Bypass Failure............................15.2-4 15.2.2.2 Sequence of Even ts and System Operation......................................15.2-4 15.2.2.2.1 Sequen ce of Events...............................................................15.2-5 15.2.2.2.1.1 Generator Lo ad Rejection - Turbine C ontrol Valve Fast Closure......15.2-5 15.2.2.2.1.2 Generator Load Rejection with Failure of Bypass........................15.2-5 15.2.2.2.2 System Operation..................................................................15.2-5 15.2.2.2.2.1 Generator Lo ad Rejection with Bypass.....................................15.2-5 15.2.2.2.2.2 Generator Load Rejection with Failure of Bypass........................15.2-5 15.2.2.2.3 The Effect of Single Failures and Operat or Errors..........................15.2-5 15.2.2.3 Core and System Performance....................................................15.2-6 15.2.2.3.1 Mathematical Model..............................................................15.2-6 15.2.2.3.1.1 Generator Lo ad Rejection with Bypass.....................................15.2-6 15.2.2.3.1.2 Generator Load Rejection with Bypass Failure............................15.2-6 15.2.2.3.2 Input Parameters and Initia l Conditions.......................................15.2-7 15.2.2.3.3 Results...............................................................................
15.2-7 15.2.2.3.3.1 Generator Lo ad Rejection with Bypass.....................................15.2-7 15.2.2.3.3.2 Generator Load Rejection with Failure of Bypass........................15.2-7 15.2.2.3.4 Considerati on of Uncertainties..................................................15.2-7 15.2.2.4 Barrier Performance.................................................................15.2-8 15.2.2.4.1 Generator Load Rejection........................................................
15.2-8 C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 Chapter 15 ACCIDENT ANALYSES
TABLE OF CONTENTS (Continued)
Section Page LDCN-10-029 15-v 15.2.2.4.2 Generator Load Rejecti on with Failure of Bypass ........................... 15.2-8 15.2.2.5 Radiological Consequences ........................................................ 15.2-8 15.2.3 TURBINE TRIP ........................................................................ 15.2-8 15.2.3.1 Identification of Causes and Frequency Classification ........................ 15.2-8 15.2.3.1.1 Identification of Causes .......................................................... 15.2-8 15.2.3.1.2 Frequency Classification ......................................................... 15.2-8 15.2.3.1.2.1 Turb ine Trip ..................................................................... 15.2-8 15.2.3.1.2.2 Turbine Trip w ith Failure of Bypass ........................................ 15.2-9 15.2.3.2 Sequence of Events and Systems Operation ..................................... 15.2-9 15.2.3.2.1 Sequence of Events ............................................................... 15.2-9 15.2.3.2.1.1 Turb ine Trip ..................................................................... 15.2-9 15.2.3.2.1.2 Turbine Trip w ith Failure of Bypass ........................................ 15.2-9 15.2.3.2.2 Systems Operation................................................................. 15.2-9 15.2.3.2.2.1 Turb ine Trip ..................................................................... 15.2-9 15.2.3.2.2.2 Turbine Trip w ith Failure of Bypass ........................................ 15.2-9 15.2.3.2.2.3 Turbine Trip at Low Po wer with Failure of Bypass ...................... 15.2-9 15.2.3.2.3 The Effect of Single Fa ilures and Operator Errors .......................... 15.2-10 15.2.3.2.3.1 Turbine Trips at Power Levels Greater Than 30% Nuclear Boiler Rated
...................................................................... 15.2-10 15.2.3.2.3.2 Turbine Trips at Power Levels Less Than 30% Nuclear Boiler Rated
...................................................................... 15.2-10 15.2.3.3 Core and Syst em Performanc e ....................................................
15.2-10 15.2.3.3.1 Mathematical Model
.............................................................. 15.2-10 15.2.3.3.1.1 Turbine Trip with Bypass ..................................................... 15.2-10 15.2.3.3.1.2 Turbine Trip with Bypass Failure ............................................ 15.2-10 15.2.3.3.2 Input Pa rameters and Initial Conditions
....................................... 15.2-10 15.2.3.3.3 Re sults ...............................................................................
15.2-11 15.2.3.3.3.1 Turb ine Trip ..................................................................... 15.2-11 15.2.3.3.3.2 Turbine Trip with Failure of By pass ........................................ 15.2-11 15.2.3.3.3.3 Turbine Trip with B ypass Valve Failure, Low Power ................... 15.2-11 15.2.3.3.4 Considerations of Uncertain ties .................................................
15.2-12 15.2.3.4 Barrier Performance
................................................................. 15.2-12 15.2.3.4.1 Turbine Trip
........................................................................ 15.2-12 15.2.3.4.2 Turbine Trip with Failure of B ypass ...........................................
15.2-12 15.2.3.4.2.1 Turbine Trip with Failure of Bypass at Low Power ......................
15.2-12 15.2.3.5 Radiological Consequences ........................................................ 15.2-12 C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 Chapter 15 ACCIDENT ANALYSES
TABLE OF CONTENTS (Continued)
Section Page LDCN-10-029 15-vi 15.2.4 MAIN STEAM LINE ISOLATION VALVE CLOSURES
..................... 15.2-12 15.2.4.1 Identification of Causes and Frequency Classification ........................ 15.2-13 15.2.4.1.1 Identificati on of Causes .......................................................... 15.2-13 15.2.4.1.2 Frequency Classification ......................................................... 15.2-13 15.2.4.1.2.1 Closure of All Main Steam Line Isolation Valves ........................ 15.2-13 15.2.4.1.2.2 Closure of One Main Steam Line Isolation Valve ........................
15.2-13 15.2.4.2 Sequence of Events and Systems Operation ..................................... 15.2-13 15.2.4.2.1 Sequence of Events
............................................................... 15.2-13 15.2.4.2.2 Systems Operation................................................................. 15.2-14 15.2.4.2.2.1 Closure of All Main Steam Line Isolation Valves ........................ 15.2-14 15.2.4.2.2.2 Closure of One Main Steam Line Isolation Valve ........................
15.2-14 15.2.4.2.3 The Effect of Single Fa ilures and Operator Errors .......................... 15.2-14 15.2.4.3 Core and Syst em Performanc e ....................................................
15.2-14 15.2.4.3.1 Mathematical Model
.............................................................. 15.2-14 15.2.4.3.2 Input Pa rameters and Initial Conditions
....................................... 15.2-15 15.2.4.3.3 Re sults ...............................................................................
15.2-15 15.2.4.3.3.1 Closure of All Main Steam Line Isolation Valves ........................ 15.2-15 15.2.4.3.3.2 Closure of One Main Steam Line Isolation Valve ........................
15.2-15 15.2.4.3.4 Considerations of Uncertain ties .................................................
15.2-16 15.2.4.4 Barrier Performance
................................................................. 15.2-16 15.2.4.4.1 Closure of All Main Steam Line Isolation Valves ........................... 15.2-16 15.2.4.4.2 Closure of One Main St eam Line Isolation Valve ...........................
15.2-16 15.2.4.5 Radiological Consequences ........................................................ 15.2-16 15.2.5 LOSS-OF-CONDE NSER VACUUM
.............................................. 15.2-17 15.2.5.1 Identification of Causes and Frequency Classification ........................ 15.2-17 15.2.5.1.1 Identificati on of Causes .......................................................... 15.2-17 15.2.5.1.2 Frequency Classification ......................................................... 15.2-17 15.2.5.2 Sequence of Events and Systems Operation ..................................... 15.2-17 15.2.5.2.1 Sequence of Events
............................................................... 15.2-17 15.2.5.2.2 Systems Operation................................................................. 15.2-17 15.2.5.2.3 The Effect of Single Fa ilures and Operator Errors .......................... 15.2-18 15.2.5.3 Core and Syst em Performanc e ....................................................
15.2-18 15.2.5.3.1 Mathematical Model
.............................................................. 15.2-18 15.2.5.3.2 Input Pa rameters and Initial Conditions
....................................... 15.2-18 15.2.5.3.3 Re sults ...............................................................................
15.2-18 15.2.5.3.4 Consideration of Uncertainties .................................................. 15.2-18 C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 Chapter 15 ACCIDENT ANALYSES
TABLE OF CONTENTS (Continued)
Section Page LDCN-10-029 15-vii 15.2.5.4 Barrier Performance
................................................................. 15.2-19 15.2.5.5 Radiological Consequences ........................................................ 15.2-19 15.2.6 LOSS OF ALTERNATIN G CURRENT POWER ............................... 15.2-20 15.2.6.1 Identification of Causes and Frequency Classification ........................ 15.2-20 15.2.6.1.1 Identificati on of Causes .......................................................... 15.2-20 15.2.6.1.1.1 Loss of Auxiliary Power Transformers ..................................... 15.2-20 15.2.6.1.1.2 Loss of All Grid Connections
................................................. 15.2-20 15.2.6.1.2 Frequency Classification ......................................................... 15.2-20 15.2.6.1.2.1 Loss of Auxiliary Power Transformers ..................................... 15.2-20 15.2.6.1.2.2 Loss of All Grid Connections
................................................. 15.2-20 15.2.6.2 Sequence of Events and Systems Operation ..................................... 15.2-20 15.2.6.2.1 Sequence of Events
............................................................... 15.2-20 15.2.6.2.1.1 Loss of Auxiliary Power Transformers ..................................... 15.2-20 15.2.6.2.1.2 Loss of All Grid Connections
................................................. 15.2-20 15.2.6.2.2 Systems Operation................................................................. 15.2-21 15.2.6.2.2.1 Loss of Auxiliary Power Transformers ..................................... 15.2-21 15.2.6.2.2.2 Loss of All Grid Connections
................................................. 15.2-21 15.2.6.2.3 The Effect of Single Fa ilures and Operator Errors .......................... 15.2-21 15.2.6.3 Core and Syst em Performanc e ....................................................
15.2-22 15.2.6.3.1 Mathematical Model
.............................................................. 15.2-22 15.2.6.3.2 Input Pa rameters and Initial Conditions
....................................... 15.2-22 15.2.6.3.2.1 Loss of Auxiliary Power Transformers ..................................... 15.2-22 15.2.6.3.2.2 Loss of All Grid Connections
................................................. 15.2-22 15.2.6.3.3 Re sults ...............................................................................
15.2-22 15.2.6.3.3.1 Loss of Auxiliary Power Transformers ..................................... 15.2-22 15.2.6.3.3.2 Loss of All Grid Connections
................................................. 15.2-23 15.2.6.3.4 Consideration of Uncertainties .................................................. 15.2-23 15.2.6.4 Barrier Performance
................................................................. 15.2-23 15.2.6.4.1 Loss of Auxiliary Power Transfor mers........................................
15.2-23 15.2.6.4.2 Loss of All Grid Connections ................................................... 15.2-24 15.2.6.5 Radiological Consequences ........................................................ 15.2-24 15.2.7 LOSS-OF-FEEDW ATER FLOW
................................................... 15.2-24 15.2.7.1 Identification of Causes and Frequency Classification ........................ 15.2-24 15.2.7.1.1 Identificati on of Causes .......................................................... 15.2-24 15.2.7.1.2 Frequency Classification ......................................................... 15.2-24 C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 Chapter 15 ACCIDENT ANALYSES
TABLE OF CONTENTS (Continued)
Section Page LDCN-10-029 15-viii 15.2.7.2 Sequence of Events and Systems Operation ..................................... 15.2-24 15.2.7.2.1 Sequence of Events
............................................................... 15.2-24 15.2.7.2.2 Systems Operation................................................................. 15.2-25 15.2.7.2.3 The Effect of Single Fa ilures and Operator Errors .......................... 15.2-25 15.2.7.3 Core and Syst em Performanc e ....................................................
15.2-25 15.2.7.3.1 Mathematical Model
.............................................................. 15.2-25 15.2.7.3.2 Input Pa rameters and Initial Conditions
....................................... 15.2-26 15.2.7.3.3 Re sults ...............................................................................
15.2-26 15.2.7.3.4 Consideration of Uncertainties .................................................. 15.2-26 15.2.7.4 Barrier Performance
................................................................. 15.2-26 15.2.7.5 Radiological Consequences ........................................................ 15.2-26 15.2.8 FEEDWATER LINE BREAK
....................................................... 15.2-26 15.2.9 FAILURE OF RESIDUAL HEAT REMOVAL SHUTDOWN COOLING ...............................................................................
15.2-27 15.2.9.1 Identification of Causes and Frequency Classification ........................ 15.2-27 15.2.9.1.1 Identificati on of Causes .......................................................... 15.2-27 15.2.9.1.2 Frequency Classification ......................................................... 15.2-27 15.2.9.2 Sequence of Events and Systems Operation ..................................... 15.2-27 15.2.9.2.1 Sequence of Events
............................................................... 15.2-27 15.2.9.2.1.1 Identification of Operator Actions ........................................... 15.2-28 15.2.9.2.2 Systems Operation................................................................. 15.2-28 15.2.9.2.3 The Effect of Single Fa ilures and Operator Errors .......................... 15.2-28 15.2.9.3 Core and Syst em Performanc e ....................................................
15.2-28 15.2.9.4 Results ................................................................................. 15.2-29 15.2.9.4.1 Full Power to Approximately 100 psig ........................................ 15.2-29 15.2.9.4.2 Approxima tely 100 psig to Cold Shutdown ................................... 15.2-30 15.2.9.4.3 Temperature Re sponse - 3462 MWt ........................................... 15.2-31 15.2.9.4.4 Temperature Re sponse - 3702 MWt ........................................... 15.2-32 15.2.9.5 Barrier Performance
................................................................. 15.2-32 15.2.9.6 Radiological Consequences ........................................................ 15.2-32 15.2.10 REFERE NCES ........................................................................
15.2-32 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE ............. 15.3-1 15.3.1 RECIRCULATIO N PUMP TRIP ................................................... 15.3-1 15.3.1.1 Identification of Causes and Frequency Classification ........................ 15.3-1 15.3.1.1.1 Identification of Causes .......................................................... 15.3-1 C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 Chapter 15
ACCIDENT ANALYSES
TABLE OF CONTENTS (Continued)
Section Page 15-ix 15.3.1.1.2 Frequency Classification.........................................................15.3-1 15.3.1.1.2.1 Trip of One Recirculation Pump.............................................15.3-1 15.3.1.1.2.2 Trip of Two Recirculation Pumps...........................................15.3-1 15.3.1.2 Sequence of Even ts and Systems Operation.....................................15.3-2 15.3.1.2.1 Sequen ce of Events...............................................................15.3-2 15.3.1.2.1.1 Trip of One Recirculation Pump.............................................15.3-2 15.3.1.2.1.2 Trip of Two Recirculation Pumps...........................................15.3-2 15.3.1.2.2 System s Operation.................................................................15.
3-2 15.3.1.2.2.1 Trip of One Recirculation Pump.............................................15.3-2 15.3.1.2.2.2 Trip of Two Recirculation Pumps...........................................15.3-2 15.3.1.2.3 The Effect of Single Failures and Operat or Errors..........................15.3-2 15.3.1.2.3.1 Trip of One Recirculation Pump.............................................15.3-2 15.3.1.2.3.2 Trip of Two Recirculation Pumps...........................................15.3-2 15.3.1.3 Core and System Performance....................................................15.3-2 15.3.1.3.1 Mathematical Model..............................................................15.3-2 15.3.1.3.2 Input Parameters and Initia l Conditions.......................................15.3-3 15.3.1.3.3 Results...............................................................................
15.3-3 15.3.1.3.3.1 Trip of One Recirculation Pump.............................................15.3-3 15.3.1.3.3.2 Trip of Two Recirculation Pumps...........................................15.3-3 15.3.1.3.4 Considerati on of Uncertainties..................................................15.3-3 15.3.1.4 Barrier Performance.................................................................15.3-3 15.3.1.4.1 Trip of One Recirculation Pump................................................
15.3-3 15.3.1.4.2 Trip of Two Recirculation Pumps..............................................
15.3-4 15.3.1.5 Radiol ogical Consequences........................................................15.3-4 15.3.2 RECIRCULATION FLOW CONTROL FAILURE - DECREASING FLOW...................................................................................
15.3-4 15.3.2.1 Identification of Causes and Frequency Classification........................15.3-4 15.3.2.1.1 Identification of Causes..........................................................15.
3-4 15.3.2.1.2 Frequency Classification.........................................................15.3-4 15.3.2.2 Sequence of Even ts and Systems Operation.....................................15.3-4 15.3.2.2.1 Sequen ce of Events...............................................................15.3-4 15.3.2.2.1.1 Speed Decrease of One Recirculation Pump...............................15.3-4 15.3.2.2.1.2 Speed Decrease of Two Recirculation Pumps.............................15.3-4 15.3.2.2.2 System s Operation.................................................................15.
3-5 15.3.2.2.2.1 Speed Decrease of One Recirculation Pump...............................15.3-5 15.3.2.2.2.2 Speed Decrease of Two Recirculation Pumps.............................15.3-5 C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Chapter 15 ACCIDENT ANALYSES
TABLE OF CONTENTS (Continued)
Section Page LDCN-08-035 15-x 15.3.2.2.3 The Effect of Single Failures and Operat or Errors..........................15.3-5 15.3.2.3 Core and System Performance....................................................15.3-5 15.3.2.3.1 Mathematical Model..............................................................15.3-5 15.3.2.3.2 Input Parameters and Initia l Conditions.......................................15.3-5 15.3.2.3.2.1 Speed Decrease of One Recirculation Pump...............................15.3-5 15.3.2.3.2.2 Speed Decrease of Two Recirculation Pumps.............................15.3-6 15.3.2.3.3 Results...............................................................................
15.3-6 15.3.2.3.3.1 Speed Decrease of One Recirculation Pump...............................15.3-6 15.3.2.3.3.2 Speed Decrease of Two Recirculation Pumps.............................15.3-6 15.3.2.3.4 Considerati on of Uncertainties..................................................15.3-6 15.3.2.4 Barrier Performance.................................................................15.3-6 15.3.2.4.1 Speed Decrease of One Recirculation Pump..................................
15.3-6 15.3.2.4.2 Speed Decrease of Two Recirculati on Pumps................................15.3-7 15.3.2.5 Radiol ogical Consequences........................................................15.3-7 15.3.3 RECIRCULATION PUMP SEIZURE.............................................15.3-7 15.3.3.1 Identification of Causes and Frequency Classification........................15.3-7 15.3.3.2 Sequence of Even ts and Systems Operation.....................................15.3-7 15.3.3.2.1 Sequen ce of Events...............................................................15.3-7 15.3.3.2.2 System s Operation.................................................................15.
3-8 15.3.3.2.3 The Effect of Single Failures and Operat or Errors..........................15.3-8 15.3.3.3 Core and System Performance....................................................15.3-8 15.3.3.3.1 Mathematical Model..............................................................15.3-8 15.3.3.3.2 Input Parameters and Initia l Conditions.......................................15.3-8 15.3.3.3.3 Results...............................................................................
15.3-9 15.3.3.3.3.1 Consider ations of Uncertainties..............................................15.3-9 15.3.3.4 Barrier Performance.................................................................15.3-9 15.3.3.5 Radiol ogical Consequences........................................................15.3-9 15.3.4 RECIRCULATION PU MP SHAFT BREAK.....................................15.3-9 15.3.4.1 Identification of Causes and Frequency Classification........................15.3-9 15.3.4.1.1 Identification of Causes..........................................................15.
3-10 15.3.4.1.2 Frequency Classification.........................................................15.
3-10 15.3.4.2 Sequence of Even ts and Systems Operation.....................................15.3-10 15.3.4.2.1 Sequen ce of Events...............................................................15.3-10 15.3.4.2.2 System s Operation.................................................................15.
3-10 15.3.4.2.3 The Effect of Single Failures and Operat or Errors..........................15.3-10 C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Chapter 15 ACCIDENT ANALYSES
TABLE OF CONTENTS (Continued)
Section Page LDCN-08-035 15-xi 15.3.4.3 Core and System Performance....................................................15.
3-11 15.3.4.3.1 Qualitative Results................................................................15.3-11 15.3.4.4 Barrier Performance.................................................................15.3-11 15.3.4.5 Radiol ogical Consequences........................................................15.
3-11 15.
3.5 REFERENCES
.........................................................................
15.3-11 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES.................15.4-1 15.4.1 ROD WITHDRAWAL ERROR - LOW POWER................................15.4-1 15.4.1.1 Control Rod Rem oval Error During Refueling.................................15.4-1 15.4.1.1.1 Identificati on of Causes and Frequency Classification......................15.4-1 15.4.1.1.2 Sequence of Events and Systems Op eration..................................15.4-1 15.4.1.1.2.1 Initial Control Rod Removal..................................................15.4-1 15.4.1.1.2.2 Fuel Inserti on With Control Rod Removed................................15.4-1 15.4.1.1.2.3 Second Control Rod Removal................................................15.4-1 15.4.1.1.2.4 Cont rol Rod Removal Without Fuel Removal.............................15.4-1 15.4.1.1.2.5 Effect of Single Failure and Operator Errors..............................15.4-2 15.4.1.1.3 Core and System Performances.................................................15.4-2 15.4.1.1.4 Barrier Performance..............................................................15.4-2 15.4.1.1.5 Radiological Consequences......................................................15.4-2 15.4.1.2 Continuous Rod Withdrawal During Reactor Startup.........................15.4-2 15.4.1.2.1 Identificati on of Causes and Frequency Classification......................15.4-2 15.4.1.2.2 Sequence of Events and Systems Op eration..................................15.4-3 15.4.1.2.2.1 Sequence of Events.............................................................15.4-3 15.4.1.2.2.2 Effects of Single Failure and Operator Errors.............................15.4-3 15.4.1.2.3 Core and System Performance..................................................15.4-3 15.4.1.2.4 Barrier Performance..............................................................15.4-3 15.4.1.2.5 Radiological Consequences......................................................15.4-3 15.4.2 ROD WITHDRAWAL ERROR - AT POWER..................................15.4-4 15.4.2.1 Identification of Caus es and Frequenc y Classifications.......................15.4-4 15.4.2.1.1 Identification of Causes..........................................................15.
4-4 15.4.2.1.2 Frequency Classification.........................................................15.4-4 15.4.2.2 Sequence of Even ts and Systems Operation.....................................15.4-4 15.4.2.2.1 Sequen ce of Events...............................................................15.4-4 15.4.2.2.2 System s Operation.................................................................15.
4-4 15.4.2.2.3 Effect of Single Failu re and Operator Errors.................................15.4-5 C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Chapter 15 ACCIDENT ANALYSES
TABLE OF CONTENTS (Continued)
Section Page LDCN-08-035 15-xii 15.4.2.3 Core and System Performance....................................................15.4-5 15.4.2.3.1 Mathematical Model..............................................................15.4-5 15.4.2.3.2 Input Parameters and Initia l Conditions.......................................15.4-5 15.4.2.3.2.1 Rod Block Monitor System Operation......................................15.4-6 15.4.2.3.3 Results...............................................................................
15.4-6 15.4.2.3.4 Considerations of Uncertainties.................................................15.4-7 15.4.2.4 Barrier Performance.................................................................15.4-7 15.4.2.5 Radiol ogical Consequences........................................................15.4-7 15.4.3 CONTROL ROD MALOPERATION (SYSTEM MALFUNCTION OR OPERATOR ERROR)................................................................
15.4-7 15.4.4 STARTUP OF IDLE RECIRCULATION PUMP...............................
15.4-7 15.4.4.1 Identification of Causes and Frequency Classification........................15.4-7 15.4.4.1.1 Identification of Causes..........................................................15.
4-7 15.4.4.1.2 Frequency Classification.........................................................15.4-7 15.4.4.1.2.1 Normal Restart of Recirculation Pump at Power..........................15.4-7 15.4.4.1.2.2 Abnormal Start up of Idle Recirculation Pump.............................15.4-7 15.4.4.2 Sequence of Even ts and Systems Operation.....................................15.4-7 15.4.4.2.1 Sequen ce of Events...............................................................15.4-7 15.4.4.2.2 System s Operation.................................................................15.
4-8 15.4.4.2.3 The Effect of Single Failures and Operat or Errors..........................15.4-8 15.4.4.3 Core and System Performance....................................................15.4-8 15.4.4.3.1 Mathematical Model..............................................................15.4-8 15.4.4.3.2 Input Parameters and Initia l Conditions.......................................15.4-8 15.4.4.3.3 Results...............................................................................
15.4-8 15.4.4.3.4 Considerati on of Uncertainties..................................................15.4-9 15.4.4.4 Barrier Performance.................................................................15.4-9 15.4.4.5 Radiol ogical Consequences........................................................15.4-9 15.4.5 RECIRCULATION FLOW CONTROL FAILURE WITH INCREASING FLOW.................................................................
15.4-9 15.4.5.1 Identification of Causes and Frequency Classification........................15.4-9 15.4.5.1.1 Identification of Causes..........................................................15.
4-9 15.4.5.1.2 Frequency Classification.........................................................15.4-9 15.4.5.2 Sequence of Even ts and Systems Operation.....................................15.4-10 15.4.5.2.1 The Effect of Single Failures and Operat or Errors..........................15.4-10 15.4.5.3 Core and System Performance....................................................15.
4-10 15.4.5.3.1 Mathematical Model..............................................................15.4-10 C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 Chapter 15 ACCIDENT ANALYSES
TABLE OF CONTENTS (Continued)
Section Page LDCN-10-029 15-xiii 15.4.5.3.2 Input Pa rameters and Initial Conditions
....................................... 15.4-11 15.4.5.3.3 Re sults ...............................................................................
15.4-11 15.4.5.3.4 Considerations of Uncertain ties .................................................
15.4-11 15.4.5.4 Barrier Performance
................................................................. 15.4-11 15.4.5.5 Radiological Consequences ........................................................ 15.4-11 15.4.6 CHEMICAL AND VOLUME CONTROL SYSTEM MALFUNCTIONS .. 15.4-12 15.4.7 MISPLA CED BUNDLE ACCIDENT ............................................. 15.4-12 15.4.7.1 Identification of Causes and Frequency Classification ........................ 15.4-12 15.4.7.1.1 Identificati on of Causes .......................................................... 15.4-12 15.4.7.1.2 Frequency of Occurrence ........................................................ 15.
4-12 15.4.7.2 Sequence of Events and Systems Operation ..................................... 15.4-13 15.4.7.2.1 Effect of Single Failure and Operator Errors
................................. 15.4-13 15.4.7.3 Core and Syst em Performanc e ....................................................
15.4-13 15.4.7.3.1 Mathematical Model
.............................................................. 15.4-13 15.4.7.3.2 Input Pa rameters and Initial Conditions
....................................... 15.4-13 15.4.7.3.3 Re sults ...............................................................................
15.4-13 15.4.7.3.4 Considerations of Uncertain ties .................................................
15.4-14 15.4.7.4 Barrier Performance
................................................................. 15.4-14 15.4.7.5 Radiological Consequences ........................................................ 15.4-14 15.4.8 SPECTRUM OF ROD EJECTION ASSEMBLIES .............................. 15.4-14
15.4.9 CONTROL ROD DR OP ACCIDENT ............................................. 15.4-14 15.4.9.1 Identification of Causes and Frequency Classification ........................ 15.4-14 15.4.9.1.1 Identificati on of Causes .......................................................... 15.4-14 15.4.9.1.2 Frequency Classification ......................................................... 15.4-15 15.4.9.2 Sequence of Events and Systems Operation ..................................... 15.4-15 15.4.9.2.1 Effect of Single Failure s and Operator Errors ...............................
15.4-16 15.4.9.3 Core and Syst em Performanc e ....................................................
15.4-16 15.4.9.3.1 Mathematical Model
.............................................................. 15.4-16 15.4.9.3.2 Input Pa rameters and Initial Conditions
....................................... 15.4-16 15.4.9.3.3 Re sults ...............................................................................
15.4-16 15.4.9.4 Barrier Performance
................................................................. 15.4-17 15.4.9.5 Radiological Consequences ........................................................ 15.4-17 15.4.9.5.1 Fission Product Re lease from Fuel ............................................. 15.4-17 15.4.9.5.2 Fission Pr oduct Transport to the Environment
............................... 15.4-18 15.4.9.5.3 Re sults ...............................................................................
15.4-18 15.4.10 REFERE NCES ........................................................................
15.4-19 C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 Chapter 15
ACCIDENT ANALYSES
TABLE OF CONTENTS (Continued)
Section Page LDCN-03-003 15-xiv 15.5 INCREASE IN REACTOR COOLANT INVENTORY...........................15.5-1 15.5.1 INADVERTENT HIGH-PRESSURE CORE SPRAY STARTUP............15.5-1 15.5.1.1 Identification of Causes and Frequency Classification........................15.5-1 15.5.1.1.1 Identification of Causes..........................................................15.
5-1 15.5.1.1.2 Frequency Classification.........................................................15.5-1 15.5.1.2 Sequence of Even ts and Systems Operation.....................................15.5-1 15.5.1.2.1 The Effect of Single Failures and Operat or Errors..........................15.5-1 15.5.1.3 Core and System Performance....................................................15.5-2 15.5.1.3.1 Mathematical Model..............................................................15.5-2 15.5.1.3.2 Input Parameter and Initia l Conditions........................................15.5-2 15.5.1.3.3 Results...............................................................................
15.5-2 15.5.1.3.3.1 Consider ation of Uncertainties...............................................15.5-2 15.5.1.4 Barrier Performance.................................................................15.5-2 15.5.1.5 Radiol ogical Consequences........................................................15.5-2 15.5.2 CHEMICAL VOLUME CONTROL SYSTEM MALFUNCTION (OR OPERATOR ERROR)..........................................................
15.5-2 15.5.3 BOILING WATER REACTOR TR ANSIENTS WHICH INCREASE REACTOR COOLANT IN VENTORY............................................15.5-2 15.
5.4 REFERENCES
.........................................................................
15.5-3 15.6 DECREASE IN REACTOR COOLANT INVENTORY..........................15.6-1 15.6.1 INADVERTENT SAFETY/R ELIEF VALVE OPENING.....................15.6-1 15.6.2 INSTRUMENT LINE PIPE BREAK...............................................15.6-1 15.6.2.1 Identification of Causes and Frequency Classification........................15.6-1 15.6.2.1.1 Identification of Causes..........................................................15.
6-1 15.6.2.1.2 Frequency Classification.........................................................15.6-1 15.6.2.2 Sequence of Even ts and Systems Operation.....................................15.6-1 15.6.2.2.1 The Effect of Single Failures and Operat or Errors..........................15.6-2 15.6.2.3 Core and System Performance....................................................15.6-2 15.6.2.3.1 Qualitative Summary - Results..................................................15.6-2 15.6.2.4 Barrier Performance.................................................................15.6-2 15.6.2.4.1 General..............................................................................
15.6-2 15.6.2.5 Radiol ogical Consequences........................................................15.6-2 15.6.2.5.1 Results...............................................................................
15.6-3 15.6.3 STEAM GENERATO R TUBE FAILURE........................................15.6-3 C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 Chapter 15
ACCIDENT ANALYSES
TABLE OF CONTENTS (Continued)
Section Page LDCN-05-009,06-043 15-xv 15.6.4 STEAM SYSTEM PIPING BREAK OUTSIDE CONTAINMENT..........15.6-3 15.6.4.1 Identification of Causes and Frequency Classification........................15.6-3 15.6.4.1.1 Identification of Causes..........................................................15.
6-3 15.6.4.1.2 Frequency Classification.........................................................15.6-3 15.6.4.2 Sequence of Even ts and Systems Operation.....................................15.6-4 15.6.4.2.1 Sequen ce of Events...............................................................15.6-4 15.6.4.2.2 System s Operation.................................................................15.
6-4 15.6.4.2.3 The Effect of Single Failures and Operat or Errors..........................15.6-4 15.6.4.3 Core and System Performance....................................................15.6-4 15.6.4.3.1 Input Parameters and Initia l Conditions.......................................15.6-4 15.6.4.3.2 Results...............................................................................
15.6-5 15.6.4.3.3 Considerations of Uncertainties.................................................15.6-5 15.6.4.4 Barrier Performance.................................................................15.6-5 15.6.4.5 Radiol ogical Consequences........................................................15.6-5 15.6.4.5.1 Results...............................................................................
15.6-6 15.6.5 LOSS-OF-COOLANT ACCIDENTS (RESULTING FROM SPECTRUM OF POSTULATED PIPING BREAKS WITHIN THE REACTOR
COOLANT PRESSURE BOUNDARY) - INSIDE CONTAINMENT.......15.6-6 15.6.5.1 Identification of Causes and Frequency Classification........................15.6-7 15.6.5.1.1 Identification of Causes..........................................................15.
6-7 15.6.5.1.2 Frequency Classification.........................................................15.6-7 15.6.5.2 Sequence of Even ts and Systems Operation.....................................15.6-7 15.6.5.3 Core and System Performance....................................................15.6-7 15.6.5.4 Radiol ogical Consequences........................................................15.6-7 15.6.5.4.1 Design Ba sis Analysis............................................................15.6-7 15.6.5.4.1.1 Fission Product Release from Fuel..........................................15.6-8 15.6.5.4.1.2 Fission Product Transport to the Environment............................15.6-8 15.6.5.4.1.3 Suppre ssion Pool pH Control.................................................15.6-9 15.6.5.4.1.4 Results............................................................................15.6-9 15.6.6 FEEDWATER LINE BREAK - OUTSIDE CONTAINMENT................15.6-9 15.6.6.1 Identification of Causes and Frequency Classification........................15.6-9 15.6.6.1.1 Identification of Causes..........................................................15.
6-9 15.6.6.1.2 Frequency Classification.........................................................15.6-9 15.6.6.2 Sequence of Even ts and Systems Operation.....................................15.6-9 15.6.6.2.1 Sequen ce of Events...............................................................15.6-9 15.6.6.2.2 System s Operation.................................................................15.
6-9 C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 Chapter 15 ACCIDENT ANALYSES
TABLE OF CONTENTS (Continued)
Section Page LDCN-09-007 15-xvi 15.6.6.2.3 The Effect of Single Fa ilures and Operator Errors .......................... 15.6-10 15.6.6.3 Core and Syst em Performanc e ....................................................
15.6-10 15.6.6.3.1 Qualitati ve Summary
............................................................. 15.6-10 15.6.6.3.2 Qualitative Results ................................................................ 15.6-10 15.6.6.3.3 Consideration of Uncertainties .................................................. 15.6-10 15.6.6.4 Barrier Performance
................................................................. 15.6-10 15.6.6.5 Radiological Consequences ........................................................ 15.6-11 15.6.6.5.1 Fission Pr oduct Release
.......................................................... 15.6-11 15.6.6.5.2 Fission Pr oduct Transport to the Environment
............................... 15.6-11 15.6.6.5.3 Re sults ...............................................................................
15.6-12 15.
6.7 REFERENCES
......................................................................... 15.6-12
15.7 RADIOACTIVE RELEASE FROM SUBSYSTEMS AND COMPONENTS ................................................................... 15.7-1 15.7.1 RADIOACTIVE GAS WASTE SYSTEM LEAK OR FAILURE ............. 15.7-1 15.7.2 LIQUID RADIOACTIVE SYSTEM FAILURE
.................................. 15.7-1 15.7.3 POSTULATED RADIOACTIVE RELEASES DUE TO LIQUID RADWASTE TANK FAILURE ..................................................... 15.7-1 15.7.3.1 Identification of Causes and Frequency Classification ........................ 15.7-1 15.7.3.1.1 Identification of Causes .......................................................... 15.7-1 15.7.3.1.2 Frequency Classification ......................................................... 15.7-1 15.7.3.2 Sequence of Events and Systems Operation ..................................... 15.7-1 15.7.3.2.1 Sequence of Events ............................................................... 15.7-1 15.7.3.2.2 Systems Operation................................................................. 15.7-2 15.7.3.2.3 The Effects of Singl e Failures and Operator Errors
......................... 15.7-2 15.7.3.3 Core and Syst em Performa nce ....................................................
15.7-2 15.7.3.4 Barrier Performance ................................................................. 15.7-2 15.7.3.5 Radiological Consequences ........................................................ 15.7-3 15.7.4 FUEL HANDLIN G ACCIDENT ................................................... 15.7-3 15.7.4.1 Identification of Causes and Frequency Classification ........................ 15.7-3 15.7.4.1.1 Identification of Causes .......................................................... 15.7-3 15.7.4.1.2 Frequency Classification ......................................................... 15.7-3 15.7.4.2 Sequence of Events and Systems Operation ..................................... 15.7-4 15.7.4.2.1 The Effects of Singl e Failures and Operator Errors
......................... 15.7-4 15.7.4.3 Core and Syst em Performa nce ....................................................
15.7-4 15.7.4.3.1 Mathematical Model .............................................................. 15.7-4 C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Chapter 15 ACCIDENT ANALYSES
TABLE OF CONTENTS (Continued)
Section Page LDCN-08-035 15-xvii 15.7.4.3.2 Input Parameters and Initia l Conditions.......................................15.7-5 15.7.4.3.3 Results...............................................................................
15.7-5 15.7.4.4 Barrier Performance.................................................................15.7-5 15.7.4.5 Radiol ogical Consequences........................................................15.7-6 15.7.4.5.1 Design Ba sis Analysis............................................................15.7-6
15.7.4.5.1.1 Fission Product Release From Fuel.........................................15.7-6 15.7.4.5.1.2 Fission Product Transport to the Environment............................15.7-6 15.7.4.5.1.3 Results............................................................................15.7-6 15.7.5 SPENT FUEL CASK DROP ACCIDENT........................................
15.7-6 15.
7.6 REFERENCES
.........................................................................
15.7-7 15.8 ANTICIPATED TRANSIENTS WITHOUT SCRAM.............................15.8-1 15.8.0 CAPABILITIES OF PRES ENT DESIGN TO ACCOMMODATE ANTICIPATED TRANSIENTS WITHOUT SCRAM.........................15.8-1 15.8.1 INADVERTENT CONTRO L ROD WITHDRAWAL.........................15.8-2 15.8.2 LOSS OF F EEDWATER............................................................15.8-2 15.8.2.1 Identification of Causes and Frequency Classification........................15.8-2 15.8.2.1.1 Identification of Causes..........................................................15.
8-2 15.8.2.1.2 Frequency Classification.........................................................15.8-3 15.8.2.2 Sequence of Even ts and System Operation......................................15.8-3 15.8.2.2.1 Sequen ce of Events...............................................................15.8-3 15.8.2.2.1.1 Identifica tion of Operator Actions...........................................15.8-3 15.8.2.2.2 System Operation..................................................................15.8-3 15.8.2.2.3 The Effect of Single Failure and Operator Errors...........................15.8-3 15.8.2.3 Core and System Performance....................................................15.8-4 15.8.2.3.1 Mathematical Model..............................................................15.8-4 15.8.2.3.2 Input Parameters and Initia l Conditions.......................................15.8-4 15.8.2.3.3 Results...............................................................................
15.8-4 15.8.2.3.4 Considerati on of Uncertainties..................................................15.8-4 15.8.2.4 Barrier Performance.................................................................15.8-5 15.8.2.5 Radiol ogical Consequences........................................................15.8-5 15.8.3 LOSS OF ALTERNA TE CURRENT POWER...................................15.8-5 15.8.4 LOSS OF ELECTRICAL LOAD...................................................
15.8-5 15.8.5 LOSS OF COND ENSER VACUUM...............................................
15.8-5 15.8.6 TURBINE TRIP........................................................................
15.8-5 C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 Chapter 15 ACCIDENT ANALYSES
TABLE OF CONTENTS (Continued)
Section Page LDCN-09-007 15-xviii 15.8.7 CLOSURE OF MAIN STEA M LINE ISOLATION VALVES ................ 15.8-6 15.8.7.1 Identification of Causes and Frequency Classification ........................ 15.8-6 15.8.7.1.1 Identification of Causes .......................................................... 15.8-6 15.8.7.1.2 Frequency Classification ......................................................... 15.8-6 15.8.7.2 Sequence of Events and System Operation ...................................... 15.8-6 15.8.7.2.1 Sequence of Events ............................................................... 15.8-6 15.8.7.2.1.1 Identification of Operator Actions ........................................... 15.8-6 15.8.7.2.2 System Operati on ..................................................................
15.8-6 15.8.7.2.3 The Effect of Single Fa ilures and Operator Errors .......................... 15.8-7 15.8.7.3 Core and Syst em Performa nce ....................................................
15.8-7 15.8.7.3.1 Mathematical Model .............................................................. 15.8-7 15.8.7.3.2 Input Parameters and Initia l Conditions ....................................... 15.8-7 15.8.7.3.3 Results ............................................................................... 15.8-7 15.8.7.3.4 Consideration of Uncertainties .................................................. 15.8-8 15.8.7.4 Barrier Performance ................................................................. 15.8-8 15.8.7.5 Radiological Consequences ........................................................ 15.8-8 15.8.8 INADVERTENT OPENIN G OF RELIEF VALVE ............................. 15.8-9 15.8.8.1 Identification of Causes and Frequency Classification ........................ 15.8-9 15.8.8.1.1 Identification of Causes .......................................................... 15.8-9 15.8.8.1.2 Frequency Classification ......................................................... 15.8-9 15.8.8.2 Sequence of Events and System Operation ...................................... 15.8-9 15.8.8.2.1 Sequence of Events ............................................................... 15.8-9 15.8.8.2.1.1 Identification of Operator Actions ........................................... 15.8-9 15.8.8.2.1.2 System Operation ............................................................... 15.8-9 15.8.8.2.2 The Effect of Single Fa ilures and Operator Errors .......................... 15.8-9 15.8.8.3 Core and Syst em Performanc e ....................................................
15.8-10 15.8.8.3.1 Mathematical Model
.............................................................. 15.8-10 15.8.8.3.2 Input Pa rameters and Initial Conditions
....................................... 15.8-10 15.8.8.3.3 Re sults ...............................................................................
15.8-10 15.8.8.3.4 Consideration of Uncertainties .................................................. 15.8-11 15.8.8.4 Barrier Performance
................................................................. 15.8-11 15.8.8.5 Radiological Consequences ........................................................ 15.8-11 15.8.9 PRESSURE REGULATOR FA ILURE - OPEN (P REGO)..................... 15.8-11 15.8.9.1 Identification of Causes and Frequency Classification ........................ 15.8-11 15.8.9.1.1 Identificati on of Causes .......................................................... 15.8-11 15.8.9.1.2 Frequency Classification ......................................................... 15.8-11 C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 Chapter 15 ACCIDENT ANALYSES
TABLE OF CONTENTS (Continued)
Section Page LDCN-09-007 15-xix 15.8.9.2 Sequence of Events and System Operation ...................................... 15.8-11 15.8.9.2.1 Sequence of Events
............................................................... 15.8-11 15.8.9.2.1.1 Identification of Operator Actions ........................................... 15.8-12 15.8.9.2.1.2 System Operation
............................................................... 15.8-12 15.8.9.2.3 The Effect of Single Fa ilures and Operator Errors .......................... 15.8-12 15.8.9.3 Core and Syst em Performanc e ....................................................
15.8-12 15.8.9.3.1 Mathematical Model
.............................................................. 15.8-12 15.8.9.3.2 Input Pa rameters and Initial Conditions
....................................... 15.8-12 15.8.9.3.3 Re sults ...............................................................................
15.8-13 15.8.9.3.4 Consideration of Uncertainties .................................................. 15.8-13 15.8.9.4 Barrier Performance
................................................................. 15.8-13 15.8.9.5 Radiological Consequences ........................................................ 15.8-14 15.8.10 SINGLE REACTOR RECIRCULATION SYSTEM PUMP OPERATION ................................................................ 15.8-14 15.8.11 EXTENDED LOAD LINE LIMIT ANALYSIS OPERATION .............. 15.8-15 15.8.12 REFERE NCES ........................................................................
15.8-15
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 Chapter 15 ACCIDENT ANALYSES
LIST OF TABLES
Number Title Page LDCN-10-003,10-029 15-xx 15.0-1 Results Summary of Transient Events Applicable to Columbia Generating Sta tion ..............................................................
15.0-17 15.0-1A Summary of Transient Peak Value Results Single-Loop Operation .... 15.0-20
15.0-2 Input Parameters and Initial Conditions for Transients................... 15.0-21
15.0-2A Input Parameters and Initial Conditions for Transients and Accidents for Single-L oop Operation
........................................ 15.0-24
15.0-2B Input Parameters a nd Initial Conditions for GNF Reload Transient ... 15.0-26
15.0-3 Summary of A ccidents .........................................................
15.0-28 15.0-4 /Q (s/m 3) values for the EAB and LPZ .................................... 15.0-29 15.0-5 Control Room Atmospheric Dispersion Factors (sec/m
- 3) ................ 15.0-30
15.1-1 Sequence of Events for Figure 15.1-1
....................................... 15.1-17
15.1-1A Sequence of Events for Figure 15.1-3
....................................... 15.1-18
15.1-2 Sequence of Events for Figure 15.1-2
....................................... 15.1-19
15.1-3 Sequence of Events for Inadvertent Safety/Relief Valve Opening ..... 15.1-20
15.1-4 Sequence of Events for In advertent Residual Heat Removal Shutdown Cooling Op eration .................................................
15.1-21 15.2-1 Sequence of Events for Figure 15.2-1
....................................... 15.2-35
15.2-2 Sequence of Events fo r Figure 15.2-2.1 .................................... 15.2-36 15.2-3 Sequence of Events fo r Figure 15.2-2.2 .................................... 15.2-37 15.2-4 Sequence of Events for Figure 15.2-3
....................................... 15.2-38 C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 Chapter 15 ACCIDENT ANALYSES
LIST OF TABLES (Continued)
Number Title Page LDCN-10-029 15-xxi 15.2-5 Sequence of Events for Figure 15.2-4
....................................... 15.2-39
15.2-6 Sequence of Events for Figure 15.2-5
....................................... 15.2-40
15.2-7 Sequence of Events for Figure 15.2-6
....................................... 15.2-41
15.2-8 Trip Signals Associated with Loss-of-Condenser Vacuum .............. 15.2-42
15.2-9 Sequence of Events for Figure 15.2-7
....................................... 15.2-43
15.2-10 Sequence of Events for Figure 15.2-8
....................................... 15.2-44
15.2-11 Sequence of Events for Figure 15.2-9
....................................... 15.2-45
15.2-12 Sequence of Events for Failure of Residual Heat Removal Shutdown Cooling .............................................................. 15.2-46
15.2-13 Evaluation of Failure of Residual Heat Removal Shutdown Cooling .............................................................. 15.2-47
15.3-1 Sequence of Events for Figure 15.3-1
....................................... 15.3-13
15.3-2 Sequence of Events for Figure 15.3-2
....................................... 15.3-14
15.3-3 Sequence of Events for Figure 15.3-3
....................................... 15.3-15
15.3-4 Sequence of Events for Figure 15.3-4
....................................... 15.3-16
15.3-5 Sequence of Events for Figure 15.3-5
....................................... 15.3-17
15.3-6 Sequence of Events for Pump Seizure (for Single Loop Operation) ... 15.3-18
15.4-1 Sequence of Events - Rod Withdrawal Error in Power Range .......... 15.4-21
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 Chapter 15
ACCIDENT ANALYSES
LIST OF TABLES (Continued)
Number Title Page LDCN-05-009 15-xxii 15.4-2 Sequence of Events for an Abnormal Startup of an Idle Recirculation Loop.............................................................
15.4-22 15.4-3 Reactor Recirculation Pump Flow Increase Input Parameters and Initial Conditions................................................................15.
4-23 15.4-4 Control Rod Drop Accident Evaluation Parameters......................
15.4-24 15.4-5 Control Rod Drop Accident Activity Airborne in the Condenser (Curies).............................................................15.4-26
15.4-6 Control Rod Drop Accident Activity Airborne to the Environment (Curies)..........................................................
15.4-29 15.4-7 Control Rod Drop Accident Radiological Effect s (rem).................15.4-32
15.5-1 Input Parameters and Initial Conditions HPCS Injection................15.5-5
15.6-1 Instrument Line Break Acci dent - Parameters Tabulated for Postulated Accident Analyses.................................................15.6-13
15.6-2 Instrument Line Failure........................................................15.
6-15 15.6-3 Instrument Line Failur e Radiological Effects..............................
15.6-16 15.6-4 Sequence of Events for Steam Line Break Outside Containment.......15.6-17
15.6-5 Steam Line Break Accident - Parameters Tabulated for Postulated Accident Analyses.................................................15.6-18
15.6-6 Steam Line Break Acci dent Activity Release to Environment (Curies)..........................................................
15.6-20 15.6-7 Steam Line Break Acci dent Radiological Effects of a Puff Release...15.6-21
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 Chapter 15
ACCIDENT ANALYSES
LIST OF TABLES (Continued)
Number Title Page LDCN-05-009 15-xxiii 15.6-8 Loss-of-Coolant Accident - Parameters Tabulated for Postulated Accident Analysis.................................................15.6-22
15.6-9 Loss-of-Coolant Accident Primary Containment Activity (Curies)....15.6-25
15.6-10 Loss-of-Coolant Accide nt Secondary Containment Activity (Curies) - 20 Minute Drawdown..................................15.
6-26 15.6-11 Loss-of-Coolant Accident Activity Released to the Environment (Curies) - 20 Minute Drawdown............................
15.6-27 15.6-12 Loss-of-Coolant Accident (Design Basis Analysis) Radiological Effects............................................................15.6-28
15.6-13 Sequence of Events for Feedwater Line Break Outside Containment......................................................................15.
6-29 15.6-14 Feedwater Line Break Acci dent - Parameters Tabulated for Postulated Accident Analysis.............................................15.
6-30 15.6-15 Feedwater Line Break A ccident Activity Release to Environment (Curies)..........................................................
15.6-32 15.6-16 Feedwater Line Break Acci dent Biological Effects of a Puff Release......................................................................15.
6-33 15.7-1 Liquid Radwaste Tanks Failure - Parameters and Concentrations.....15.7-9 15.7-2 Fuel Handling Accident Parameters Tabulated for Postulated Accident Analysis.................................................15.7-10 15.7-3 Fuel Handling Accident Activity Airborne in Secondary Containment (Curies)...........................................................
15.7-12 15.7-4 Fuel Handling Accident Activity Released to the Environment (Curies)..........................................................
15.7-13 C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 Chapter 15
ACCIDENT ANALYSES
LIST OF TABLES (Continued)
Number Title Page LDCN-07-011 15-xxiv 15.7-5 Fuel Handling Accident (Design Basis Analysis) Radiological Effects............................................................15.7-14
15.8-1 Anticipated Transients Without Scram Analysis Initial Conditions....15.8-17
15.8-2 Anticipated Transients W ithout Scram Analysis Equipment Performance Characteristics..................................................15.8-18
15.8-3 Summary of Anticipated Transi ents Without Scram Results............15.8-19
15.8-4 Sequence of Events fo r Loss of Feedwater.................................15.8-20
15.8-5 Sequence of Events for Main Steam Line Isolation Valve Closure (Long Term Transient)..............................................15.8-21
15.8-6 Sequence of Events for Main Steam Line Isolation Valve Closure with Four Safety/Relief Valves Out-of-Service.................15.8-22
15.8-7 Sequence of Events for Inadvertent Open Relie f Valve..................15.8-23
15.8-8 Sequence of Events for Pre ssure Regulator Failure Open (Long Term Transient).........................................................
15.8-24 C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Chapter 15 ACCIDENT ANALYSES
LIST OF FIGURES
Number Title LDCN-08-035 15-xxv 15.0-1 Scram Position and Reactivity Characteristics 15.0-2 Illustration of Single Recirculation Loop Operation Flows 15.1-1 Feedwater Controlle r Failure, Maximum Demand
15.1-2 Pressure Regulator Failure - Open at 106.2% Uprated Power, 100% Flow (Sheets 1 through 11)
15.1-3 Feedwater Controller Failure, Maximum Demand, EOC RPT OOS, Single Loop Operation and 75% Uprated Power, 57% Flow (Sheets 1 through 4)
15.2-1 Pressure Regulatory Failure - Down Scale Failure at 104.1% Uprated Power, 106% Flow
15.2-2.1 Generator Load Reject ion with Bypass On - Original Rated Power (Sheet 1)
15.2-2.2 Generator Load Re jection with BP Failure 15.2-3 Turbine Trip, Trip Sc ram, Bypass, and RPT - On
15.2-4 Turbine Trip with Bypass Failure
15.2-5 Main Steam Line Isolation Valv e Closure at 106.2% Uprated Power, 100% Rated Flow
15.2-6 Loss of Condenser Vacuum at 104.
1% Uprated Power, 100% Rated Flow 15.2-7 Loss of Auxiliary Power Transformers - 106.2% Uprated Power, 100% Rated Flow 15.2-8 Loss of All Grid Connections - 104.
1% Uprated Power, 100% Rated Flow
15.2-9 Loss of All Feedwater Flow - 106.
2% Uprated Power, 100% Rated Flow
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Chapter 15 ACCIDENT ANALYSES
LIST OF FIGURES (Continued)
Number Title 15-xxvi 15.2-10 Automatic Depressu rization System/Residual H eat Removal Cooling Loops (Sheets 1 and 2) 15.2-11 Summary of Paths Availa ble to Achieve Cold Shutdown
15.2-12 Activity C1 Alternate Shutdown Coo ling Path Utilizing Residual Heat Removal Loop B 15.2-13 Residual He at Removal Loop C
15.2-14 Residual Heat Rem oval Loop A (B) (Suppression Pool Cooling/Rated Pump Flow Test Mode)
15.2-15 Activity C2 Alternate Shutdown Coo ling Path Utilizing Residual Heat Removal Loop A 15.2-16 Vessel Temperature and Pressure Versus Time (Activity C1.b.1 or C2)
15.2-17 Vessel Temperature and Pressure Versus Time (Activity C1.b.2)
15.2-18 Suppression Pool Temperature Versus Time (with 87°F Service Water Temperature) (Activity C1.b.1 or C.2)
15.2-19 Suppression Pool Temperature Versus Time (with 87°F Service Water Temperature) (Activity C1.b.2)
15.3-1 One Recirculation Pump Trip at 106.2% Uprated Power, 100% Flow (Sheets 1 through 5)
15.3-2 Two Recirculation Pump Trip at 106.2% Uprated Power, 100% Flow (Sheets 1 through 5)
15.3-3 Recirculation Flow Control Failure - Decreasing Flow in One Loop at 106.2% Uprated Power, 100% Flow (Sheets 1 through 5)
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Chapter 15 ACCIDENT ANALYSES
LIST OF FIGURES (Continued)
Number Title LDCN-08-035 15-xxvii 15.3-4 Recirculation Flow Control Failure - Decreasing Flow in Two Loops, (5%/Sec Ramp) at 106.2% Uprated Power, 100% Flow (Sheets 1 through 5)
15.3-5 Recirculation Pump Seizure at 106% Uprated Power, 100% Flow (Sheets 1 through 5)
15.3-6 SLO Recirculation Pump Seizure Results
15.4-1 Abnormal Startup of an Idle Reci rculation Loop at 57.
9% Uprated Power, 34.1% Flow (Sheets 1 through 5)
15.4-2 Leakage Path Model for Rod Drop Accident (Origi nal Rated Power)
15.5-1 Inadvertent Start of Hi gh-Pressure Core Spray Pump at 102% Uprated Power, 88% Flow 15.6-1 Leakage Path for Instrument Line Break
15.6-2 Steam Flow Schematic for Steam Break Outside Containment
15.6-3 Leakage Path for LOCA
15.6-4 Leakage Path for Feedwater Line Break Outside Containment
15.7-1 Leakage Path for Fuel Handling Accident
15.8-1 Loss of Feedwater Ev ent (Sheets 1 through 5) 15.8-2 Main Steam Isolation Valve Cl osure Event (Sheets 1 through 5) 15.8-3 Main Steam Isolation Valve Closure Event with Four SRVs Out-of-Service (Sheets 1 through 5)
15.8-4 Inadvertent Opening of Relief Valve Event (Sheets 1 through 5)
15.8-5 Pressure Regulator Failure -
Open Event (Sheets 1 through 5)
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 15.0-1 Chapter 15
ACCIDENT ANALYSES
15.0 GENERAL
This chapter discusses th e effects of anticipated process disturbances and postulated component failures, their consequences, and the capabilities built into the plant to control or accommodate such failures and events. The analyses have been reviewed and revised, as needed, for the:
- Reactor power uprate,
- Installation of the adjustable speed dr ive for the reactor recirculation pumps,
- Implementation of an alternative source term, and
- Cycle specific changes.
These changes to the plant licen sing and design basis and their impact are discussed in this chapter.
The scope of the situations analyzed includes anticipated (expected) operational occurrences, off-design abnormal (unexpect ed) transients that indu ce system operating condition disturbances, postulated accidents of low probability, and hypotheti cal events of extremely low probability. For each reload, the events are evaluated by the fuel vendor(s). The events identified as limiting during the evaluation are analyzed and the sections are revised.
The plant was originally licensed at 3323 MWt.
In 1995, an amendment to the plant Operating License authorized an increas e in power to 3486 MWt. The power uprate analysis was performed in accordance with the NRC-approved General Electric Company (GE) generic power uprate program for bo iling water reactors (BWRs).
The postulated events in this chapter have b een analyzed for power uprate conditions. The only exceptions to using uprated power are some non-limiting single loop operation (SLO) transients that were not reanal yzed as part of the GE power uprate transient analysis. Their text and figures are clearly marked with ORIGINAL POWER designation. Limiting events in
terms of setting the fuel opera ting limits (e.g., Loss of Feedwa ter Heating, Generator Load Rejection Without Bypass) are r eanalyzed on a cycle specific ba sis and therefore, may include fuel vendor results and references.
The events in this chapter have been analyzed for application of the adjustable speed drives (ASD) in place of the former reactor recirculation control system that used flow control valves (FCV). The uprated power for Columbia Ge nerating Station is 3486 MWt which is 4.9% higher than the original license d power of 3323 MWt. All transient initial conditions are specified in Table 15.0-2 , 15.0-2A , 15.0-2B or the individual transient event description sections. Several performance improvement packages have been included in the analysis:
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 15.0-2
- 1. Final Feedwater Temperature Reduction (FFWTR) and Feedwater Heaters Out-Of-Service (FWHOOS) - Analyses were performe d at a reduced feedwater temperature at rated thermal power for operations at end-of-cycle and during the cycle for limiting transients.
- 2. Increased Core Flow (ICF) - Increased co re flow allows for ope ration at 106% of the rated core flow. The limiting transients were performed for the end-of-cycle with control rods fully withdrawn. This envelopes the operation at increased core flow condition throughout the cycle.
- 3. Extended Load Line Limit Analysis (ELLLA) - The consequences of the transients were evaluated to determine if operating limit adjustments are n ecessary for operation in the extended operating domain and compar ed with the evaluati on at rated thermal power and increased core flow region.
This comparison ensures bounding of the results at the exte nded operating domain.
- 4. Single Loop Operation (SLO) - Limiting transients were re-analyzed for operation at SLO. Using adjustable speed drives (ASD), GE determined the maximum active loop's recirculation flow at 105% of rated pump speed with resultant analyzed power and core flow conditions of 75% UP a nd 57% of rated core flow for SLO. Prior to power uprate, a comprehensive SLO analysis was performed. For non-limiting events, this analysis has been retained for completeness and historical purposes. These analyses are clearly marked with ORIG INAL POWER designation.
- 5. End-of-Cycle RPT Out-of-Service (RPT OOS) - The recirculation pump trip (RPT) mitigates several transients that are more severe at end-of-cycle. The limiting transients were re-analyzed with RPT OOS at various power/flow conditions.
- 6. Turbine Bypass Out-of-Service - Limiting transient events have been analyzed with turbine bypass valves out-o f-service at limiting pow er and flow conditions.
Additional updates address the implementation of the use of alternative source terms (AST) as described in the Regulatory Guide 1.183, July 2000, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nucl ear Power Reactors." This methodology is based on the advances that have been made in understanding the timing, magnitude, and the chemical form of fission prod uct releases from severe nuclear power plant accidents. The accidents that were reanalyzed by Energy Northwest with the AST:
- Loss of coolant accident (LOCA)
- Fuel handling accident (FHA)
- Control rod drop accident (CRDA)
- Main steam line break, outside containment, accident (MSLB)
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 15.0-3 The radiological consequences of these accidents were de termined based on AST approved for use under 10 CFR 50.67. In accordance with the guidance provided in Regulatory Guide 1.183, the licensing and desi gn basis are revised to reflect the application of full scope AST methodology (with the exception that the TID-14844 will continue to be used as the basis for equipment qualification (EQ) and radiati on zone maps/shielding calculations). The accidents analyzed as part of the implementation of AST are subject to the limits specified in 10 CFR 50.67 and guidelines of Regulatory Guide 1.183.
15.0.1 ANALYTICAL OBJECTIVE
The spectrum of postulated initiating events is divided into categories based on the type of disturbance and the expected fr equency of the initiating occurren ce. The limiting events in each combination of category and frequency are quantitatively analyzed.
15.0.2 ANALYTICAL CATEGORIES
Transient and accident events contained in this report are provided in individual categories as specified by Regulatory Guide 1.70, Revision 2. The results of the events are summarized in Table 15.0-1. Events evaluated are assigned to one of the following app licable categories:
- a. Decrease in reactor coolant temperature:
Reactor vessel water (moderator) temperature reduction results in an increase in core reactivity. This could lead to fuel-cladding damage.
- b. Increase in reactor pressure:
Nuclear system pressure increases threaten to rupture the reactor coolant pressure boundary (RCPB). Increasing pre ssure also collapses the voids in the core-moderator, thereby increasing core reactivity and power level that could threaten fuel cladding due to overheating.
- c. Decrease in reactor coolant system flow rate:
A reduction in the core coolant flow ra te could overheat th e cladding as the coolant becomes unable to adequately re move the heat generated by the fuel.
- d. Reactivity and power distribution anomalies:
Transient events included in this category are those that could cause rapid increases in power due to increased core flow disturbance events. Increased C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 15.0-4 core flow reduces the void content of the moderator increa sing core reactivity and power level.
- e. Increase in reactor coolant inventory:
Increasing coolant inventory could resu lt in excessive moisture carryover to components such as the main turb ine, feedwater turbines, etc.
- f. Decrease in reactor coolant inventory:
Reductions in coolant invent ory could threaten the fuel as the coolant becomes less able to remove heat generated in the core.
- g. Radioactive release from subsystems and components:
Loss of integrity of a ra dioactive containment co mponent is postulated.
- h. Anticipated transients without scram:
To determine the capability of plant de sign to accommodate an extremely low probability event, a multi-system maloperation situation is postulated.
15.0.2.1 Single Loop Operation (SLO)
Operation with one recirculati on loop results in a maximum pow er output that is 20% to 30% below that which is attainable for two-pump operation. Therefore, the consequences of abnormal operation transients from one-loop operation will be considerably less severe than those analyzed from a two-loop operational m ode because of the as sociated reduction in operation power level.
For pressurization, flow decrease, and cold water increase transients, results presented bound
both the thermal and overpressure consequences of one-loop opera tion. The consequences of flow decrease transients are also bounded by the full power analys is. A single pump trip from one-loop operation is less severe than a two-pump trip from full power because of the reduced initial power level.
Cold water increase transients can result from either recircula tion flow controller failure or introduction of colder water into th e reactor vessel by events such as loss of feedwater heater. For the former, the flow-dependent minimum critical power ratio (MCPR) values are derived assuming both recirculation loop controllers fail. This condition produces the maximum possible power increase and, hence, maximum MCPR for transients in itiated from less than rated power and flow. When operating with only one recirculation loop, the flow and power
increase associated with this failure with only one recircul ation loop will be less than that C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 15.0-5 associated with both loops; therefore, the MC PR values derived with the two-pump assumption are conservative for SLO. The latter event, loss of feedwater heating, is generally the most severe cold water increase event with respect to increase in core power. This event is caused by positive reactivity insertion fr om core inlet subcooling and it is relatively insensitive to initial power level. A generic st atistical loss of feedwa ter heater analysis using different initial power levels and other core design parameters c oncluded one-pump operation with lower initial power level is conservatively bounded by th e full power two-pump analysis. Inadvertent restart of the idle recirculation pump has been analyzed and is applicable for SLO.
From the above discussions, th e transient consequence from SLO is bounded by previously submitted full power analyses.
The maximum power level that can be attained with one-loop operation is only restricted by the MCPR and overpressure limits established from a full-power analysis.
The following most limiting transien ts of coldwater increase, pr essurization and flow decrease events are analyzed for SLO and the results are shown in Table 15.0-1A
- a. Feedwater flow controller failure (maximum demand),
- b. Generator load rejecti on with bypass failure, and c. One pump seizure accident.
15.0.3 EVENT EVALUATION
15.0.3.1 Identification of Causes and Frequency Classification
Situations and causes that lead to the initiating event analyzed are described within the analytical categories. The fr equency of occurrence of each event is summarized based on operating plant history for the transient event.
Events for which inconc lusive data exist are discussed separately within each event section.
Each initiating event within th e major groups is assigned to one of the following frequency groups:
- a. Incidents of moderate frequency - these are incide nts that may occur during a calendar year to once per lifetime. This event is referred to as an "anticipated (expected) operational transient."
- b. Infrequent incidents - these are incidents that may occur during the life of the particular plant. This event is re ferred to as an "a bnormal (unexpected) operational transient."
- c. Limiting faults - these are occurrences that are not expected to occur but are postulated because their consequences may result in the release of significant C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 15.0-6 amounts of radioactive materi al. This event is referr ed to as a "design basis (postulated) accident."
15.0.3.1.1 Unacceptable Results for Incidents of Moderate Frequency [Anticipated (Expected) Operational Transients]
The following are unacceptable safety results for incidents of moderate frequency:
- a. Release of radioactive material to the environs that ex ceeds the limits of 10 CFR 20,
- b. Reactor operation induced fuel cladding failure,
- c. Nuclear system stresses in excess of that allowed for the transient classification by applicable industry codes, and
- d. Containment stresses in excess of that allowed for the transient classification by applicable industry codes.
15.0.3.1.2 Unacceptable Results for Infrequent Incidents [Abnormal (Unexpected)
Operational Transients]
The following are unacceptable safety results for infrequent incidents:
- a. Release of radioactivity that results in dose consequences that exceed a small fraction of 10 CFR 50.67 values,
- b. Fuel damage that would preclude resumption of normal operation after a normal restart,
- c. Generation of a condition that results in consequential loss of function of the reactor coolant system, and
- d. Generation of a condition that results in a consequential loss of function of a necessary containment barrier.
15.0.3.1.3 Unacceptable Results for Limiting Faults [Design-Basis (Postulated)
Accidents]
The following are unacceptable safety results for limiting faults:
- a. Radioactive material release that resu lts in dose consequences that exceed the requirements of 10 CFR 50.67, C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 15.0-7 b. Failure of fuel cladding that would cause changes in core geometry such that core cooling woul d be inhibited,
- c. Nuclear system stresses in excess of those allowed for the accident classification by applicable industry codes, d. Containment stresses in excess of those allowed for the accident classification by applicable industry codes when containment is required, and
- e. Radiation exposure to pl ant operations personnel in the main control room in excess of 5 rem total effective dose equi valent (TEDE) for the duration of the accident.
15.0.3.2 Sequence of Events and Systems Operation
Each transient or accident is discussed and evaluated in terms of
- a. A step-by-step sequence of events from initiation to final stabilized condition (e.g., termination of the accident),
- b. The extent to which normally operating plant instrumentation and controls are assumed to function,
- c. The extent to which plant and reactor protection systems are required to function,
- d. The credit taken for the functioning of normally operating plant systems,
- e. The operation of engineered safe ty systems that is required, and
- f. The effect of a single failure or an operator error on the event.
The transient or accident discussion is specific to the event in that it is limited to the events and system operations related to the reactor core pe rformance and postulated damage. In general, the step-by-step description ends when the analys is has demonstrated that the core performance results are within established limits. The stabilized condition does not imply that all actions to stabilize plant parameters or to recover from the transient or accident have been completed by plant personnel. In the stabiliz ed condition, either the core ha s demonstrated compliance with requirements or the postulated or deterministic damage is complete. At this point, the transient or accident is terminated. After termination of the event, the operator actions or system operations are not event specific. The required actions and expected system operations, needed to establish cold shutdown or to initiate recovery actions, are symptom based and described in C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 15.0-8 procedures. The events associated with a radiological releas e and radiological consequences of the transient or accident are also discussed.
15.0.3.2.1 Single Failures or Operator Errors
15.0.3.2.1.1 General. The events considered in this section were evaluated and are provided in this chapter in accordance with Regulatory 1.70, Revision 2.
15.0.3.2.1.2 Initia ting Event Analysis. a. The undesired opening or closing of any single valv e (a check valve is not assumed to close against normal flow),
- b. The undesired starting or st opping of any single component,
- c. The malfunction or maloperation of any single control device,
- d. Any single electrical component failure, or
- e. Any single operator error.
Operator error is defined as an active devia tion from written operating procedures or nuclear plant standard operating pr actices. The set of actions is limited as follows:
- a. Those actions that could be performed by one person,
- b. Those actions that would have constituted a correct procedure had the initial decision been correct, and
- c. Those actions that are s ubsequent to the initial operato r error and have an effect on the designed operation of the plant, but are not necessarily directly related to the operator error.
Examples of single operator errors are as follows:
- a. An increase in power above the established flow control power limits by control rod withdrawal in the specified sequences,
- b. The selection and complete withdrawal of a single control rod out of sequence,
- c. An incorrect calibration of an average power range monitor (APRM), and
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 15.0-9 d. Manual isolation of the main steam lines as a result of operator misinterpretation of an alarm or indication.
15.0.3.2.1.3 Single Active Component Failure or Single Operator Error Analysis. a. The undesired action or maloperatio n of a single active component, or
- b. Any single operator error where op erator errors are defined as in Section 15.0.3.2.1.2.
15.0.3.3 Core and Sy stem Performance
Fuel thermal and hydraulic de sign are described in Section 4.4.
The fuel cladding integrity safety li mit is set so that no fuel damage is calculated to occur if the limit is not violated. Exceeding unacceptable results criteria for fuel cladding integrity for anticipated operational transients is avoided by meeting the fo llowing criteria provided in the NRC Standard Review Plan (NUREG-0800) Section 4.4:
- a. The expected number of fuel rods in boiling transition should not exceed 0.1% of the fuel rods in the core. This criterion is met by ensuring that the MCPR for any anticipated operational transient is calculated to be not less than the safety limit MCPR values given in the cycle-specific Core Operating Limits Report (COLR).
- b. No fuel centerline melting nor uniform total cladding strain in excess of 1% will occur. This criterion is met by complia nce with the operati ng limits for linear heat generation rate (LHGR) given in the cycle-specific COLR.
The operating limit for MCPR is developed as follows:
The MCPR calculated during the tran sient is compared to the safety limit. The MCPR safety limit is established using the critical power ev aluation methods and incl udes consideration of the operating domain and manufacturing uncertainties and a conservative core power distribution as inputs. The operating limit MCPR is establis hed such that the transient CPR for the dynamic anticipated operational occurre nces and quasi stea dy-state anticipated operational occurrences are included in the ev aluation. Thus, the operating limit MCPR is specified to maintain an adequate margin to boiling transition.
The MCPR operating limit is the maximum of (a) the applicable exposure dependent, full power and full flow MCPR limit, (b) the applicable exposure and power dependent MCPR limit, and (c) the flow dependent MCPR limit as specified in the cycle-specific COLR. This stipulation ensures that the safety limit MC PR will not be violated throughout the CGS C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 15.0-10 operating regime. Full power MCPR limits are specified to de fine operating limits at rated power and a range of flow conditions that support extended load line operation. Power
dependent MCPR limits are speci fied to define operating limits at other than rated power conditions. A flow dependent MC PR limit is specified to define operating limits at other than rated flow conditions.
Extended load line limit analysis (ELLLA) opera tion extends the power and flow operating regime for CGS above the rated rod line. Th e COLR defines the maximum allowable rod line for ELLLA operation. The cy cle specific Supplemental Reload Licensing Report (Reference 15.0-1) documents the reload analyses in support of ELLLA operation.
The CGS cycle-specific COLR provides the average planar linear heat generation rate (APLHGR) limits, the MCPR limits, and the linear heat gene ration rate (LHGR) limits as required by the Technical Specifications.
15.0.3.3.1 Mathematical Model
Unless otherwise stated in the Mathematical M odel description for the event being discussed, the following mathematical model was used to perform the Chapter 15 transient analyses.
Transients are analyzed using one of two transient analysis m odels described in Reference 15.0-3. The one-dimensional transient analys is model ODYN was used to analyze transients involving significant re actor pressurization (i.e., limiti ng events). The point-kinetics transient analysis model REDY was used for transients not involving significant reactor pressurization (i.e., non-limiting ev ents). The transient analysis model determines the transient pressure, power, heat flux, and average core flow which are re quired as input to both the ISCOR hot channel analysis and to the TASC transient critical pow er methodology. The ODYN transient analysis model also calculates the transient peak reactor vessel pressure to demonstrate conformance to the reactor pressu re vessel safety limit, which is based on the reactor pressure vessel design pressure.
The overall system model consis ts of a one-dimensional representation (ODYN) or point kinetics representation of the core (REDY), a nd representation of th e nuclear steam supply system including the reactor vessel, steamline, recirculation, feedwater system, recirculation control system, feedwater control system, and pr essure regulator. The main steamline model incorporates mass and momentum balances over multiple nodes allowing for the modeling of the acoustic wave phenomena presen t in the steamline dur ing transients. In addition, the model provides the capability for simulating the high pressure flooding system, the reactor core isolation cooling system, and the standby liquid control system as necessary for the event to be simulated.
The input data to the transient analysis model come from two sources: (1) the plant model or base deck, and (2) the BWR three-dimensi onal simulator PANACEA. The plant model C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 15.0-11 provides the necessary input data for the simulati on of the plant, including the plant and control systems performance characteristic
- s. The BWR three-dimensiona l simulator supplies the core state, the neutron kinetics cross sections, and other data necessary to characterize the reactor core.
For the event to be evaluated, a steady-state initializ ation is performed a nd then the parameter changes during the transient are calculated. The steady state initialization includes the recirculation loops and reactor vessel internal pressure drops, core exit pr essure, and core inlet flow and enthalpy inputs to the reactor core mode
- l. The values are used in the reactor core model to calculate the neutron kinetics, therma l-hydraulics and fuel temp erature at steady-state conditions. During the transien t, the system model calculates the time dependent response of pressure, flow, neutron flux and heat flux. Plan t control responses for sy stems such as turbine control and recirculation flow control and feedwater flow control are also calculated.
The hot channel analysis is performed using ISCOR to determine the flow distribution in the core during the transient, and to establish the flow to the limiting channels of each type in the core to be analyzed using the transient critical power methodology.
The hot channel analysis is based on the transient parameter changes pr ovided by the transien t analysis model.
The TASC transient critical power calculation me thodology is used to calculate the change in CPR from the initial CPR assumed for the transient being evaluated. This defines the delta-CPR during the transient. The GEXL transient critical power cal culation method is used to calculate CPR.
The above mathematical models de scribe the models used in the power uprate analysis. Cycle specific analyses are performed using vendor specific models for the vendor supplying the reload fuel for the current cycle as described elsewhere in this chapter.
15.0.3.3.2 Input Paramete rs and Initial Conditions for Analyzed Events
This section discusses the important input parameters used in the analysis for the event discussed. In some cases , the discussion references Table 15.0-2 (or 2A or 2B).
15.0.3.3.3 Consideration of Uncertainties
Except for total core flow and TIP reading, the uncertainties used in the statistical analysis to determine the MCPR fuel cladding integrity sa fety limit are not dependent on whether coolant flow is provided by one or two recirculati on pumps. Uncertainties used in the two-loop operation analysis are documented in the FSAR. A 6% core flow measurement uncertainty has been established for single loop operation (compared to 2.5%
for two-loop operation). This value conservatively reflects the one standard deviation (one sigma) accuracy of the core flow measurement system documented in Reference 15.0-2. In SLO, measurement and prediction uncertainties for radial power distribution and axial power distribution also increase. In the C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 15.0-12 current methodology, axial power uncertainty is not an important parameter.
The net effect of these revised uncertainties is an incremental increase in the required MCPR fuel cladding integrity safety limit. The MC PR safety limit for SLO is given in the cycle specific COLR.
15.0.3.3.3.1 Core Flow Uncertainty Analysis
The uncertainty analysis procedure used to es tablish the core flow uncertainty for one-pump operation is essentially the same as for two-pump opera tion, with some ex ceptions. The core flow uncertainty analysis is described in Reference 15.0-2. The analysis of one-pump core flow uncertainty is summarized below.
For SLO, the total core flow can be expressed as follows (see Figure 15.0-2
): WWW CAI= where:
W C=total core flow W A=active loop flow, and
W I= inactive loop (true) flow.
By applying the "propagation of errors" method to the above equa tion, the variance of the total flow uncertainty can be approximated by:
W C W sys W Arand W Irand C a a a 22 2 2 2 22 1 11=+++ where:
W C = uncertainty of total core flow; Wsys = uncertainty systematic to both loops; W Arand = random uncertainty of active loop only; W Irand = random uncertainty of inactive loop only; C = uncertainty of "C" coefficient; and C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 15.0-13 a = ratio of inactive loop flow (WI) to active loop flow (WA).
From an uncertainty analysis, the conservative, bounding values of W sys W Arand W Irand C,,, and are 1.6%, 2.6%, 3.5% and 2.8% respectively.
Based on the above uncertainties and a bounding value of 0.36
- for "a", the variance of the total flow uncertainty is approximately:
()()()()()()W C 2 2 2 2 2222 16 11036 260361036352850=+++=........% When the effect of 4.1% core bypass flow split uncertainty at 12% (bounding case) bypass flow fraction is added to the above total co re flow uncertainty, th e active coolant flow uncertainty is:
()()()2 2 2 22 5001210124151active coolant
=+=.%...%.% which is less than the 6% core flow uncer tainty assumed in the statistical analysis.
In summary, core flow during one-pump operation is measured in a conservative way and its uncertainty has been conservatively evaluated.
15.0.3.3.4 Results
This section discusses the results, in terms of core and syst em performance, of the event analyzed. The COLR provides ope rating limits that are the results of analytical evaluations that impact core operati ng parameters for the current cycle. In addition, critical parameters for the complete set of transients analyzed are shown in Table 15.0-1. From the data in Table 15.0-1 , an evaluation of the limiting event for that particular category and parameter can be made. The limiting events are reanal yzed for the current operating cycle.
Table 15.0-3 provides a summary of accid ents that may have ra diological consequences.
- This flow split ratio varies from about 0.13 to 0.36. The 0.36 value is a conservative bounding value. The analytical expected value of the flow split ratio for CGS is 0.23.
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 15.0-14 15.0.3.4 Barrier Performance
This section addresses the performance of the RCPB and the cont ainment system during transients and accidents.
During transients that occur with no release of coolant to the containment, only RCPB performance is considered. If release to the containment occurs as in the case of limiting faults, then challenges to the containment are evaluated as well.
Piping systems within the secondary containmen t structure (i.e., the reactor building) have been analyzed for pipe break effects including jet impingement, jet reac tion, pipe whip, and subcompartment pressurization. Where necessary, these loads were included in the design of the structure to ensure that the secondary containment can pe rform its required functions as defined in Section 6.2.3. 15.0.3.5 Radiological Consequences
This section addresses the radiological release consequences during the incidents of moderate frequency (anticipated operati onal transients), infrequent in cidents (abnormal operational transients), and limiting faults (design basis accidents [DBA]) events. For all events where consequences are limiting a de tailed quantitative evaluation is presented. For nonlimiting events, a qualitative evaluation is presented or the results are referenced from a more limiting or enveloping case or event.
For limiting faults (DBA), conservative assumptions considered to be acceptable to the NRC for the purpose of worst case bounding of the even t and determining the ad equacy of the plant design to meet 10 CFR 50.67 requirements are assumed. This is referred to as the "design basis analysis."
The atmospheric dispersion coefficients are presented in Tables 15.0-4 and 15.0-5. Reference will be made to these tables in the discussion of the analyses.
15.
0.4 REFERENCES
15.0-1 Supplemental Reload Licensing Repor t for Columbia (m ost recent version referenced in COLR).
15.0-2 General Electric Comp any, General Electric BWR Thermal Analysis Basis (GETAB); Data, Correlation, and De sign Application, NEDO-10958-A, January 1977.
15.0-3 GE Nuclear Energy, "WNP-2 Power Uprate Transient Analysis Task Report,"
GE-NE-208-08-0393, Revision 0, September 1993 (Proprietary).
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-003 15.0-15 15.0-4 GE-Hitachi Report, "Eva luation of Steam Flow I nduced Error Impact on the L3 Setpoint Analysis Limit,"
GEH-NE-0000-0077-46 03, December 2007.
Table 15.0-1 Results Summary of Transient Events App licable to Columbia Generating Station a Paragraph I.D. Figure I.D. Description Maximum Neutron Flux (%NBR) Maximum Dome Pressure (psig) Maximum Vessel Pressure (psig) Maximum Steam Line Pressure (psig) Maximum Core Average Surface Heat Flux (% of Initial) DCPRb,c,e Frequency Category C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORTDecember 2011LDCN-10-029 15.0-17 15.1 DECREASE IN REACTOR COOLANT TEMPERATURE 15.1.1 Loss of Feedwater Heating, Manual Flow Control 0.14 (d) 15.1.2 15.1-1 Feedwater Controller Failure, Max Demand 210 1146 1168 1145 114 0.27 (d) 15.1.3 15.1-2 Pressure Regulator Fail-Open 131 1151 1172 1151 100 <0.01 15.1.4 Inadvertent Opening of Safety or Relief Valve 104 1020 1061 1012 100 <0.01 (d) 15.1.6 RHR Shutdown Cooling Malfunction Decreasing Temperature See text (d) 15.2 INCREASE IN REACTOR PRESSURE 15.2.1 Pressure Regulator Fail-Closed 163 1188 1220 1187 106 n/a (d) 15.2.2 15.2-1 Generator, Load Rejection, Bypass-On f See text (d) 15.2.2 15.2-2 Generator Load Rejection, Bypass-Off 275 1238 1260 1235 111 0.30 (d) 15.2.3 15.2-3 Turbine Trip, Bypass-On f See text (d) 15.2.3 15.2-4 Turbine Trip, Bypass-Off 278 1238 1260 1235 111 0.30 (d) 15.2.4 15.2-5 Inadvertent MSIV Closure 206 1200 1234 1198 100 0.022 (d) 15.2.5 15.2-6 Loss of Condenser Vacuum 256 1173 1199 1166 111 0.12 (d) 15.2.6 15.2-7 Loss of Auxiliary Power Transformers 106 h 1169 1185 1166 100 <0.01 (d) 15.2.6 15.2-8 Loss of All Grid Connections 196 1173 1196 1166 106 0.079 (d) 15.2.7 15.2-9 Loss of all Feedwater Flow 106 h 1142 1152 1142 100 <0.01 (d) 15.2.8 Feedwater Piping Break See Section 15.6.6 15.2.9 Failure of RHR Shutdown Cooling See text
Table 15.0-1 Results Summary of Transient Events App licable to Columbia Generating Station a (Continued)
Paragraph I.D. Figure I.D. Description Maximum Neutron Flux (%NBR) Maximum Dome Pressure (psig) Maximum Vessel Pressure (psig) Maximum Steam Line Pressure (psig) Maximum Core Average Surface Heat Flux (% of Initial) DCPRb,c,e Frequency Category C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORTDecember 2011LDCN-10-029,11-000 15.0-18 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE 15.3.1 15.3-1 Trip of One Recirculation Pump Motor 106 h 1020 1059 1012 100 <0.01 (d) 15.3.1 15.3-2 Trip of Both Recirculation Pump Motors 106 h 1077 1088 1076 100 <0.01 (d) 15.3.2 15.3-3 Speed Decrease of One Main Recirc Motor 106 h 1020 1059 1012 100 <0.01 (d) 15.3.2 15.3-4 Speed Decrease of Two Main Recirc Motors 106 h 1061 1072 1061 100 <0.01 (d) 15.3.3 15.3-5 Seizure of One Recirculation Pump 106 h 1099 1110 1098 100 <0.01 (i) 15.3.4 Recirc Pump Shaft Break See 15.3.3 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.1.1 RWE - Refueling See text (j) 15.4.1.2 RWE - Startup See text (j) 15.4.2 RWE - At Power See text (d) 15.4.3 Control Rod Misoperation See Sections 15.4.1 and 15.4.2 15.4.4 15.4-1 Abnormal Startup of Idle Recirculation Loop 124 c 1004 1026 998 190 0.53 (d) 15.4.5 15.4-2 Speed Increase of One Main Recirc Motor 136 c 990 1009 986 127 0.15 (d) 15.4.5 Speed Increase of Both Main Recirc Motors 153 c 1006 1033 1001 149 0.27 (d) 15.4.7 Misplaced Bundle Accident See text (j) 15.4.9 Rod Drop Accident (i) 15.5 INCREASE IN REACTOR COOLANT INVENTORY 15.5.1 15.5-1 Inadvertent HPCS Pump Start f 102 h 1020 1052 1012 100 <0.01 g (d) 15.5.3 BWR Transients See appropriate events in Sections 15.1 and 15.2 a Results reflect GE14 fuel introduction, some of which are dependent on fuel design and core loading pattern. Compliance with the event acceptance criteria is demonstrated by cycle-dependent analysis of potentially limiting events just prior to the operation of that cycle. The results are reported in the Supplemental Reload Licensing Report (Reference 15.0-1).
Table 15.0-1 Results Summary of Transient Events App licable to Columbia Generating Station a (Continued)
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORTDecember 2011 15.0-19 b MCPR operating limits are based on the delta-CPR (DCPR) results from the limiting transient event and the MCPR safety limit def ined in the Technical Specifications.
c Option B DCPR results are reported.
d Moderate frequency.
e This value is only for the more limiting GE14 fuel.
f Non-limiting event under power uprate conditions (event not reanalyzed).
g ODYN results without the adjustment factors delineated in the ODYN Report NEDO-24154, NEDE-24154P.
h No increase from initial value.
i Limiting fault.
j Infrequent incident.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.0-20 Table 15.0-1A Summary of Transient Peak Value Results Single-Loop Operation Paragraph/
Figure Description Maximum Neutron Flux (% NBR) Maximum System Pressure (psig) Frequency Category Initial condition 75 1020 N/A 15.1.2/ 15.1-3 Feedwater flow controller failure (maximum demand) uprated power 89 1118 (a) 15.2.2 Generator load rejection - uprated power 131 1184 (a) 15.3.3/ 15.3-6 Seizure of active recirculation pump 75 1014 (b) a Moderate frequency incident.
b Limiting fault.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.0-21 Table 15.0-2 Input Parameters and Initia l Conditions for Transients REDY (ASD Events)
REDY a ODYN 1. Thermal power level, MWt Licensed value 3486 3323 3486 Analysis value 3702 3464 3629 2. Steam flow, lbs/hr analysis value 16.09 x 10 6 14.98 x 10 6 15.73 x 10 6 3. Core flow, lbs/hr 108.5 x 10 6 108.36 x 10 6 95.5-115.0 x 10 6 4. FW flow rate, lb/sec analysis value 4471 4161 4362 5. Feedwater temperature, °F 426 424 426 6. Vessel dome pressure, psig 1020 1020 1020 7. Vessel core pressure, psig 1031 1031 1031 8. Turbine bypass capacity, %NBR 22.7 25 22.7 9. Core coolant inlet enthalpy, Btu/lb 528.3 529.3 529.6 10. Turbine inlet pressure, psig 992 975 997 11. Fuel lattice 8 x 8/9 x 9 8x8 Simulated 8x8/9x9 12. Core average fuel cladding gap conductance, Btu/sec-ft 2-°F 0.3608 0.1667 Fuel specific 13. Core leakage flow, %
10.20 11.84 Cycle specific 14. Required MCPR operating limit (b) 1.24 (c) 15. MCPR safety limit (b) 1.06 (c) 16. Doppler coefficient (-)¢/° Nominal EOC-1 0.311 0.227 (d) Analysis data ASD events 1. Increase power
- 2. Decrease power
0.295 0.327 0.215 17. Void coefficient (-)¢/% Rated Nominal EOC-1 7.48 (d) Analysis data for power increase events 15.93 12.70 (d) Analysis data for power decrease events 12.10 7.065 (d) 18. Core average rated void fraction, % (Steady state) 41.24 41.32 43.1 19. Scram reactivity, $k analysis data Figure 15.0-1 Figure 15.0-1 (d) 20. Control rod drive speed, position versus time Figure 15.0-1 Figure 15.0-1 Figure 15.0-1 21. Jet pump ratio, M 2.36 2.41 2.39 C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-003 15.0-22 Table 15.0-2
Input Parameters and Initial Conditions for Tr ansients (Continued)
REDY (ASD Events)
REDY a ODYN 22. Safety/relief valve capacity,% NBR safety valve capacity @ 1241 psig 108.6 111.5 108.6 Relief valve capacity @ setpoint values in item 25 of this table @ 1121 psig 98.3 101.8 98.3 @ 1131 psig 99.1 102.8 99.1 @ 1141 psig 100.0 103.7 100 @ 1151 psig 100.9 104.6 100.9 @ 1161 psig 101.7 105.5 101.7 Manufacturer Crosby Crosby Quantity installed 18 18 23. Relief function delay, sec 0.4 0.4 0.4 24. Relief function response, sec 0.15 0.1 0.15 25. Setpoints for safety/relief valves Safety function, psig
1200, 1210
1177, 1187, 1200, 1210 1221, 1231 1197, 1207, 1221, 1231 1241 1217 1241 Relief function, psig 1121, 1131 1091, 1101, 1121, 1131 1141, 1151 1111, 1121, 1141, 1151 1161 1131 1161 26. Number of valve groupings simulated Safety function, number 5 5 Relief function, number 5 5 27. High flux trip analysis setpoint (123 x 1.041),% NBR 128.0 126.20 128 e 28. High pressure scram setpoint, psig 1086 1071 1086 29. Vessel level trips, inches with respect to dryer skirt bottom Level 8 - (L8), in.
59.5 55.5 59.5 Level 4 - (L4), in.
31.5 30 Level 3 - (L3), in.
7.5 i 12.5 i (f) Level 2 - (L2), in.
(-38) (f) 30. APRM thermal trip analysis setpoint (117 x 1.041)% NBR @ 100% core flow 121.8 122.030 121.8 e 31. Recirculation pump trip delay, sec 0.190 0.140 0.190 32. Recirculation pump trip inertia time constant for analysis, sec 6 g 6 g 6 g 33. RPS response time delay (h) (h) (h)
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-003 15.0-23 Table 15.0-2
Input Parameters and Initial Conditions for Tr ansients (Continued) a REDY values reflect the pre-uprate initial cond itions. Only limiting events were analyzed for power uprate conditions with FCV.
b See COLR.
c Not applicable to reload 7/cycle 8 simulation.
d ODYN values are calcu lated within the code.
e The thermal multiplier (1.041 = 3629/3486) is us ed to give a conservative margin that is proportional to the core power.
f Parameter not used in the analysis.
g The inertia time constant is defined by the expression:
t Jn gT O O2 where t = inertia time constant (sec)
J o = pump motor inertia (lb-ft
- 2) n = rated pump speed (rps) g = gravitational constant (ft/sec
- 2) T o = pump shaft torque (lb-ft) h The "maximum overall response time" as addressed in the LC S is utilized for each scram encountered in the Chapter 15 events. i The impact of steam flow induced error on the analytical limit does not impact event descriptions or conclusions (Reference 15.0-4).
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.0-24 Table 15.0-2A Input Parameters and Initia l Conditions for Transients and Accidents for Single-Loop Operation Original Rated Power Uprated Power 1. Thermal power analysis value (MWt) 2596.9 2615 2. Flow Steam (1b/hr) 10.79 x 10 6 10.76 x 10 6 Core (lb/hr) 59.0 x 10 6 61.85 x 10 6 Core bypass (lb/hr) 5.88 x 10 6 6.22 x 10 6 Feedwater (lb/hr) 10.79 x 10 6 10.76 x 10 6 Turbine bypass (lb/hr) 5.88 x 10 6 N/A Turbine bypass (% rated) N/A 23% 3. Core Inlet Enthalpy (Btu/lb) 510.8 513.7
- 4. Pressure Vessel dome (psia) 1020 1008 Vessel core (psia) 1029.7 1017.7 Turbine inlet (psia) a 960.5 1000 5. Jet pump ratio (M) 3.2 3.4 6. Safety/relief valve capacity % NBR @ 1,164 psig 107.1 N/A Manufacturer Crosby Crosby Quantity installed 18 18 % NBR @ 1241 psig N/A 108.6 7. Relief function Delay (sec) 0.4 0.4 Response (sec) 0.1 0.15 C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-003 15.0-25 Table 15.0-2A
Input Parameters and Initia l Conditions for Transients and Accidents for Single-Loop Operation (Continued)
Original Rated Power Uprated Power 8. Setpoints for safety/relief valves Safety function (psig) 1177, 1187, 1197, 1207, 1217 1200, 1210, 1221, 1231, 1241 Relief function (psig) 1106, 1116, 1126, 1136, 1146 1121, 1131, 1141, 1151, 1161 9. Number of valve groupings simulated Safety function (number) 5 5 Relief function (number) 5 5
- 10. Setpoints High flux trip anal ysis (1.21 x 1.043) (% NBR) 126.2 128 High pressure scram (psig) 1071 1086 APRM thermal trip (% NBR @
100% core flow) 122.03 121.8 11. Vessel level trips (ft above instrument zero) Level 8 - (L8) (ft) 4.542 Level 4 - (L4) (ft) 2.625 Level 3 - (L3) (ft) 1.083 b Level 2 - (L2) (ft) (-)4.167
- 2) 24,500 N/A 13b. RPT inertia time constant (sec) N/A 6 a Pressure specified at rated power condition. Off-rated power pressure drop is calculated by transient analysis code.
b The impact of steam flow induced error on the analytical limit does not impact event descriptions or conclusions (Reference 15.0-4).
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.0-26 Table 15.0-2B Input Parameters a nd Initial Conditions for GNF Reload Transient Analyses Parameter Value 1. Thermal power level, MWt 3486 2. Steam flow, lbs/hr analysis value 15.01 x 10 6 3. Core flow, lbs/hr 95.5 - 115.0 x 10 6 4. FW flow rate, lb/sec analysis value 4161.6 5. Feedwater temperature, °F 421.2 6. Vessel dome pressure, psig 1020 7. Vessel core pressure, psig 1032 8. Turbine bypass cap acity, %NBR 23.75 9. Core coolant inlet enthal py, Btu/lb (rated flow) 527.2 10. Turbine inlet pressure, psig 978 11. Fuel lattice 10 x 10 mixed core 12. Required MCPR operating limit See COLR 13. MCPR safety limit See COLR 14. Control rod drive speed, position versus time a 15. Jet pump ratio, M 2.285 16. Safety/relief valve capacity, % NBR safety valve capacity See Table 5.2-3 Manufacturer Crosby Quantity installed 18 17. Relief valve function delay, sec 0.4 response, sec 0.15 18. Safety valve function delay, sec 0.0 response, sec 0.3
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-003 15.0-27 Table 15.0-2B
Input Parameters and Initial Conditions for GNF Reload Transient (Continued)
Parameter Value 19. Setpoints for safety/relief valves Safety function, psig # in Group Setpoint 2 b 1200 4 b 1210 4 1221 4 1231 4 1241 Relief function, psig 2 b 1156 4 b 1166 4 1176 4 1186 4 1196 20. Number of valve groupings simulated Safety function, number (actual/credited) 5 / 3 Relief function, number (actual/credited) 5 / 3 21. High flux trip analysis setpoint,% NBR 123.0 22. High pressure scram setpoint, psig 1086 23. Vessel level trips, inches with respect to instrument zero Level 8 - (L8), in.
59.5 Level 4 - (L4), in. 30 Level 3 - (L3), in. 2.5 c Level 2 - (L2), in. -90 LOCA
-70 Non LOCA 24. APRM thermal trip analysis setpoint Not credited 25. Recirculation pump trip delay, sec 0.200 26. RPS response time delay See LCS a See Section 4.6.1.1.2.5.3. b Valve group function not cred ited in safety analyses.
c This allows the correction due to steam flow induced error (Reference 15.0-4).
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.0-28 Table 15.0-3 Summary of Accidents Paragraph I.D. Title Failed Fuel Calculated Value 15.3.3 Seizure of one recirculation pump None 15.3.4 Recirculation pump shaft break None 15.4.9 Rod drop accident 850 rods 15.6.2 Instrument line break None 15.6.4 Steam system pipe break outside containment None 15.6.5 Loss-of-coolant accident within RCPB 100%
15.6.6 Feedwater line break None 15.7.1 Main condenser gas treatment system failure N/A 15.7.3 Liquid radwaste tank failure N/A 15.7.4 Fuel handling accident 250 rods 15.8 Anticipated transients without scram None
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.0-29 Table 15.0-4
/Q (s/m 3) values for the EAB and LPZ Time Period EAB /Q (s/m 3) LPZ /Q (s/m 3) 0 - 2 hrs 1.81 E-4 4.95 E-5 2 - 8 hrs 4.95 E-5 8 - 24 hrs 3.69 E-5 1 - 4 d 1.95 E-5 4 - 30 d 7.81 E-6 C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.0-30 Table 15.0-5 Control Room Atmospheric Dispersion Factors (sec/m 3) Hours Turbine Building Secondary Containment SGT System Release Filtered Intake Release Path 0 - 2 2 - 8 8 - 24 24 - 96 96 - 720 8.81E-4 3.75E-4 1.93E-4 1.50E-4 1.44E-4 2.82E-4 2.17E-4 8.77E-5 7.42E-5 6.40E-5 1.43E-4 1.05E-4 4.14E-5 3.52E-5 3.03E-5 Unfiltered Intake Release Path 0 - 2 2 - 8 8 - 24 24 - 96 96 - 720 4.70E-3 2.00E-3 1.03E-3 8.01E-4 7.69E-4 7.02E-4 3.19E-4 1.30E-4 1.05E-4 9.00E-5 6.95E-4 3.36E-4 1.28E-4 9.72E-5 7.69E-5
Scram Position and Reactivity Characteristics 900547.61 15.0-1 Figure Amendment 53 November 1998 Form No. 960690 Draw. No.Rev.1.0 0.8 0.6 0.4 0.2 0012345678 40 30 20 10 0$40.21$37.05 Scram Curve Used in Analysis 67B Scram Rod Drive Control Fraction (Per Unit Insertion)
Time (sec)
End of Cycle 1 Scram Curve 67 Scram Analysis EOC 1 Scram Curve
Revised Portion of Scram Speed Used in ODYN Analysis Scram Reactivity ($)
Columbia Generating Station Final Safety Analysis Report 15.0-2 Amendment 58 December 2005 Illustration of Single Recirculation Loop Operation Flows 910402.42 W C W I W A Core W C - Total Core Flow W A - Active Loop Flow W - Inactive Loop Flow Columbia Generating Station Final Safety Analysis ReportDraw. No.Rev.Figure Form No. 960690FH LDCN-05-019 C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-07-011 15.1-1 15.1 DECREASE IN REACTOR COOLANT TEMPERATURE
15.1.1 LOSS OF FEEDWATER HEATING
15.1.1.1 Identification of Causes and Frequency Classification
15.1.1.1.1 Identification of Causes
A feedwater heater can be lost in at least two ways:
- a. Steam extraction line to heater is closed, and b. Steam is bypassed around heater.
The first case produces a gradual cooling of th e feedwater. In the second case, the steam bypasses the heater and no heating of that feedwater occurs. In either case, the reactor vessel receives cooler feedwater. Th e maximum number of feedwater heaters that can be tripped or bypassed by a single event represents the most severe transient for analysis considerations. This event has been conservatively estimated to incur a loss of up to 100°F of the feedwater heating capability of the plant and causes an increase in core inlet subcooling. This increases core power due to the negative void reactivity coefficient.
15.1.1.1.2 Frequenc y Classification
This event is considered to be an incident of moderate fre quency and is analyzed under worst case conditions of a 100°F loss at full power.
15.1.1.2 Sequence of Events and Systems Operation
The loss of feedwater hea ting leads to a gradual decrease in the temperature of the feedwater entering the reactor vessel. The decrease in feedwater temperature results in an increase in the core inlet subcooling which co llapses voids, and increases th e core average power. The gradual power change allows fuel thermal response to maintain pace with the increase in neutron flux. For this analysis, it was assume d that the initial feed water temperature dropped 100°F. In establishing the expected sequence of events and simulating the plant performance, it was assumed that normal functioning occurred in th e plant instrumentation and controls, plant protection, and reactor protection systems. Engineered safety feature (ESF) system initiation is not anticipated or required to pr event or mitigate the transient.
The Average Power Range Monitor (APRM) Flow Biased Simulated Thermal Power trip setpoint provides protection against transients such as the Loss of Feedwater Heating where thermal power increases slowly. While the sequence of events may produce sufficiently high C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 15.1-2 flux levels to initiate an APRM reactor protection system trip, no credit is taken for a reactor trip in the analysis of the even
- t. A description of the APRM system and operation is provided in Sections 7.2.1.1.1.2 and 7.6.1.4.3.
15.1.1.2.1 The Effect of Single Failures and Operator Errors
The loss of feedwater heating generally leads to an increase in reactor power level. The APRM system is the mitigating system and is de signed to be single failu re proof. Therefore, single failures are not expected to result in a more severe event than analyzed.
15.1.1.3 Core and Sy stem Performance
15.1.1.3.1 Mathematical Model
The analytical dynamic behavior has been determined using the steady state boiling water reactor (BWR) simulator code PANACEA (Reference 15.1-2). This code does not provide plots of the dynamic behavior of basic parameters as a function of time nor does it provide information for a sequence of events table. Therefore, no figures or tables are available.
Reference 15.1-6 approves the use of PANACEA for modeling the Loss of Feedwater Heating event.
The loss of feedwater heating (LF WH) event analysis supports an assumed 100°F decrease in the feedwater temperature. The result is an increase in core inlet subcoo ling, which collapses voids, thereby, increasing the core power and sh ifting the axial power distribution toward the bottom of the core. As a result of the axial power shift and increased core power, voids begin to build up at the botto m of the core, acting as negative feedback to the void collapse process. The negative feedback moderates the core power increase.
The PANACEA code is used to determine the change in minimu m critical power ratio (MCPR) during the event. Analyses were performed for a range of cycle exposures to ensure that appropriate limits are set.
Although there is a substantial increase in core thermal power during the event, the increase in steam flow is much less because a large part of the added power is used to overcome the increase in inlet subcooling. The increase in steam flow is accommodated by the pressure control system by the turbine control (governor) valves or the turbine bypass valves, so no pressurization occurs (Reference 15.1-3). 15.1.1.3.2 Input Paramete rs and Initial Conditions These analyses have been perfor med, unless otherwise noted, with plant conditions tabulated in Table 15.0-2B.
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 15.1-3 15.1.1.3.3 Results
The LFWH transient is analyzed for each reload core to quantify the reduction in thermal margins. The results of the an alysis are provided in the cycl e specific Supplemental Reload Licensing Report (Reference 15.1-3).
15.1.1.3.4 Considerations of Uncertainties
Factors such as exposure and magn itude of feedwater temperature change are assumed to be at the worst configuration so that any deviations seen in the actual plant operation reduce the severity of the event.
15.1.1.4 Barrier Performance
The consequences of this event do not result in any temperature or pressure transient in excess of the criteria for which the fu el, pressure vessel, or containm ent are designed. Therefore, barrier integrity and function is maintained.
15.1.1.5 Radiological Consequences
Since this event does not result in any additional fuel failures or any release of primary coolant to either the secondary containment or to the environment, there are no radiological consequences associated with this event.
15.1.2 FEEDWATER CONTROLLER FAILURE - MAXIMUM DEMAND
15.1.2.1 Identification of Causes and Frequency Classification
15.1.2.1.1 Identification of Causes
This event is postulated on the basi s of a single failure of a control device, specifically one that can directly cause an increase in coolant inventory by increa sing the feedwater flow. The most severe applicable event is a feedwater controller failure (FWCF) during maximum flow demand. The feedwater controller is forced to its upper limit at the beginning of the event. The event is evaluated for both single and two reactor recirculation l oop operations. Because the two-loop operation event is bounding, the core performance analysis is limited to the feedwater controller failure during two-loop op eration. However, the MCPR operating limit for single loop operation (SLO) is obtained by adding the CPR from two-loop operation to the MCPR safety limit (SLMCPR) for SLO.
15.1.2.1.2 Frequenc y Classification
This event is considered to be an incident of mode rate frequency.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 15.1-4 15.1.2.2 Sequence of Events and Systems Operation
The increase in feedwater flow, due to a failur e of the feedwater control system to maximum demand, results in an increase in the water level and a decrease in the coolant temperature at the core inlet. The increase in core inlet subcooling causes an increase in core power. As the feedwater flow continues at maxi mum demand, the water level con tinues to rise and eventually reaches the high water level trip setpoint. The initial water level is c onservatively assumed to be at the low level normal opera ting range of 30 inches above instrument zero to delay the high-level trip and maximize the core inlet subcooling that resu lts from the FWCF. The high water level trip causes th e turbine throttle (stop) valves to close in orde r to prevent damage to the turbine from excessive liquid inventory in the steam line. The va lve closures create a compression wave that travels to the core cau sing a void collapse and su bsequent rapid power excursion. In addition to the turbine throttle valve closure, the turbine governor valves also close in the fast closure mode.
The closure of the governor (cont rol) turbine valves initiates a reactor scram and a recirculation pump trip.
Because of the partially opened initial position of the governor valves, they will clos e faster than the th rottle valves and initia te the pressurization portion of the event. The tu rbine bypass valves are assume d operable and provide some pressure relief. The core power excursion is mitigated in part by the pressure relief, but the primary mechanisms for termination of the event are reactor sc ram and revoiding of the core.
The high-pressure core spray (HPCS) system and reactor core isolation cooling (RCIC) system initiate on a low reactor water level (L2) to maintain long-term water level control following tripping of feedwater pumps. The analysis of this event assumes norma l functioning of plant instrumentation and controls, plant protection and reactor protection systems.
Table 15.1-1 lists the sequence of events for Figure 15.1-1. The figure shows the changes in variables during this transient.
15.1.2.2.1 Sequence of Events and Syst ems Operation - Single Loop Operation
The simulated feedwater contro ller transient is shown in Figure 15.1-3 for the case of 75% power, 57% core flow. The high-water level turbine trip and feedwater pump trip are initiated at approximately 8.4 sec. A scram occurs simultaneous ly with the turbine trip and limits the neutron flux peak and fuel thermal transient so no fuel damage occurs.
Table 15.1-1A lists the sequen ce of events for Figure 15.1-3. The figures show the changes in important variables during this transient.
Identification of Operator Actions
- a. Observe high feedwater pump trip ha s terminated the failure event, C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 15.1-5 b. Switch the feedwater controller from auto to manual control to try to regain a correct output signal, and
- c. Conduct follow-up assessment.
15.1.2.2.2 The Effect of Single Failures and Operator Errors
The first sensed event to initiate corrective action to the transient is the vessel high water level (L8) trip. Multiple level sensors are used to sense and detect when the water level reaches the L8 setpoint. At this point in the logic, a single failure will not initiate or prevent a turbine trip signal. Turbine trip signal transmission, however , is not built to single failure criterion. The result of a failure at this point would have the effect of delaying, but not impacting, the pressurization "signature."
Scram trip signals from the turbine are designed such that a single failure will neither initiate nor impede a reactor scram trip initiation.
15.1.2.3 Core and Sy stem Performance
15.1.2.3.1 Mathematical Model
The predicted dynamic behavior has been determined using a computer simulated, analytical model of a direct-cycle BWR. This model is described in detail in Reference 15.1-4. Results from the two-loop operation bound the SLO event.
Therefore, the discussion of core and system performance is limited to the descri ption of the analysis for two-loop operation.
The nonlinear computer simulated analytical model is designed to predict associated transient
behavior of the reactor. Some of the significant features of the model are the following:
- a. An integrated one-dimensional core m odel is assumed which includes a detailed description of hydraulic feedback effe cts, axial power shape changes, and reactivity feedbacks;
- b. The fuel is represented by an average cylindrical fuel and cladding model for each axial location in the core;
- c. The steam lines are modeled by eight pressure nodes incorporating mass and momentum balances which will predict any wave phenomena present in the steam line during a pressurization transient;
- d. The core average axial water density and pressure distribution is calculated using a single channel to represent the heated active flow and a single channel to represent the bypass flow. A model, representing liquid and vapor mass and C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-029 15.1-6 energy conservation and mixt ure momentum conservation, is used to describe thermal-hydraulic behavior. Changes in the flow split between the bypass and active channel flow are accounted for during transient events;
- e. Principal controller functi ons such as feedwater flow, recirculation flow, reactor water level, pressure, an d load demand are represented together with their dominant nonlinear characteristics; and
- f. The ability to simulate necessary reactor protect ion system functions is provided.
15.1.2.3.2 Input Paramete rs and Initial Conditions
These analyses have been perfor med with the plant conditions in Table 15.0-2B.
All rods out scram characteristics are assumed. The safety/relief valve (SRV) action is conservatively assumed to occur w ith higher than nominal setpoints.
The transient is simulated by programming an upper limit failure in the feed water system such that 139% feedwater flow occurs at the nominal reactor operating pressure of 1035 psia.
An increase in feedwater flow will cause a corresponding drop in feedwater temperature. However, the relatively large time constant of the feedwater heater s (order of minutes) plus the flow transport time (10 sec from heaters to vessel and 3 sec from sparger to core) would preclude any effect of temperatur e reduction on the transient since the transient is essentially over in about 20 sec. Therefore, feedwater temperature is assumed to remain constant.
15.1.2.3.3 Results
The simulated feedwater contro ller transient is shown in Figure 15.1-1. The high water level turbine trip and feedwater pump trip are initiated as stated in Table 15.1-1. Results reflect GE14 fuel introduction, some of which are dependent on fuel de sign and core loading pattern.
Compliance with the event acceptance criteria is demonstrat ed by cycle-dependent analysis of potentially limiting events just prior to the operation of that cycl
- e. The results are reported in the Supplemental Reload Li censing Report (Reference 15.1-3). Because the total change in f eedwater flow is greatest from reduced power conditions, the feedwater controller failure (FWCF) transient wa s analyzed for several reduced power states.
The power dependent MCPR limits are established to protect the fuel during the FWCF event.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.1-7 15.1.2.3.4 Consideration of Uncertainties
All systems used for protection in this event were assumed to have the most conservative response characteristics. Therefore, actual plant behavior is exp ected to lead to a less severe transient.
15.1.2.4 Barrier Performance
The consequences of this event do not result in any temperature or pressure transient in excess of the criteria for which the fu el, pressure vessel, or containm ent are designed. Therefore, barrier integrity and function is maintained.
15.1.2.5 Radiological Consequences
The consequence of this event does not result in fuel failure. It does result in the discharge of normal coolant activity to the suppression pool by means of SRV operation, which is contained in the primary containment.
This event does not result in an uncontrolled release to the environment, so the plant operator can choose to hold the activity in containment or discharge it when conditions permit. If purging of the containment is chosen, th e release would be in accordance with established requirements.
15.1.3 PRESSURE REGULATOR FAILURE - OPEN
15.1.3.1 Identification of Causes and Frequency Classification
15.1.3.1.1 Identification of Causes
The total steam flow rate to the main turbine resulting from a pr essure regulator malfunction in the Digital Electro-Hydraulic (DEH) control system is limited by a maximum flow limiter imposed at the turbine controls. This limiter is set to limit maximu m steam flow demand to approximately 130% NBR.
If the triple redundant DEH control system fails such that the turbine control (governor) valves fully open and the turbine bypass valves partially open, the maximum steam flow is
established.
15.1.3.1.2 Frequenc y Classification
This event is categorized as an incident of mode rate frequency.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.1-8 15.1.3.2 Sequence of Events and Systems Operation 15.1.3.2.1 Sequence of Events
Table 15.1-2 lists the sequence of events for Figure 15.1-2. Figure 15.1-2 depicts how the high water level turbine trip a nd isolation valve closure stops vessel depressurization and produces a normal shutdown of the reactor.
15.1.3.2.2 Systems Operation Depressurization results in formation of voids in the reactor coolant and causes a decrease in reactor power almost immediately. In this simu lation, the depressurizati on rate is large enough such that water level swells to the sensed level trip setpoi nt (L8), initiating main turbine and feedwater turbine trips. Position switches on the turbine stop (throttle) va lves initiate a reactor scram and RPT and shut down the reactor. After the turbine trip, the failed DEH control
system signals the bypass to ope n to full bypass flow of 25% NBR steam flow. After the pressurization resulting from the turbine stop (throttle) valve clos ure, the pressure increase opens the relief valves and pressure drops and continues to drop until turbine inlet pressure is below the low turbine pressure isolation setpoi nt when main steam line isolation limits the duration and severity of the depressurization.
In order to properly simulate the expected sequence of events, the analysis of this event assumed normal functioning of plant instrumentation and controls, plant protection, and reactor protection systems except as otherwise noted.
Initiation of HPCS and RCIC syst em functions will occur when the vessel water level reaches the L2 setpoint although this is not included in the analysis.
Normal startup and actuation can take up to 30 sec before effects are realized.
If these events o ccur, they will follow some time after the primary concerns of fuel thermal margin and overpressure effe cts have occurred and are expected to be less severe than those already experienced by the system.
15.1.3.2.3 The Effect of Single Failures and Operator Errors
This transient leads to a loss of pressure control such that the increased steam flow demand causes a depressurization. Instrumentation for pre ssure sensing of the turbine inlet pressure is designed to be single failure proof for initiation of MSIV closure.
Reactor scram sensing, originating from limit switches on the MSIVs, is designed to be single failure proof. It is, therefore, concluded that the basi c phenomenon of pre ssure decay is adequately terminated.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.1-9 15.1.3.3 Core and Sy stem Performance 15.1.3.3.1 Mathematical Model
The point-kinetics REDY model described in Section 15.0.3.3.1 is used to simulate this event.
15.1.3.3.2 Input Paramete rs and Initial Conditions
This transient is simulated by setting the DEH control system demand signal to a high value, which causes the turbine control (governor) valves to open fully and the turbine bypass valves to open partially. A DEH control failure with 130% steam flow dema nd signal was simulated as a worst case since 130% is the normal maximum flow limit in order to conform with Table 15.1-2. A 5-sec isolation valve closure is assumed when the turbine pressure decreases below the turbine inlet low pressure setpoint for main steam line isolation initiation.
Reactor scram is initiated when the isolation valv es reach the 10% closed position. This is the maximum travel from the full ope n position allowed by specification.
This analysis has been perfor med, unless otherwise noted, with the plant conditions listed in Table 15.0-2.
15.1.3.3.3 Results
Results are summarized in Table 15.0-1.
No significant reductions of fuel thermal margins occur. No significant thermal stresses are imposed on the reactor coolant pressure boundary (RCPB).
15.1.3.3.4 Consideration of Uncertainties
If the maximum flow limiter were set higher or lower than normal, a faster or slower loss in nuclear steam pressure would result. The ra te of depressurization may be limited by the bypass capacity, but it is unlikely. For example, the turbine valves will open to the valves-wide-open state admitting sl ightly more than the rated st eam flow, and with the limiter in this analysis set to fail at 130%, it is expected that less than 25% would be bypassed. This is, therefore, not a limiting factor for the plant. If the rate of depressurization does change, it will be terminated by the low turbin e inlet pressure trip setpoint.
Depressurization rate has a proportional effect upon the voiding action in the core and the flashing in the vessel bulk water regions. If the rate is low enough, th e water level may not swell to the high water level trip setpoint and the isolation will occur earlier when pressure at
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.1-10 the turbine decreases below 850 psia. The reactor will scram as a result of the MSIV closure.
Since power is being depressed as the pressure decreases (due to a dditional voiding in the core), this transient is less severe when a slower depressurization rate is assumed. Therefore, the assumed L8 trip provides the most restrictive margins on MCPR and peak vessel pressure.
15.1.3.4 Barrier Performance
Barrier performance analyses we re not required since the consequences of this event do not result in any temperature or pressure transient in excess of the criteria for which fuel, pressure vessel, or containment are designed. Peak pressure in the bottom of the vessel is below the ASME code upset limit for the RCPB.
15.1.3.5 Radiological Consequences
The consequence of this event does not result in fuel failure. It does result in the discharge of normal coolant activity to the suppression pool by means of SRV operation, which is contained in the primary containment.
This event does not result in an uncontrolled release to the environment, so the plant operator can choose to hold the activity in containment or discharge it when conditions permit. If purging of the c ontainment is chosen, th e release would be in accordance with established requirements.
15.1.4 INADVERTENT SAFETY/RELIEF VALVE OPENING
The event is defined as the ina dvertent opening of an SRV which stays in the "open" position.
It was determined that this event is not limiting from a core performance standpoint.
15.1.4.1 Identification of Causes and Frequency Classification
15.1.4.1.1 Identification of Causes
Cause of inadvertent opening is attributed to malfunction of th e valve or an operator initiated opening. Opening and closing circuitry at the individual valv e level (as opposed to groups of valves) is subject to a single failure impact. It is therefore simply postulated that a failure occurs and the event is analyzed accordingly. Detailed di scussion of the valve is provided in Section 5.2.2. 15.1.4.1.2 Frequenc y Classification
This event is categorized as an incident of mode rate frequency.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.1-11 15.1.4.2 Sequence of Events and Systems Operation 15.1.4.2.1 Sequence of Events
Table 15.1-3 lists the sequence of events.
15.1.4.2.2 Systems Operation
In this transient, the analysis assumes normal functioning of plant instrumentation and controls, specifically, the relief valve discharge line temperature se nsors and the suppression pool temperature sensors and reactor pressure vessel level control systems. The opening of an SRV allows steam to be discharged into the suppression pool. The sudden increase in the rate of steam flow leaving the reactor vessel causes a mild depressurization transient. The pressure regulator senses the nuclear sy stem pressure decrease and with in a few seconds closes the turbine control (governor) valve far enough to stabilize reactor ve ssel pressure at a slightly lower value and reactor power settles at nearly the initial power level. Additionally, although not credited in the analysis to mitigate the consequences of this transient, minimum reactor and plant protection systems, emergency core c ooling system flow, a nd RHR suppression pool cooling, are available.
15.1.4.2.3 The Effect of Single Failures and Operator Errors
From a core performance standpoint, a single failure or operator error would simply activate the reactor protection system resulting in a plant shutdown. A single failure or operator error cannot increase the seve rity of this event.
The instrumentation which detects and audibly alarms the resulting suppression pool temperature rise, and the RHR containment heat removal system are designed to meet the single failure criteria. The operator must manually initi ate suppression pool cooling.
15.1.4.3 Core and Sy stem Performance
15.1.4.3.1 Mathematical Model
The one-dimensional ODYN m odel described in Section 15.0.3.3.1 is used to simulate this event.
15.1.4.3.2 Input Paramete rs and Initial Conditions
These analyses have been perfor med, unless otherwise noted, with plant conditions tabulated in Table 15.0-2 , the ODYN column. A discussion of the SRV is provided in Section 5.2.2.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.1-12 15.1.4.3.3 Results
Thermal margins decrease only slig htly through the tran sient, and no fuel da mage results from the transient. The MCPR is essentially unchanged and, therefore, the safety limit margin is unaffected.
15.1.4.4 Barrier Performance
The transient resulting from a stuc k open relief valve is a mild de pressurization which is within the range of normal load following. Therefore, there is no significant effect on RCPB and
containment design pressure limits.
Since quenchers are used as steam discharge devices on the steam relief lines, no unstable condensation oscillations are expected which could damage the containment vessel. This is discussed in Appendix 3A.
Therefore, barrier integrity and function is maintained.
15.1.4.5 Radiological Consequences
While the consequence of this event does not result in fuel failure, it does result in the discharge of normal coolant activity to the suppression pool by means of SRV operation. Since this activity is contained in the primary containment there will be no exposures to operating personnel. Since this event does not result in an uncontrolled release to the environment the plant operator can choose to hold the activity in containment or discharge it to the environment when conditions permit. If pur ging of the containment is chosen, the release would be in accordance with established requirements.
15.1.5 SPECTRUM OF STEAM PIPING FAILURES INSIDE AND OUTSIDE OF CONTAINMENT IN A PRESSURIZED WATER REACTOR
This event is not applicable to BWR plants.
15.1.6 INADVERTENT RESIDUAL HEAT REMOVAL SHUTDOWN COOLING OPERATION This transient is classified as a nonlimiting event for both original and uprated power conditions. Therefore, no further analysis has been performed.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.1-13 15.1.6.1 Identification of Causes and Frequency Classification 15.1.6.1.1 Identification of Causes
At design power conditions no conceivable malfunction in the s hutdown cooling system could cause temperature reduction.
In startup or cooldown operation, where the reactor is at or near critical, a very slow increase in reactor power could result. A shutdown cooling malfunction lead ing to a moderator temperature decrease could result from misoperation of the coo ling water controls for the RHR heat exchangers. The resulting temperature de crease would cause a slow insertion of positive reactivity into the core.
If the operator did not act to cont rol the power level, a high neutron flux reactor scram would termin ate the transient without viol ating fuel thermal limits and without any measurable increase in nuclear system pressure.
15.1.6.1.2 Frequenc y Classification
This event is categorized as an incident of mode rate frequency.
15.1.6.2 Sequence of Events and Systems Operation
15.1.6.2.1 Sequence of Events
A shutdown cooling malfunction leading to a moderator temperature decrease could result from misoperation of the cooling wa ter controls for RHR heat exchangers. The resulting
temperature decrease causes a slow insertion of positive reactiv ity into the core. Scram will occur before any thermal limits are reached if the operator does not take action. The sequence of events for this event is shown in Table 15.1-4.
15.1.6.2.2 System Operation
A shutdown cooling malfunction cau sing a moderator temperature d ecrease must be considered in all operating states. However, this event is not considered while at power operation since
pressure is too high to permit operation of RHR s hutdown cooling.
No unique safety actions are required to avoid unacceptable safety result s for transients as a result of a reactor coolant temperature decrease induced by misope ration of the shutdown cooling heat exchangers. In startup or cooldown operation, where the reactor is at or near critical, the slow power increase resulting from the cooler moderator temperature would be controlled by the operator in the same manner normally used to control power in the source or intermediate power ranges.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.1-14 15.1.6.2.3 Effect of Single Fa ilures and Operator Action No single failures can cause this event to be more severe.
If the operator takes action, the slow power rise will be controlled in the normal manner. If no operator action is taken, a scram will terminate the power increase before thermal limits are reached.
15.1.6.3 Core and Sy stem Performance
The increased subcooling caused by misoperation of the RHR shutdown cooling mode could result in a slow power increase due to the reactivity insertion. This power rise would be terminated by a flux scram before fuel thermal limits are approached. Therefore, only a qualitative descripti on is provided here.
15.1.6.4 Barrier Performance
The consequences of this event do not result in any temperature or pressure transient in excess of the criteria for which the fu el, pressure vessel, or containm ent are designed. Therefore, barrier integrity and function is maintained.
15.1.6.5 Radiological Consequences
Since this event does not result in any fuel failures, no analysis of radiological c onsequences is required for this event.
15.
1.7 REFERENCES
15.1-1 For Power Uprate: GE Nuclear En ergy, "WNP-2 Power Uprate Transient Analysis Task Report," GE-NE-208 0393, September 1993 (Proprietary).
15.1-2 NEDE-30130-P-A, "Steady State Nuclear Methods," April 1985.
15.1-3 Supplemental Reload Licensing Repor t for Columbia (m ost recent version referenced in COLR).
15.1-4 NEDC-24154-P-A, "Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors," Volume s 1, 2, 3 and 4, February 2000.
15.1-5 GE Nuclear Energy, "WNP-2 Powe r Uprate Project NSSS Engineering Report," GE-NE-208-17-0993, Re vision 1, December 1994.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-003 15.1-15 15.1-6 "General Electric Standard Appli cation for Reactor Fuel," NEDE-24011-P-A and "Supplement for United States,"
NEDE-24011-P-A-US (most recent approved revision referenced in COLR).
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.1-17 Table 15.1-1 Sequence of Events for Figure 15.1-1
Feedwater Controller Failure 100% Reactor Power / 106% Core Flow Time (sec) Event 0 Initiate simulated failure of 139% upper limit on feedwater flow.
10.06 L8 vessel level setpoint trips main turbine and feedwater pumps. Turbine bypass operation initiated.
10.16 Turbine control (governor) or stop (throttle) valves fully closed.
10.08 Reactor scram trip actuated from ma in turbine control (governor) valve fast closure. 10.16 Turbine bypass va lves start to open. 10.27 Recirculation pump motor circuit breakers open causing decrease in core flow to natural circulation.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.1-18 Table 15.1-1A Sequence of Events for Figure 15.1-3
Feedwater Controller Failure Single Loop Operation 75% Power / 57% Flow Time (sec) Event 0 Initiate an upper limit failure of 146% of rated feedwater flow. 8.39 L8 vessel level setpoint trips main turbine and feedwater pumps. 8.39 Recirculation pump tr ip (RPT) actuated by st op valve position switches. 8.40 Reactor scram trip actuated from main turbine stop valve position switches. 8.49 Turbine stop valves closed and main turbine bypa ss valves start to open. 8.58 Recirculation pump motor circuit breakers open causing decrease in core flow to natural circulation.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.1-19 Table 15.1-2 Sequence of Events for Figure 15.1-2
Pressure Regulator Fail ure - Open Uprated Power Time (sec) Event 0 Simulate maximum limit on steam flow, (130%) to main turbine. 0.2 Main turbine bypass valves open. 3.31 Vessel water level (L8) trip initiates turbin e and feedwater trips. 3.32 Main turbine stop valves reac h 90% open position initiating a reactor scram. 3.50 Both recirculation pumps trip. 6.15 Feedwater recirc ulation valves trip. 6.95 Group 3 relief valves actuated. 7.40 Group 4 relief valves actuated.
10 a Pressure relief valves closed.
57.98 a Main steam line isolation valves closed on turbine inlet pressure (approximately 850 psia). 77 High-pressure core spray and RCIC system initiation on low level (L2).
a Estimated.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.1-20 Table 15.1-3 Sequence of Events for Inadvertent Safety/Relief Valve Opening
Uprated Power Time Event 0 Initiate opening of one SRV which re mains open throughout the event.
1 a Reactor dome pressure decreases.
3 a DEH turbine control system pressure regulator initiates closure of the turbine control (governor) valves to stabilize reactor vessel pressure. 8+ Reactor power settles near the initial power level.
a Approximately.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.1-21 Table 15.1-4 Sequence of Events for Inadvertent Residual Heat Removal Shutdown Cooling Operation
Original Rated Power Timea Event 0 Residual heat removal shutdown cooling inadvertently activated. 0-10 minutes Slow rise in reactor power. +10 minutes Operator may take action to limit power rise. Flux scram will occur if no action is taken.
a Approximately.
-25.0 25.0 75.0 125.0 175.00.02.04.06.08.010.012.014.016.0 Time (sec)% Rated Vessel Press Rise (psi)
Safety Valve Flow Relief Valve Flow Bypass Valve Flow
-50.0-25.0 0.0 25.0 50.0 75.0 100.0 125.0 150.0 175.0 200.00.02.04.06.08.010.012.014.016.0 Time (sec)% Rated Level - Inch above Sep Skirt Vessel Steam Flow Turbine Steam Flow Feedwater Flow
-2.0-1.5-1.0-0.5 0.0 0.5 1.0 1.5 2.0 2.50.02.04.06.08.010.012.014.016.0 Time (sec)
R eactivity Com ponents ($) Void Reactivity Doppler Reactivity Scram Reactivity Total Reactivity 0.0 50.0 100.0 150.0 200.0 250.00.02.04.06.08.010.012.014.016.0 Time (sec)% Rated Neutron Flux Average Surface Heat Flux Core Inlet Flow Core Inlet Subcooling Figure Form No. 960690 LDCN-08-035 Draw. No.Rev.910402.34 15.1-1 Columbia Generating Station Final Safety Analysis Report Amendment 60 December 2009 Feedwater Controller Failure, Maximum Demand System Response 100% Power, 106% Flow Amendment 53 November 1998 Columbia Generating Station Final Safety Analysis Report Pressure Regulator Failure - Open at 106.2%
Uprated Power, 100% Flow 020361.57 15.1-2.1 Amendment 53 November 1998 Columbia Generating Station Final Safety Analysis Report Pressure Regulator Failure - Open at 106.2%
Uprated Power, 100% Flow 020361.58 15.1-2.2 Amendment 53 November 1993 Columbia Generating Station Final Safety Analysis Report Pressure Regulator Failure O pen at 106.2% Uprated Power, 100% Flow 020361.59 15.1-2.3 Amendment 53 November 1998 Columbia Generating Station Final Safety Analysis Report Pressure Regulator Failure - Open at 106.2%
Uprated Power, 100% Flow 020361.60 15.1-2.4 Amendment 53 November 1998 Columbia Generating Station Final Safety Analysis Report Pressure Regulator Failure Open at 106.2% Uprated Power, 100% Flow 020361.61 15.1-2.5 Amendment 53 November 1998 Columbia Generating Station Final Safety Analysis Report Pressure Regulator Failure Open at 106.2% Uprated Power, 100% Flow 020361.62 15.1-2.6 Amendment 53 November 1998 Columbia Generating Station Final Safety Analysis Report Pressure Regulator Failure - Open at 106.2%
Uprated Power, 100% Flow 020361.63 15.1-2.7 Amendment 53 November 1998 Columbia Generating Station Final Safety Analysis Report Pressure Regulator Failure - Open at 106.2%
Uprated Power, 100% Flow 020361.64 15.1-2.8 Amendment 53 November 1998 Columbia Generating Station Final Safety Analysis Report Pressure Regulator Failure - Open at 106.2%
Uprated Power, 100% Flow 020361.65 15.1-2.9 Amendment 53 November 1998 Columbia Generating Station Final Safety Analysis Report Pressure Regulator Failure - Open at 106.2%
Uprated Power, 100% Flow 020361.66 15.1-2.10 Amendment 53 November 1998 Columbia Generating Station Final Safety Analysis Report Pressure Regulator Failure Open at 106.2% Uprated Power, 100% Flow 020361.67 15.1-2.11 Amendment 57December 2003Feedwater Controller Failure, Maximum Demand,EOC RPT OOS, Single Loop Operation and 75%Uprated Power, 57% Flow 910402.41 15.1-3.1 Figure Form No. 960690Draw. No.Rev.150 100 50 0 0 5 10 15 20Time (Sec)(Percent of Rated)
Neutron FluxAve Surface Heat Flux
Core Inlet Flow Columbia Generating StationFinal Safety Analysis Report LDCN-03-003 Amendment 57December 2003 LDCN-03-003 910402.40Feedwater Controller Failure, Maximum Demand,EOC RPT OOS, Single Loop Operation and 75%Uprated Power, 57% Flow 15.1-3.2 Figure Form No. 960690Draw. No.Rev.200 100 0-100 0 5 10 15 20Time (Sec)Level (Inch Ref SIP Skirt)Vessel Steamflow Turbine Steamflow
Feedwater Flow Columbia Generating StationFinal Safety Analysis Report Amendment 57December 2003 LDCN-03-003 910402.39Feedwater Controller Failure, Maximum Demand,EOC RPT OOS, Single Loop Operation and 75%Uprated Power, 57% Flow 15.1-3.3 Figure Form No. 960690Draw. No.Rev.300 200 100 0 0 5 10 15 20Time (Sec)Vessel Pres Rise (PSI)Safety Valve Flow Relief Valve Flow Bypass Valve Flow Columbia Generating StationFinal Safety Analysis Report Amendment 57December 2003 LDCN-03-003 910402.38Feedwater Controller Failure, Maximum Demand,EOC RPT OOS, Single Loop Operation and 75%Uprated Power, 57% Flow 15.1-3.4 Figure Form No. 960690Draw. No.Rev.1 0-1-2 0 5 10 15 20Time (Sec)
Reactivity ComponentsVoid Reactivity Doppler Reactivity
Scram Reactivity Total Reactivity Columbia Generating StationFinal Safety Analysis Report C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-06-057,07-011 15.2-1 15.2 INCREASE IN REACTOR PRESSURE
15.2.1 PRESSURE REGULATOR FAILURE - CLOSED
This transient is classified as a nonlimiting event for both original and uprated power conditions. Therefore, cycle specific analyses are not performed for this event. The analysis results presented in the section are based on uprated power conditions and a representative reload core (Cycle 8) as documented in Reference 15.2-5. 15.2.1.1 Identification of Causes and Frequency Classification 15.2.1.1.1 Identification of Causes
A triple redundant control system is provided to maintain primar y system pressure control.
The pressure upstream of the main turbine stop (thro ttle) valves is sens ed by three redundant throttle pressure transmi tters and the control system uses a median select logic to determine which pressure transmitter is used to control throttle pressure. The pressure control system compares the detected throttle pres sure to a pressure setpoint to control the position of the main turbine control (governor) valves in order to control pressure.
It is assumed for purposes of this transient analysis that a single failure occurs on the controlling pressure transmitter wh ich erroneously causes the DEH control system to close the turbine control (governor) valves and thereby increases reactor pressure. If this occurs, the self diagnostics ability and triple redundant control system is available.
15.2.1.1.2 Frequenc y Classification
This event is categorized as an incident of mode rate frequency.
15.2.1.2 Sequence of Events and Systems Operation
15.2.1.2.1 Sequence of Events
A failure of a DEH control system component that causes the turbine control (governor) valves or turbine bypass valves to m ove towards the closed position will momentarily result in an initial pressure increase because the reactor is still generating th e initial steam flow. The DEH control system is self diagnostic. It will detect the faulty component and disable it. The control system is redundant and will continue to perform its f unctions, and will restore steady state operation.
For a failure that causes the DE H turbine control pressure regu lator to initiate a demand signal to close the turbine control (governor) valves (requires multiple co mponent failures), there will be an increase in system pressure and reactor power. A scram will be initiated when the high
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-06-057,07-011 15.2-2 neutron flux scram setpoint is reached. The pressure rises to the pressure relief setpoint, part of the relief valves open, disc harging steam to the suppression pool. The plant response is given in Table 15.2-1.
15.2.1.2.2 Systems Operation
Normal plant instrumentation and control is assumed to function except for the pressure regulator failure. The event is analyzed from 104.1% uprated power and 106% of rated core flow. The event results in a high flux trip initiated by the reactor protection system.
15.2.1.2.3 The Effect of Single Failures and Operator Errors
The first assumed failure produces a slight pr essure increase in th e reactor until the DEH control system adjusts to the singl e failure and gains cont rol. No other action is significant in restoring normal operation. If subsequent failures occur such that the DEH control system further closes the turbine control (governor) valves the reactor pr essure could rise to the point where a flux or pressure scram trip would be initiated to shutdown the reactor. This event is
less severe than the turbine trip for the following reasons:
- a. For the DEH control system failure-closure event the reactor scrams on high neutron flux or pressure but the recirc ulation pumps do not trip. As a result, core flow remains at 100% or greater throughout the cr itical portion of the transient with respect to the critical power ratio (CPR). This provides improved heat transfer capability in relation to the turbine trip transient; and
- b. Since the turbine control (governor) valves close in response to a pressure error signal, their closure rate is not as fast as the turbin e stop (throttle) or control (governor) valve response to a trip signal. This produces a slower pressurization rate for the DE H control system failure rela tive to the turbine trip event. This in turn resu lts in a lower peak neutr on flux and therefore a lower peak surface heat flux than the turbine trip event.
15.2.1.3 Core and Sy stem Performance 15.2.1.3.1 Mathematical Model The one-dimensional ODYN m odel described in Section 15.0.3.3.1 is used to simulate this event.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-07-011 15.2-3 15.2.1.3.2 Input Paramete rs and Initial Conditions
The analyses have been perf ormed with plant conditions at 104.1% of uprated power and 106% of rated core flow. The input parameters are gi ven in detail in Table 15.0-2 under the ODYN column.
15.2.1.3.3 Results
The closure of the turbine governor (control) valves results in a rise in reactor pressure, collapsing the coolant voids which in turn increases the neutron flux. One sec after the initiation of the event the neutron flux increases to the high flux setpoint signal and initiates a reactor scram. Two sec into the event the pressure in the reactor reaches the ATWS high pressure trip setpoint, initiating a recirculation pump trip signal. As the pressure in the reactor system continues to rise, the re lief valves begin to open starting with Group 3. The maximum pressure is reached at 3.25 sec and is calculated to be 1220 psig at the bottom of the reactor vessel. Table 15.2-1 provides the sequence of events and Figure 15.2-1 depicts the plant parameters responses.
Key transient peak valu es are presented in Table 15.0-1. This event is nonlimiting in that the pressurization event is less severe than the Generator Load Rejection with Bypass Failure and Turbine Trip with Bypass Failure events.
15.2.1.3.4 Consideration of Uncertainties
The uncertainties included in the initial power and flow considerations maximize the consequences of the plant re sponse. The independent pre ssure regulators normally respond such that failure of one would be compensated by the other regulator with plant not experiencing a trip.
15.2.1.4 Barrier Performance
The consequences of this even t do not result in any temperature or pressure transient (see Table 15.0-1) in excess of the criteria for which the fuel, pressure vessel, or containment are designed. Therefore, barrier inte grity and function is maintained.
15.2.1.5 Radiological Consequences Since this event does not result in any fuel failures or any release of primary coolant to either the secondary containment or to the environment, there are no radiological consequences associated with this event.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-07-011 15.2-4 15.2.2 GENERATOR LOAD REJECTION
15.2.2.1 Identification of Causes and Frequency Classification
15.2.2.1.1 Identification of Causes
Fast closure of the turbine c ontrol (governor) valves (TCVs) is initiated whenever electrical grid disturbances occur which re sult in significant loss of electri cal load on the generator. The TCVs are required to close as rapidly as possible to prevent excessive overspeed of the turbine-generator rotor. Clos ure of the main TCVs will ca use a sudden reduction in steam flow which results in an increase in system pressure, which may cause a reactor shutdown due to a high flux or high steam pressure condition.
15.2.2.1.2 Frequenc y Classification
15.2.2.1.2.1 Genera tor Load Rejection. This event is cat egorized as an incident of moderate frequency.
15.2.2.1.2.2 Generator Load Rejection with Bypass Failure. This event is categorized as a moderate frequency event.
15.2.2.2 Sequence of Events and System Operation
The generator load rejection with bypass failure event is the most limiting (with respect to thermal margin) of the class of transients characterized by rapid vessel pressurization, including load rejection with the bypass valves operating. The generator load rejection causes
a TCV (governor valve) fast clos ure, which initiates a reactor scram and a recirculation pump trip (RPT). The compression wave produced by the TCV fa st closure travels through the steam lines into the vessel and pressurizes the reactor vessel and core. Bypass flow to the condenser, which would mitigate the pressurization effect, is conservative ly not allowed. The excursion of core power due to void collapse is primarily terminated by reactor scram and void growth due to RPT. The recirc ulation pump speed remains c onstant until tripped by the RPT system.
Events caused by low water level trips, includi ng closure of main steam line isolation valves (MSIVs), and initiation of high-pressure core spray (HPCS) and r eactor core isolation cooling (RCIC) are not included in the simulation. Shou ld these events occur, they will follow after the primary concerns of fuel thermal margin and overpressure effects have occurred, and are expected to be less severe than those already experienced by the system.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-07-011 15.2-5 15.2.2.2.1 Sequence of Events
15.2.2.2.1.1 Generator Lo ad Rejection - Turbine C ontrol Valve Fast Closure. This transient is classified as a nonlimiting event for both original and uprated power conditions. Therefore, the original rated power (3323 MWt) analysis has not been updated.
A loss of generator electrical lo ad from high power conditions pr oduces the sequence of events listed in Table 15.2-2. 15.2.2.2.1.2 Generator Load Rejection with Failure of Bypass. A loss of generator electrical load at 3486 MWt with bypass failure produ ces the sequence of events listed in Table 15.2-3. 15.2.2.2.2 System Operation
15.2.2.2.2.1 Generator Lo ad Rejection with Bypass. To properly simulate the expected sequence of events, the anal ysis of this event assume s normal functioning of plant instrumentation and controls, pl ant protection, reactor pressure vessel (RPV) safety/relief valves (SRV), and reactor protec tion systems (RPS) unless stated otherwise. The bypass valve opening characteristics reflect the specified delay together with the specified opening characteristic required for bypass system operation.
Turbine control valve fast clos ure initiates a scram tr ip signal for power le vels greater than 30% nuclear boiler rated (NBR). In addition, recirculation pump trip (RPT) is initiated. Both of these trip signals satisfy single failure criterion and credit is taken for these protection features.
The pressure relief system, which operates th e SRVs independently when system pressure exceeds relief valve instrumentation setpoints is assumed to function normally during the time period analyzed.
15.2.2.2.2.2 Generator Load Rejection with Failure of Bypass. Same as Section 15.2.2.2.2.1 except that failure of the main turbine bypass valv es is assumed for the entire transient. In addition, the pressure relief system, which operates the SRVs independently when system pressure exceeds relief valve in strumentation setpoints, fails to operate. Pressure relief is provided by the safety function of the SRVs.
15.2.2.2.3 The Effect of Single Failures and Operator Errors
Mitigation of pressure increase, the basic nature of this transi ent, is accomplished by the RPS functions. Turbine control valve trip scram and RPT are designed to satisfy the single failure criterion. An evaluation of the most limiting sing le failure (i.e., failu re of the bypass system) was considered in this event.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 15.2-6 15.2.2.3 Core and Sy stem Performance 15.2.2.3.1 Mathematical Model
15.2.2.3.1.1 Generator Lo ad Rejection with Bypass. The predicted dynamic behavior for the generator load reject with by pass valves operable has been determined using a computer simulated, analytical model of a generic direct-cycle BWR. Th is model is described in detail in Reference 15.2-4. The nonlinear computer simulated analytical model is designed to predict associated transient behavior of the reactor. Some of the significant features of the model are the following:
- a. A point kinetic model is assumed with reactivity feedbacks from control rods (absorption), voids (moderation), and Doppler (cap ture) effects.
- b. The fuel is represented by three four-node cylindrical elements , each enclosed in a cladding node. One of the cylindrical elements is used to represent core average power and fuel temperature conditions, providing the source of Doppler
feedback. The other two are used to re present "hot spots" in the core, to simulate peak fuel center temperature and cladding temperature.
- c. Four primary system pressure nodes are simulated. Th e nodes represent the core exit pressure, vessel dome pressu re, steam line pressure (at a point representative of the safety/relief valve location), a nd turbine inlet pressure.
- d. The active core void fraction is calculat ed from a relationship between core exit quality, inlet subcooling, and pressure.
This relationship is generated from multimode core steady-state calculations. A second-order voi d dynamic model, with the void boiling sweep time calculated as a function of core flow and void conditions, is also utilized.
- e. Principal controller functi ons such as feedwater flow, recirculation flow, reactor water level, pressure an d load demand are represented together with their dominant nonlinear characteristics.
- f. The ability to simulate necessary reactor protect ion system functions is provided.
15.2.2.3.1.2 Generator Load Rejection with Bypass Failure. The predicted dynamic behavior for the load rejection with bypa ss inoperable has been determined using a computer simulated, analytical model of a direct-cycle BWR that is disc ussed in Section 15.1.2.3.1. This model is described in deta il in Reference 15.2-2.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-029 15.2-7 15.2.2.3.2 Input Paramete rs and Initial Conditions
These analyses have been pe rformed, unless otherwise noted , with the plant conditions tabulated in Table 15.0-2 for the load rejection with bypass and in Table 15.0-2B for load rejection with bypass failure.
The turbine digital elec trohydraulic control system power/l oad imbalance device detects load rejection before a measurable speed change takes place.
The closure characteristics of the TCVs are assumed such that the valves operate in the full arc (FA) mode and have a full stroke closure time, from fully open to fully closed, of 0.15 sec. In FA mode, at 100% power, the TCVs are not fu lly open, so the analys is assumes a closure time that is a fraction of the full stroke time proportional to the TCV initial position.
15.2.2.3.3 Results
Analyses were performed to analyze combina tions of RPT operable/
inoperable and Option A and Option B scram speeds (Reference 15.2-3). The excursion of core power due to void collapse is primarily terminated by reactor scram and void growth due to RPT.
15.2.2.3.3.1 Generator Load Rejection with Bypass.
Figure 15.2-2.1 shows the results of the generator trip from original rated power.
Peak neutron flux rises 156.8% above NBR conditions.
The average surface heat flux p eaks at 102.9% of the initial va lue and minimum critical power ratio (MCPR) does not significantly decrease below its initial value.
15.2.2.3.3.2 Generator Load Rejection with Failure of Bypass. Figure 15.2-2.2 shows that, for the case of bypass failure, p eak neutron flux reaches about 236% power, average surface heat flux reaches 111% of its in itial value. Results reflect GE14 fuel introduction, some of which are dependent on fuel design and core loading pattern. Compliance w ith the event acceptance criteria is demonstrat ed by cycle-dependent analysis of potentially limiting events just prior to the operation of th at cycle. The results are repo rted in the Supplemental Reload Licensing Report (Reference 15.2-3). As discussed in Section 15.0.2.1 , when this event is initiated during single loop operation, the consequences are less severe than the consequences analyzed for the two loop operation.
15.2.2.3.4 Consideration of Uncertainties
The full stroke closure time of the TCV of 0.15 sec is conservative. Typically, the actual closure time is closer to 0.2 s ec. The less time it takes to close, the more severe the pressurization effect.
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 15.2-8 All systems used for pr otection in this event were assu med to have the poorest allowable response. Expected plant behavior is, therefore, expected to re duce the actual severity of the transient.
15.2.2.4 Barrier Performance
15.2.2.4.1 Generator Load Rejection
Peak pressure remains within normal operating range and no threat to the barrier exists.
15.2.2.4.2 Generator Load Reject ion with Failure of Bypass
The peak steam line pre ssure reaches 1235 psig. The peak reactor coolant pressure boundary (RCPB) pressure reaches 1260 psig. The peak pressure remains well below the nuclear barrier transient pressure limit of 1375 psig.
15.2.2.5 Radiological Consequences
While the consequences of this event do not result in fuel failures, the result includes the discharge of normal coolant activity to the suppression pool by means of safety/relief valve (SRV) operation. Since this activity is contained in the primary containment, there will be no
exposure to the public. Since th is event does not result in an uncontrolled release to the environment, the plant operator can choose to hold the activity in containment or filter the discharge prior to release to the environm ent when conditions pe rmit in accordance with established re quirements.
15.2.3 TURBINE TRIP
15.2.3.1 Identification of Causes and Frequency Classification
15.2.3.1.1 Identification of Causes
A variety of turbine or nuclear system malfunctions will initiate a turbine trip. Some examples are moisture separator high levels, operator lockout, loss of contro l fluid pressure, low condenser vacuum, and reactor high water level.
15.2.3.1.2 Frequenc y Classification
15.2.3.1.2.1 Turbine Trip. This event is categor ized as an incident of moderate frequency.
In defining the frequency of this event, turb ine trips which occur as a by-product of other transients such as loss of condenser vacuum or reactor high level trip events are not included.
However, spurious low vacuum or high level trip signals which cause an unnecessary turbine trip are included in defining the frequency.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-029 15.2-9 15.2.3.1.2.2 Turbine Trip with Failure of Bypass. This transien t disturbance is categorized as a moderate fre quency incident.
15.2.3.2 Sequence of Events and Systems Operation
15.2.3.2.1 Sequence of Events
15.2.3.2.1.1 Turbine Trip. This transient is classified as a nonlimiting event for both original and uprated power conditions. Therefore, the or iginal rated power (3323 MWt) analysis has not been updated. Turbine trip at high power produces the sequence of events listed in
Table 15.2-4.
15.2.3.2.1.2 Turbine Trip with Failure of Bypass. Turbine trip at high power with bypass failure produces the sequence of events listed in Table 15.2-5.
15.2.3.2.2 Systems Operation
15.2.3.2.2.1 Turbine Trip. All plant contro l systems maintain normal operation unless specifically designate d to the contrary.
Turbine stop (throttle) valve closure initiates a reactor scram trip by means of valve position signals to the protection system.
Turbine stop valve closure initiates RPT thereby terminati ng the jet pump drive flow.
The pressure relief system, which operates th e relief valves indepe ndently when system pressure exceeds relief valve instrumentation se tpoints, is assumed to function normally during the time period analyzed.
The severity of turbine trips from lower initial power levels decreases to th e point where a scram can be avoided if auxiliary power is av ailable from an external source and the power level is within the bypass capability.
15.2.3.2.2.2 Turbine Trip with Failu re of Bypass. Same as Section 15.2.3.2.2.1 except that failure of the main turbine bypass system is assumed for th e entire transient time period analyzed. During the transient the SRVs open and close se quentially as the stored energy is dissipated until the pressure fall s below the valve setpoints.
15.2.3.2.2.3 Turbine Trip at Low Power with Failure of Bypass. Same as Section 15.2.3.2.2.1 except that failure of the main turbine bypass system is assumed.
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 15.2-10 Below 30% NBR power level, a main stop valve sc ram trip inhibit signal derived from the first stage pressure of the tu rbine is activated. This is done to eliminate the stop valve scram trip signal from scramming the reactor provided the bypass system functions properly. In other words, the bypass would be sufficient at this lo w power to accommodate a turbine trip without the necessity of shutting down the reactor. All other protection system functions remain functional as before and credit is taken for those protection system trips.
15.2.3.2.3 The Effect of Single Failures and Operator Errors
15.2.3.2.3.1 Turbine Trips at Power Levels Greater Than 30% Nuclear Boiler Rated. Mitigation of pressure increase, the basic nature of this transi ent, is accomplished by the RPS functions. Main stop valve clos ure scram trip and RPT are desi gned to satisfy single failure criterion.
15.2.3.2.3.2 Turbine Trips at Power Levels Less Than 30% Nuclear Boiler Rated. Same as Section 15.2.3.2.3.1 except RPT and stop valve closure scram trip is normally inoperative. Since protection is still provi ded by high flux, high pressure, etc., th ese will continue to function and scram the reactor should a single failure occur.
15.2.3.3 Core and Sy stem Performance
15.2.3.3.1 Mathematical Model
15.2.3.3.1.1 Turbine Trip with Bypass. The predicted dynamic beha vior for the turbine trip has been determined using a computer simulated, analytical model of a generic direct-cycle BWR, as discussed in Section 15.2.2.3.1. This model is described in detail in Reference 15.2-4.
15.2.3.3.1.2 Turbine Tr ip with Bypass Failure. The one-dimensional ODYN model described in Section 15.0.3.3.1 is used to simulate this event.
15.2.3.3.2 Input Paramete rs and Initial Conditions
These analyses have been perfor med, unless otherwise noted, with plant conditions tabulated in Table 15.0-2 for the turbine trip with bypass and Table 15.0-2B for the turbine trip with bypass failure.
The turbine trip analysis was pe rformed at the 105% of the or iginal rated steam flow. The turbine trip with bypass failure was analyzed at an initial condition of 100% rated power (3486 MWt) and 106% rated core flow.
Turbine stop (throttle) valves full stroke closure time is 0.1 sec.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-029 15.2-11 A reactor scram is initiated by position switches on the stop valves when the valves are 90% open or less. This stop valve scram trip signal is automatically bypassed when the reactor is below 30% NBR power level.
Reduction in core recirculation flow is initiat ed by position switches on the main stop valves, which actuate trip circuitry which trips the recirculation pumps.
15.2.3.3.3 Results
15.2.3.3.3.1 Turbine Trip. The results of a turbine trip w ith the bypass system operating normally are shown in Figure 15.2-3. Neutron flux increases rapidly because of the void reduction caused by the pressure increase.
However, the flux increase is limited to 147.5%
of rated by the stop va lve scram and the RPT system. Peak fuel surface heat flux does not exceed 101.7% of its initial value.
15.2.3.3.3.2 Turbine Trip with Failure of Bypass. The results of a turbine trip w ith failure of the bypass system are shown in Figure 15.2-4.
The peak neutron flux reaches 278%
of its rated value, and p eak surface heat flux reaches
111% of its initial value.
Results reflect GE14 fuel introduc tion, some of which are depende nt on fuel design and core loading pattern. Compliance with the event acceptance criteria is demonstrated by cycle-dependent analysis of po tentially limiting events just prior to the operation of that cycle. The results are reported in the Supplementa l Reload Licensing Report (Reference 15.2-3).
15.2.3.3.3.3 Turbine Trip with Bypass Valve Failure, Low Powe
- r. This transient is less severe than a similar one at high power. Below 30% of rated power, the turbine stop valve closure and TCV (governor valve) closure scrams are automatica lly bypassed. At these lower power levels, turbine firs t stage pressure is used to initiate the scram logic bypass. The scram which terminates the transien t is initiated by high vessel pr essure. The bypa ss valves are assumed to fail; therefor e, system pressure will increase until th e pressure relief setpoints are reached. At this time, because of the relatively low power of this transient event, relatively few relief valves will open to limit reactor pressure. Peak pressures are not expected to greatly exceed the pressure relief valve setpoints and will be significantly be low the reactor coolant pressure boundary (RCPB) transient limit of 1375 psig. Peak surf ace heat flux and peak fuel center temperature remain at relatively low values and MCPR remain s well above the GETAB safety limit.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-12 15.2.3.3.4 Considerati ons of Uncertainties Uncertainties in these an alyses involve protection system settings, system capacities, and system response characteristics.
In all cases, the most conser vative values are used in the analyses. For example:
- a. Slowest allowable control rod scram motion is assumed, b. Scram worth shape for all-r od-out conditions is assumed, c. Minimum specified valve capacities are utilized for overpressure protection, and d. Setpoints of the SRVs include errors and uncertainties (high) for all valves.
15.2.3.4 Barrier Performance
15.2.3.4.1 Turbine Trip
Peak pressure in the bottom of the vessel reaches 1163 psig, which is below the American Society of Mechanical Engin eers (ASME) Code limit of 1375 ps ig for the RCPB. Vessel dome pressure does not exceed 1136 psig.
15.2.3.4.2 Turbine Trip with Failure of Bypass
The peak steam line pre ssure reaches 1235 psig. The peak reactor coolant pressure boundary (RCPB) pressure reaches 1260 psig. The peak pressure remains well below the nuclear barrier transient pressure limit of 1375 psig.
15.2.3.4.2.1 Turbine Trip with Failure of Bypa ss at Low Power. Qualitative discussion is provided in Section 15.2.3.3.3.3.
15.2.3.5 Radiological Consequences
The consequence of this event does not result in fuel failure. It does result in the discharge of normal coolant activity to the suppression pool by means of SRV operation, which is contained in the primary containment.
This event does not result in an uncontrolled release to the environment, so the plant operator can choose to hold the activity in containment or discharge it when conditions permit. If purging of the c ontainment is chosen, th e release would be in accordance with established requirements.
15.2.4 MAIN STEAM LINE ISOLATION VALVE CLOSURES
This transient is classified as a nonlimiting event for both original and uprated power conditions. Therefore, cycle specific analyses are not performed for this event. The analysis results presented in the section are based on uprated power conditions and a representative reload core (Cycle 8) as documented in Reference 15.2-5.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-13 15.2.4.1 Identification of Causes and Frequency Classification
15.2.4.1.1 Identification of Causes
Various steam line and nuclear system malfunctions, or operator actions, can initiate MSIV closure. Examples are low-steam line pressu re, high-steam line flow, low-water level, or manual action.
15.2.4.1.2 Frequenc y Classification
15.2.4.1.2.1 Closure of All Main Steam Line Isolation Valves. This event is categorized as an incident of moderate frequency. To define the frequency of this event as an initiating event and not the byproduct of another transient, only the following contribute to the frequency: Manual action (purposely or inadvertent); spurious signals such as low pressure, low reactor water level, and low condenser vacuum; and equipment malfunctions such as faulty valves or operating mechanisms. A closur e of one MSIV may cause an immediate closure of all the other MSIVs depending on reactor conditions. If this occurs, it is also included in this category. During the MSIV closure, position switches on the valves provide a reactor scram when the valves in three or more main steam lines are less than 90% open (except for interlocks which permit pr oper plant startup). Protection system logic permits the test closure of one valve without initiating scram from the position switches.
15.2.4.1.2.2 Closure of One Main Steam Line Isolation Valve. This event is categorized as an incident of moderate frequency. One MSIV at a time may be manually closed for testing purposes. Operator error or equipment malf unction may cause a single MSIV to be closed inadvertently. If reactor power is greater than about 75% when this occurs, a high flux or high steam line flow condition may result in a scram. If all MSIVs cl ose as a result of the single event, the event is considered as a closure of all MSIVs. The results presented for this event assume all MSIVs close as a result of an unspecified initiating event.
15.2.4.2 Sequence of Events and Systems Operation
15.2.4.2.1 Sequence of Events
Table 15.2-6 lists the sequence of events for Figure 15.2-5. When the MSIV's reach their 85% open position, a reactor scram is initiated by the reactor protection system. The valve closure results in a system pressure increase which in turn results in a spike in reactor neutron flux. The reactor vessel pressure increase also results in an ATWS recirculation pump trip (RPT). As the pressure increases, the relief valves begin to open terminating the pressure increase.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-14 15.2.4.2.2 Systems Operation
15.2.4.2.2.1 Closure of All Main Steam Line Isolation Valves. The MSIV closures initiate a reactor scram trip by means of position signals to the protection system. Credit is taken for successful operation of the protection system.
The pressure relief system wh ich initiates opening of the relief valves when system pressure exceeds relief valve instrumentation setpoints is assumed to function normally during the time period analyzed.
All plant control systems maintain normal ope ration unless specifically designated to the contrary.
15.2.4.2.2.2 Closure of One Main Steam Line Isolation Valve.
A closure of a single MSIV will not initiate a reactor scram by means of the position signal to the protection system. This is because the valve position scram trip logic is designed to accommodate single valve operation and testability during no rmal reactor operation at limited power levels. Credit is taken for the operation of the pressure and flux signals to initiate a reactor scram.
All plant control systems maintain normal ope ration unless specifically designated to the contrary.
15.2.4.2.3 The Effect of Single Failures and Operator Errors
Mitigation of pressure increase is accomplished by initiation of the reactor scram by means of MSIV position switches and the protection system. Relief valves also operate to limit system pressure. All of these aspects are designed to single failure criteri on and additional single failures would not alter the results of this analysis.
Failure of a single relief valve to open is not expected to have any significant effect. Such a failure is expected to result in less than a 5 psi increase in th e maximum vessel pressure rise. The peak pressure will still remain considerably below 1375 psig.
15.2.4.3 Core and Sy stem Performance 15.2.4.3.1 Mathematical Model
The point-kinetics REDY model described in Section 15.0.3.3.1 is used to simulate these transient events.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-15 15.2.4.3.2 Input Paramete rs and Initial Conditions It is assumed the closure of all MSIVs occurs with the plan t operating at 106%
of uprated power and 100% core flow. The input parameters are defined with the plant conditions tabulated in Table 15.0-2 for power uprate.
The MSIVs close in 3 to 5 sec. The worst case, the 3-sec clos ure time, is assumed in this analysis.
Position switches on the valves in itiate a reactor scram when the valves are less than 90% open as described in Section 7.2 (85% is assumed in the analysis).
Closure of these valves inhibits steam flow to the feedwater turb ines terminating feedwater flow.
Valve closure indirectly causes a trip of the main turbine and generator.
Because of the loss of feedwate r flow, water level within the vessel decreases sufficiently to initiate trip of the recircul ation pump and to initiate th e HPCS and RCIC systems.
15.2.4.3.3 Results
15.2.4.3.3.1 Closure of All Main Steam Line Isolati on Valves. The reacto r scram is initiated at 0.45 sec when the MSIVs reach 85% open pos ition. The nuclear syst em relief valves begin to open at 3.08 sec after the start of isolation. The valves close sequentially as the stored heat is dissipated but continue to discharge the decay heat intermittently.
Table 15.2-6 provides the sequence of events and Figure 15.2-5 depicts the plant parameters responses. Key transient peak values are presented in Table 15.0-1. This event is non-limiting in that the pressurization event and change in CPR margin is less severe than the Genera tor Load Rejection with Bypass Failure and Turbine Trip with Bypass Failure events.
15.2.4.3.3.2 Closure of One Main Steam Line Isolation Valve.
Only one isolation valve is permitted to be closed at a time for testing purposes to prevent sc ram. Normal test procedure requires an initial power reduction to approximat ely 65% to 70% of design conditions to avoid high-flux scram, high-pressure scram, or full isolation from a high-steam flow condition in the
open steam lines. With a 3-sec closure of one MSIV during 10 5% of original rated power conditions, the steam flow disturbance raises vessel pressure and r eactor power enough to initiate a high neutron flux scram.
This transient is considerab ly milder than closure of all MSIVs at full power. No quantitative analysis is furnished for this event. No significant change in thermal margins is e xperienced and no fuel damage occurs. Peak pressure remains below SRV setpoints.
Inadvertent closure of one or all of the isolat ion valves while the reactor is shut down will produce no significant transient.
Closures during plant heatup will be less severe than the maximum power cases (maximum stored and decay heat).
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-16 15.2.4.3.4 Considerati ons of Uncertainties
Uncertainties in these an alyses involve protection system settings, system capacities, and system response characteristics.
In all cases, the most conser vative values are used in the analyses. For example:
- a. Slowest allowable control rod scram motion is assumed, b. Scram worth shape for all-r od-out conditions is assumed, c. Minimum specified valve capacities are used for overpressure protection, and
- d. Setpoints of the SRVs are assumed to be 15 psi higher than the valve's nominal setpoint.
15.2.4.4 Barrier Performance
15.2.4.4.1 Closure of All Main Steam Line Isolation Valves
The nuclear system relief valves begin to open at approximately 3.1 sec after the start of isolation. The valves close se quentially as the stored heat is dissipated but continue to discharge the decay heat interm ittently. Peak pressure at th e vessel bottom reaches 1234 psig, clearly below the pressure limits of the RCPB. Peak pressure in the main steam line is 1198 psig.
15.2.4.4.2 Closure of One Main Steam Line Isolation Valve
No significant effect is imposed on the RCPB, since if closure of the valve occurs at a high operating power level a flux or pressure scram will result. The main turbine bypass system will continue to regulate system pressure by means of the other three steam lines.
15.2.4.5 Radiological Consequences
While the consequence of this event does not result in fuel failure, it does result in the discharge of normal coolant activity to the suppression pool via SRV operation. Since this activity is contained in the prim ary containment, there will be no exposure to the public. Since this event does not result in an uncontrolled release to the environment, the plant operator can choose to hold the activity in containment or discharge it to the enviro nment when conditions permit. If purging of the containment is chosen, the release would be in accordance with established re quirements.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-17 15.2.5 LOSS-OF-CONDENSER VACUUM
This transient is classified as a nonlimiting event for both original and uprated power conditions. Therefore, cycle specific analyses are not performed for this event. The analysis results presented in the section are based on uprated power conditions and a representative reload core (Cycle 8) as documented in Reference 15.2-5.
15.2.5.1 Identification of Causes and Frequency Classification 15.2.5.1.1 Identification of Causes
Various malfunctions can cause a loss-of-condenser vacuum.
The causes and estimated vacuum decay rates include failu re or isolation of steam jet air ejectors (<1 in. Hg/mm), loss of sealing steam shaft gland seals (1 to 2 in. Hg/minute), ope ning of vacuum breaker valves (2 to 12 in. Hg/minute), and loss of one or more circulating water pumps (4 to 24 in. Hg/minute).
15.2.5.1.2 Frequenc y Classification
This event is categorized as an incident of mode rate frequency.
15.2.5.2 Sequence of Events and Systems Operation
15.2.5.2.1 Sequence of Events
Table 15.2-7 lists the sequence of events for Figure 15.2-6.
15.2.5.2.2 Systems Operation
It is conservatively assumed that condenser vacuum is lost at a rate of 2 inches of Hg per second. The bypass system is si gnaled to close approximately 10 inches of Hg less than the stop (throttle) valve closure vacuum setpoint level which means the b ypass is available for approximately 5 sec before the turbine stop (t hrottle) valves close. The loss of vacuum initiates a main turbine tr ip and feedwater turbine trip. Up on reaching 90% close, the turbine throttle (stop) valves closure results in a reactor scram. As th e reactor pressure increases, the relief valves will open. Subsequently, this result s in the main steam line isolation valves to close. However, the effect of the MSIV clos ure is minimal since th e turbine stop (throttle) valve and bypass valve closure have alrea dy terminated main steam line flow.
Tripping functions incurred by sensing main turbine condenser vacuum are designated in Table 15.2-8.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-18 15.2.5.2.3 The Effect of Single Failures and Operator Errors
Single failure will not effect the vacuum mo nitoring and turbine trip devices which are redundant. The protective sequences of the anticipated operationa l transient are shown to be single failure proof.
15.2.5.3 Core and Sy stem Performance
15.2.5.3.1 Mathematical Model The one-dimensional ODYN m odel described in Section 15.0.3.3.1 was used to simulate this transient event.
15.2.5.3.2 Input Paramete rs and Initial Conditions
This analysis was performed with plant conditions tabulated in Table 15.0-2 and at 104.1% of uprated power and 100% core flow. Turbine st op (throttle) valves full stroke closure time used in this analysis is 0.1 second and a react or scram is initiated by position switches on the stop valves when the valves are less than 90% open. The 2 in ches of Hg per second assumed in the analysis is conservativ e with respect to normal loss of vacuum and no operator actions are assumed.
Thus, the bypass system is available for several seconds since the bypass is signaled to close at a vacuum level of about 10 in. Hg less than the stop valve closure.
15.2.5.3.3 Results
The loss of condenser vacuum in itiates a main turbin e trip, which then initiates turbine bypass operation. The bypass is availabl e for approximately 5 sec until both the turbine bypass valves and the main steam line isolat ion valves receive a signal to close on low condenser vacuum.
The effect of MSIV closure tends to be minimal since the closure of main turbine stop valves and subsequently the bypass valv es have already shut off the main steam line flow.
Figure 15.2-6 shows the transient expected for this event. It is assumed that the plant is initially operating at 105% of uprated NBR steam flow conditions. Peak neutron flux reaches 256% of NBR power while average fuel surface heat flux reaches 111% of rated value. The SRVs open to limit the pressure rise then se quentially reclose as the stored energy is dissipated.
15.2.5.3.4 Consideration of Uncertainties
The reduction or loss of vacuum in the main turbine condenser will seque ntially trip the main and feedwater turbines and close the MSIVs a nd turbine bypass valves. While these are the major events occurring, other resultant actions will include scram (fro m stop valve closure)
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-19 and bypass opening with the main turbine trip.
Because the protective ac tions are actuated at various levels of condenser vac uum, the severity of the resul ting transient is dependent upon the rate at which the vacuum is lost. Normal loss of vac uum due to loss-of-cooling water pumps or steam jet air ejector problem produces a very slow rate of loss of vacuum (minutes, not seconds). If corrective actions by the reactor operators are not successful, then simultaneous trips of the main and feedwater turbines, and ultimately complete isolation by closing the bypass valves (opened with the main turbine trip) and the MSIVs, will occur.
A faster rate of loss of the condenser vacuum would reduce the anticipatory action of the scram and the overall effectiveness of the bypass valves since they would be closed more quickly.
Other uncertainties in these analyses involve protection syst em settings, syst em capacities, and system response characteristics.
In all cases, the most conser vative values are used in the analyses. For example:
- a. Slowest allowable control rod scram motion is assumed,
- b. Scram worth shape for all-r od-out conditions is assumed,
- c. Minimum specified valve capacities are utilized for overpressure protection, and
- d. Setpoints of the SRVs are assumed to be 15 psi higher than the valve's nominal setpoint.
15.2.5.4 Barrier Performance
The maximum calculated pressure for this event as presented in Table 15.0-1 is below the ASME Code limit of 1375 psig for the RCPB and the ASME Service Level C of 1500 psig. The consequences of this event do not result in any temperature or pressure transient in excess of the criteria for which the fu el, pressure vessel, or containm ent are designed. Therefore, barrier integrity and function is maintained.
15.2.5.5 Radiological Consequences
While the consequence of this event does not re sult in fuel failures, it does result in the discharge of normal coolant activity to the suppression pool by m eans of SRV operation. Since this activity is contained in th e primary containment, there will be no exposure to the public.
Since this event does not result in an uncontrolled release to the e nvironment, the plant operator can choose to hold the activity in containment or discharge it to the environment when conditions permit. If pur ging of the containment is chosen, the release would be in accordance with established requirements.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-20 15.2.6 LOSS OF ALTERNATING CURRENT POWER
This transient considers the loss of AC power to the plant fr om both an onsite cause (loss of auxiliary power transformer) and an offsit e cause (loss of all grid connections).
This transient is classified as a nonlimiting event for both original and uprated power conditions. Therefore, cycle specific analyses are not performed for this event. The analysis results presented in the section are based on uprated power conditions and a representative reload core (Cycle 8) as documented in Reference 15.2-5.
15.2.6.1 Identification of Causes and Frequency Classification
15.2.6.1.1 Identification of Causes
15.2.6.1.1.1 Loss of Auxiliary Power Transforme rs. Causes for interr uption or loss of the auxiliary power transformers can arise from normal operation or malfunctioning of transformer protection circuitry. These can include high transformer oil temperature, reverse of high current operation, and opera tor error which trips the transformer breakers.
15.2.6.1.1.2 Loss of All Grid Connections. Loss of all grid connections can result from major shifts in electrical loads, loss of loads, lightning, st orms, wind, etc., which contribute to electrical grid instab ilities. These in stabilities will cause equipment damage if unchecked. Protective relay schemes automatically disconnect electrical s ources and loads to mitigate damage and regain elec trical grid stability.
15.2.6.1.2 Frequenc y Classification
15.2.6.1.2.1 Loss of Auxiliary Power Transformers. This event is categorized as an incident of moderate frequency.
15.2.6.1.2.2 Loss of All Grid Connections. This event is ca tegorized as an incident of moderate frequency.
15.2.6.2 Sequence of Events and Systems Operation 15.2.6.2.1 Sequence of Events
15.2.6.2.1.1 Loss of Auxiliary Power Transformers.
Table 15.2-9 lists the sequence of events for Figure 15.2-7.
15.2.6.2.1.2 Loss of All Grid Connections.
Table 15.2-10 lists the sequence of events for Figure 15.2-8.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-21 15.2.6.2.2 Systems Operation
15.2.6.2.2.1 Loss of Auxiliary Power Transfor mers. This event, unless otherwise stated, assumes and takes credit for nor mal functioning of plant instrumentation and controls, plant protection, and reactor protection systems.
The reactor is subjected to a complex sequence of events when the plant loses all auxiliary power. Estimates of the responses of the va rious reactor systems (assuming loss of the auxiliary transformers) provide the following simulation sequence:
- a. Recirculation pumps a nd condenser circulatory wa ter pumps trip off at time = 0. A 4 sec recircula tion pump trip inertia time constant is assumed for
this analysis;
- b. Reactor scram and MSIV closure is initiated at 2 sec due to loss of power to the scram and MSIV re lay solenoids; and
Operation of the HPCS and RCIC are not simulated in this analys is. Their operation occurs at a time beyond the primary concerns of fuel thermal margin and overpressu re effects of this analysis.
15.2.6.2.2.2 Loss of All Grid Connections. Sa me as Section 15.2.6.2.2.1 with the following additional concern.
The loss of all grid connections would add a generator load reje ction to the above sequence at time, t=0. The load rejection immediately causes the TCVs (gove rnor valves) to close, causes a scram, and initiates RPT [already tripped at reference time t = 0].
15.2.6.2.3 The Effect of Single Failures and Operator Errors
Loss of the auxiliary power transformers in general leads to a reduction in power level due to rapid pump coastdown with pressu rization effects due to MSIV closure resulting from loss of power to the solenoids. Additiona l failures of the other systems assumed to protect the reactor would not result in an effect different from those reported. Fa ilures of the protection systems have been considered and satisfy single failure criteria and, as such, no change in analyzed consequences is expected.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-22 15.2.6.3 Core and Sy stem Performance 15.2.6.3.1 Mathematical Model
The point-kinetics REDY model described in Section 15.0.3.3.1 was used to simulate this event.
15.2.6.3.2 Input Paramete rs and Initial Conditions 15.2.6.3.2.1 Loss of Auxiliary Power Transformers. It is assumed the loss of the auxiliary power transformer occurs with the plant opera ting at 106% of uprated power and 100% core flow. The input parameters are defined with the plant conditions tabulated in Table 15.0-2 except as noted below.
- a. The recirculation pump trip in ertia time constant is 4 sec.
- b. The relay-type Reactor Trip System (RTS) circuitry generates a reactor scram and Main Steam Isolation Va lves (MSIV) closure signal due to loss of power to the scram and MSIV solenoids. This occurs 2 sec after the loss of offsite power. c. The feedwater pumps trip due to MSIV closure 2 sec after the MSIV begin to close as a result of the loss of power to the MSIV solenoids.
15.2.6.3.2.2 Loss of All Grid Connections. It is assumed th e loss of all grid connections occurs with the plant operating at 104% of uprated power and 100% core flow. The input parameters are defined with th e plant conditions tabulated in Table 15.0-2 except as noted below.
- a. The recirculation pump trip in ertia time constant is 4 sec.
- b. The relay-type Reactor Trip System (RTS) circuitr y generates a Main Steam Isolation Valves (MSIV) closure signal due to loss of power to the MSIV solenoids. This occurs 2 sec after the loss of offsite power.
- c. The feedwater pumps trip due to MSIV closure 2 sec after the MSIV begin to close as a result of the loss of power to the MSIV solenoids.
15.2.6.3.3 Results
15.2.6.3.3.1 Loss of Auxiliary Power Transfor mers. Initially the offsite power is cutoff causing both recirculation pumps to trip. The lo ss of power to the sc ram and MSIV solenoids causes a reactor scram, MSIV isolation and a feedwater pump trip 2 sec after the MSIV C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-23 isolation. Subsequently, the feedwater recirculation valves trip and the relief valves begin to open due to the rising pressure caused by the main steam line isolation.
Table 15.2-9 provides the sequence of events and Figure 15.2-7 depicts the plant parameters responses. Key transient peak values are presented in Table 15.0-1. This event is non-limiting in that the pressurization event and change in MCPR margin are less severe than the Generator Load Rejection with Bypass Failure and Turbine Trip with Bypass Failure events.
15.2.6.3.3.2 Loss of All Grid Connections. Loss of all grid connections is a more general form of loss of auxiliary power. It essentially takes on the characteristic response of the standard full load reject ion discussed in Section 15.2.2. Initially the offsite power is cutoff to the grid causing the turbine-generator to detect a loss of electrical load, and a power-load unbalance. The turbine genera tor overspeed protection control (OPC) initiates a control (governor) valve fast closure, turbine bypass valves opening and a reactor scram. At the same time both recirculation pump motors trip. Subsequently, MSIV isol ation occurs and both feedwater pumps trip. The rising pressure due to the isolation of the steam line causes the relief valves to open.
Table 15.2-10 provides the sequence of events and Figure 15.2-8 depicts the plant parameter responses. Key transient peak values are presented in Table 15.0-1. This event is non-limiting in that th e pressurization event and cha nge in MCPR margin are less severe than the Generator Load Rejection with Bypass Failure and Turb ine Trip with Bypass Failure events.
15.2.6.3.4 Consideration of Uncertainties
The most conservative characteristics of protection features are assumed. Any actual deviations in plant performance are expected to make the results of this event less severe.
Operation of the HPCS and RCIC systems are not included in the simulation of the first 50 sec of this transient. Startup of the pumps occurs in the latter part of this time period but the system has no significant effect on the results of this transient.
Following main steam line isolation and prior to RHR initiati on the reactor pressure is expected to increase until the SRV setpoints are reached. During this time the valves operate in a cyclic manner to discharge d ecay heat to the suppression pool.
15.2.6.4 Barrier Performance
15.2.6.4.1 Loss of Auxiliary Power Transformers
Safety/relief valves open in the pressure relief mode of operation as the pressure increases beyond their setpoints. The pre ssure in the dome is limited to a maximum value of 1169 psig well below the vessel pressu re limit of 1375 psig.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-09-007 15.2-24 15.2.6.4.2 Loss of All Grid Connections
Safety/relief valves open in the pressure relief mode of operation as the pressure increases beyond their setpoints. The pre ssure in the dome is limited to a maximum value of 1173 psig well below the vessel pressu re limit of 1375 psig.
15.2.6.5 Radiological Consequences
The consequence of this event does not result in fuel failure. It does result in the discharge of normal coolant activity to the suppression pool by means of SRV operation, which is contained in the primary containment.
This event does not result in an uncontrolled release to the environment, so the plant operator can choose to hold the activity in containment or discharge it when conditions permit. If purging of the c ontainment is chosen, th e release would be in accordance with established requirements.
15.2.7 LOSS-OF-FEEDWATER FLOW
This transient is classified as a nonlimiting event for both original and uprated power conditions. Therefore, cycle specific analyses are not performed for this event. The analysis results presented in the section are based on uprated power conditions and a representative reload core (Cycle 8) as documented in Reference 15.2-5. The analysis has not been updated for the change in MSIV isolation setpoint from Level 2 to Level 1 because the analysis is bounding and conclusions of the analysis are not affected (Reference 15.2-8).
15.2.7.1 Identification of Causes and Frequency Classification
15.2.7.1.1 Identification of Causes
A loss of feedwater flow could occur from pump failures, feedwater controller failures, operator errors, or reactor system variables such as a high vessel water level (L8) trip signal.
15.2.7.1.2 Frequenc y Classification
This event is categorized as an incident of mode rate frequency.
15.2.7.2 Sequence of Events and Systems Operation 15.2.7.2.1 Sequence of Events
Feedwater flow terminates at approximately 5 sec. Subcooling decreases causing a reduction in core power level and pressure. As power level is lowered, the turbine steam flow starts to drop off. Water level continue s to drop until the vessel level (L3) scram trip setpoint is reached, whereupon the reactor is shut down.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-25 Main steam line isolation initiation occurs due to vessel water droppi ng to the L2 trip. Also at this time, the recirculation system is trippe d and HPCS and RCIC operation is initiated. Operation of the HPCS and RCIC systems is not included in the simulation of the first
50 seconds of this transient sinc e startup of the pumps occurs in the latter part of this time period. Therefore, the system has no significant effects on the results of this transient.
Table 15.2-11 lists the sequence of events for Figure 15.2-9. 15.2.7.2.2 Systems Operation
Loss of feedwater flow results in a proportional reduction of vessel inventory causing the vessel water level to drop. The first correc tive action is the low le vel (L3) scram trip actuation. Reactor protection sy stem responds after this trip to scram the reactor. The low level (L3) scram trip function meets single failure criterion.
Vessel water level (L2) trip initiates main steam line isolatio n, recirculation pump trip and HPCS/RCIC system operation (not simulated). The recirculation pump motor circuit breakers then open causing decrease in core flow to natural circulation.
15.2.7.2.3 The Effect of Single Failures and Operator Errors
The nature of this event results in a lowering of vessel water le vel. Key correc tive efforts to shut down the reactor are automatic and designed to satisfy single failure criterion. Therefore, any additional failure in these shutdown methods woul d not aggravate or ch ange the simulated transient.
The potential exists for a single relief valve failing to close on ce it is opened. This would result in a complete depressurization of the reactor. Either the RCIC or the HPCS system is capable of maintaining adequate core coverage and will provide long-term inventory control.
For the complete loss of feedwa ter flow event, operation of RC IC or HPCS is sufficient to avoid initiation of ADS on low vessel level (L1).
15.2.7.3 Core and Sy stem Performance 15.2.7.3.1 Mathematical Model
The point-kinetics REDY model described in Section 15.0.3.3.1 was used to simulate this event.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-26 15.2.7.3.2 Input Paramete rs and Initial Conditions The simultaneous trip of both feedwater pumps is assumed to occur while the plant is operating at 106% uprated power and 100% core flow.
These analyses have b een performed, unless otherwise noted, with plan t conditions tabulated in Table 15.0-2.
15.2.7.3.3 Results
Table 15.2-11 provides the sequence of events and Figure 15.2-9 depicts the plant parameter responses. Key transient peak values are presented in Table 15.0-1. This event is non-limiting in that the pressurization event and change in MCPR are less severe than the Generator Load Rejection with Bypass Failure and Turbin e Trip with Bypass Failure events.
15.2.7.3.4 Consideration of Uncertainties
End-of-cycle scram characteristics are assumed.
This transient is most severe from high power conditions, because the rate of level decrease is greatest and the amount of stored decay heat to be dissi pated is highest.
15.2.7.4 Barrier Performance
Peak pressure in the bottom of the vessel reaches 1152 psig, wh ich is below the ASME Code limit of 1375 psig for the RCPB. Vessel dome pressure does not exceed 1142 psig. The consequences of this event do no t result in any temperature or pr essure transient in excess of the criteria for which the fuel , pressure vessel, or containment are designed. Therefore, barrier integrity and function are maintained.
15.2.7.5 Radiological Consequences
The consequence of this event does not result in fuel failure. It does result in the discharge of normal coolant activity to the suppression pool by means of SRV operation, which is contained in the primary containment.
This event does not result in an uncontrolled release to the environment, so the plant operator can choose to hold the activity in containment or discharge it when conditions permit. If purging of the containment is chosen the release will be in accordance with established requirements.
15.2.8 FEEDWATER LINE BREAK
See Section 15.6.6.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-27 15.2.9 FAILURE OF RESIDUAL HEAT REMOVAL SHUTDOWN COOLING This transient is classified as a nonlimiting event for both original and uprated power conditions.
Normally, in evaluating component failures associated with the RHR shutdown cooling mode of operation, active pumps or in strumentation (all of which are redundant for the safety related portions of the RHR system) would be assumed to be the component failure. For purposes of a worst case analysis, a valve on the single recirculation suction line to the otherwise redundant RHR shutdown cooling loops is assumed to fa il. Manual attempts to open the valve are assumed unsuccessful. Discovery is conservatively assumed to occur at 100 psig. This envelops discovery at normal RHR shutdown cooling operating limits (see Section 5.4.7). This failure disables the shutdown cooling mode but does not affect the remaining RHR modes of operation.
Reference 15.2-1 establishes additional assumptions.
15.2.9.1 Identification of Causes and Frequency Classification
15.2.9.1.1 Identification of Causes
The plant is operating at 105% of original NBR steam flow when an event occurs , e.g., a long-term loss of offsite power, causing a plant shutdown. Reactor vessel depressurization is initiated to bring the reactor pressure to approx imately 100 psig. Concurrent with the loss of offsite power a failure of a valve in the shutdown cooling suc tion line occurs which prevents the operator from establishing the normal shutdown cooling path through the RHR shutdown cooling lines. An additional fa ilure is assumed which complete ly disables the RHR equipment in one division. The operator then establishes a shutdown cooling path for the vessel through the SRV valves.
15.2.9.1.2 Frequenc y Classification
This event is categorized as an incident of mode rate frequency.
15.2.9.2 Sequence of Events and Systems Operation
15.2.9.2.1 Sequence of Events The sequence of events for this event is shown in Table 15.2-12.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-28 15.2.9.2.1.1 Identification of Operator Actions. For the early part of the transient, the operator actions are to restor e and maintain reactor water level. The ope rator should reestablish reactor cooling by one or more of the following:
- a. Maintain reactor water inventory with the RCIC (when single failure is not assumed to be a loss of Division 1 dc power) and HPCS systems, b. At approximately 10 minutes into the transient, initiate suppression pool cooling, it is assumed that only one RHR heat exchanger is available,
- d. Attempts to open one of the two RHR shutdown cooling su ction valves are assumed unsuccessful (reactor pressu re is approximately 100 psig), and
- e. Continue RPV depressurization by opening SRVs and es tablish a reactor cooling path as described in the notes for Figure 15.2-10.
Time required to initiate the nece ssary steps to maintain reactor pressure and level control is approximately 10 minutes.
15.2.9.2.2 Systems Operation
Plant instrumentation and contro l is assumed to be functioning normally except as noted. In this evaluation credit is taken for the plant and reactor protection systems and/or the ESF used.
15.2.9.2.3 The Effect of Single Failures and Operator Errors
The worst case single failure (loss of division power) has already b een analyzed in this event. Therefore, no single failure or operator error can increase the consequences of this event.
15.2.9.3 Core and Sy stem Performance
The earliest time the shutdown cooling system can be actuated is 2 to 3 hr after shutdown is initiated. During this time MCPR remains hi gh and nucleate boiling heat transfer is not exceeded at any time. Therefore, the core thermal safety margin remains essentially unchanged. The 10-minute time period approximated for operator act ion is an estimate of how long it would take the operator to initiate the necessary actions. It is not a time by which action must be initiated.
The transient behavior of the core during this event has been evaluated in Section 15.2.6.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-29 15.2.9.4 Results For most single failures that could result in loss of shutdown cooling, no unique safety actions are required. In these cases, shutdown cooling is simply reestablished using the redundant shutdown cooling loop. In cases where the RHR shutdown cooling suction line valves cannot be opened, alternate paths are available to accomplish the shutdown cooling function (Figure 15.2-11
). An evaluation has been performed a ssuming a failure that disables the RHR shutdown cooling suction line valves.
This evaluation demonstrates the capability to safely transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the RCPB are not exceeded.
The alternate cooldown path chosen to accomplish the shutdown coo ling function uses the RHR and ADS or normal relief valve systems (see Reference 15.2-1 and Figure 15.2-10
). The alternate shutdown systems are capable of performing the function of transferring heat from the reactor to the environment using only safety systems. The systems are capable of bringing the reactor to a cold shutdown in approximately 36 hr or less after the transient occurs.
The systems have suitable redundanc y in components such that ev en for onsite electrical power operation (offsite power is not available), th e safety function of the systems can be accomplished assuming an additional single failure. The systems can be fully operated from the main control room.
The design evaluation is divided into two phases: (a) full power operation to approximately 100 psig vessel pressure, and (b) approximately 100 psig vessel pressure to cold shutdown (14.7 psia 200°F) conditions.
15.2.9.4.1 Full Power to Approximately 100 psig
Independent of the event that initiated plant shutdown (whether it be a normal plant shutdown or a forced plant shutdown), the reactor is nor mally brought to approx imately 100 psig using either the main condenser or, in the case wher e the main condenser is unavailable, the HPCS and RCIC systems together with the nuclear bo iler pressure relief syst em and the RHR heat exchanger in the suppression pool cooling mode.
For evaluation purposes, however, it is assumed that plant shutdown is initiated by a transient event (loss of offsite pow er) which results in relief valve actuation and subs equent suppression pool heatup. For this postulated condition, the reactor is shut down and the reactor vessel pressure is reduced to approximately 100 psig. Manual operation of the SRVs is used to depressurize the reactor vessel. Reactor ve ssel makeup water is automatically provided by means of the RCIC (until reduced vessel pressu re is reached) and HPCS systems. While in C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-30 this condition, the RHR system (suppression pool c ooling mode) is used to maintain the suppression pool temperature within shutdown limits.
These systems are designed to routinely perform their functions for both normal and forced plant shutdown. Since the HPCS and RHR syst ems are divisionally separated and the HPCS and RCIC systems are divisionally separated, no single failure together with the loss of offsite power, is capable of preventing reaching the 100 psig level.
15.2.9.4.2 Approximately 100 psig to Cold Shutdown
The following assumptions are used for the analyses of the pr ocedures for attaining cold shutdown from a pressure of approximately 100 psig:
- a. The vessel is at 100 ps ig and saturated conditions,
- b. A worst-case single failure is assume d to occur (i.e., loss of a division of emergency power), and
- c. There is no offsite power available.
In the event that the RHRs shutdown suction line is not avai lable because of single failure, the first action to be taken will be to control reactor pressure. If a single electrical failure caused the suction line to fail in the closed position, a hand wheel is provided on the valve to allow manual operation. If for some reason the normal shutdown c ooling suction line cannot be restored to service, the capabilities described below will satis fy the normal shutdown cooling requirements and thus fully comply with GDC 34.
The RHR shutdown cooling line valves are in two divisions (Division 1 - the outboard valve, and Division 2 - the inboard valve) to satisfy containment isolation criteria. For evaluation purposes, the worst-case failure is assumed to be the loss of a division of emergency power, since this also prevents electr ical actuation of one shutdown c ooling line valve. Engineered safety features equipment and safe shutdown RCIC equipment (until reduced reactor pressure is reached) available for accomplishing the shutdo wn cooling function include (for the selected path):
ADS (dc Division 1 and dc Division 2)
RHR Loop A (Division 1)
HPCS (Division 3)
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-31 RCIC (dc Division 1)
LPCS (Division 1)
Since availability or failure of Division 3 equipment does not aff ect the normal shutdown mode, normal shutdown cooling is easily available through equipment powered from only Divisions 1 and 2. It should be noted that, HPCS is always available for cool ant injection if either of the other two divisions fails. For failure of Division 1 or 2, the following systems are assumed functional:
Assuming the single failure is a failure of Division 1 emergency power, the safety function is accomplished by establishing one of the coo ling loops described in Activity C1 of Figure 15.2-10. b. Division 2 Fails, Divi sion 1 and 3 Functional
Failed Systems Functional Systems
RHR Loop B and C HPCS ADS RHR Loop A LPCS RCIC (until reduced reactor pressure is reached)
Assuming the single failure is the failure of Di vision 2, the safety f unction is accomplished by establishing one of the cooling loops described in Activity C2 of Figure 15.2-10. Figures 15.2-12 through 15.2-15 show RHR loops A, B, and/or C (simplified).
15.2.9.4.3 Temperature Response - 3462 MWt
The reactor vessel temperature and pressure response versus time fo r the core conditions defined in Table 15.2-13 (105% of original rated stea m flow, 3462 MWt rated power) are presented in Figures 15.2-16 and 15.2-17. Figure 5.2-16 presents the results for C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-32 Activity C1.b.1 or C2 described in Figure 15.2-10. Figure 5.2-17 presents the results for Activity C1.b.2. The bulk suppression pool temper ature responses from th e same analysis are presented in Figures 15.2-18 and 15.2-19. Figure 5.2-18 presents the results for Activity C1.b.1 or C2 and Figure 5.2-19 presents the result s for Activity C1.b.2.
15.2.9.4.4 Temperature Response - 3702 MWt
Reference 15.2-7 analyzed the same two s cenarios (Activities C2 a nd C1.b.2) for 106.2% of power uprate conditions (3702 MW t) to determine the peak bul k suppression pool temperature and the time required to cool th e reactor vessel to cold shutdo wn (14.7 psia and 200°F). The analysis at power uprate conditions calculated a relative 4°F increase in the peak bulk pool temperature due to the power uprate. However, the peak temperature calculated was lower than the temperatures presented in this section due to the use of more realistic assumptions.
These assumptions include a more realistic decay heat model, lower initial suppression pool temperature (90°F), and more realistic treatment of pump heat addition.
15.2.9.5 Barrier Performance
As noted above, the consequences of this event do not result in any temperature or pressure transient in excess of the criteria for which the fuel, pressure vessel, or containment are designed. Release of coolant to the containment occurs by means of SRV actuation.
15.2.9.6 Radiological Consequences
The consequence of this event does not result in fuel failure. It does result in the discharge of normal coolant activity to the suppression pool by means of SRV operation, which is contained in the primary containment.
This event does not result in an uncontrolled release to the environment, so the plant operator can choose to hold the activity in containment or discharge it when conditions permit. If purging of the c ontainment is chosen, th e release would be in accordance with established requirements.
15.2.10 REFERENCES
15.2-1 Letter - R. S. Boyd to I. F. Stua rt; dated November 12, 1975.
Subject:
Requirements delineated for RHRS - Shutdown Cooling System - Single Failure Analysis.
15.2-2 NEDC-24154-P-A, "Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors," Volume s 1, 2, 3 and 4, February 2000.
15.2-3 Supplemental Reload Licensing Repor t for Columbia (m ost recent version referenced in COLR).
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-09-007 15.2-33 15.2-4 R. B. Linford, "Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor," NEDO 10802, April 1973.
15.2-5 For Power Uprate: GE Nuclear En ergy, "WNP-2 Power Uprate Transient Analysis Task Report," GE-NE-208 0393, September 1993 (Proprietary).
15.2-6 Deleted.
15.2-7 GE Nuclear Energy, "WNP-2 Powe r Uprate Project NSSS Engineering Report," GE-NE-208-17-0993, Revision 1, December 1994 (Proprietary).
15.2-8 AREVA NP, Inc., "Columbia Generati ng Station MSIV Closure Level Setpoint Change - Loss of Feedwater Flow Tr ansient Analysis," 51-9084418-000, July 2008.
15.2-9 GE Hitachi Nuclear Energy, "Li cense Amendment Request for Proposed Changes to Columbia Technical Spec ifications: Changing Group 1 Isolation Valves' Low Reactor Water Level Isolation Signal from the Current Level 2 to Level 1," 0000-0081-6730-R1, July 2008.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-35 Table 15.2-1 Sequence of Events for Figure 15.2-1
Pressure Regulator Failure - Closed 104.1% Uprated Power - 106% Flow Time (sec) Event 0 Failure of the pressure regulator causes closure of the turbine control valves. 1.0 Scram signal initiated at high neutron flux. 2.0 Recirculation pump motor circuit breakers open causing decrease in core flow to the natural circulation. 2.56 Group 3 relief valves actuated. 2.67 Group 4 relief valves actuated. 2.77 Group 5 relief valves actuated. 2.84 Turbine control valves close.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-36 Table 15.2-2 Sequence of Events for Figure 15.2-2.1
Generator Load Rejection with Bypass On Original Rated Power Time (sec) Event (-)0.015 a Turbine generator detection of loss of electrical load.
0 Turbine generator overspeed protection control (OPC) devices trip to initiate turbine control (governor) valve fast closure.
0 Turbine generator OPC trip in itiates main turbine bypass system operation.
0 Fast control valve closure initiates scram trip.
0 Fast control valve closure initiates an RPT. 0.07 Turbine control valves closed. 0.11 Turbine bypass va lves start to open. 0.19 Recirculation pump motor circuit breakers open causing decrease in core flow to natural circulation. 1.70 Group 1 relief valves actuated. 1.86 Group 2 relief valves actuated. 2.01 Group 3 relief valves actuated. 2.27 Group 4 relief valves actuated.
a Approximately.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-37 Table 15.2-3 Sequence of Events for Figure 15.2-2.2
Generator Load Rejection with Bypass Failure 100% Power 106% Flow Time (sec) Event (-)0.003 a Turbine generator detection of loss of electrical load.
0 Turbine generator OPC devices trip to initiate turbine control (governor) valve fast closure.
0 Turbine bypass valves fail to operate. 0.03 Time of scram trip. 0.15 Turbine control valves fully closed.
0.20 Time of RPT trip.
0.28 Start of control blade motion. (b) Group 1 MSRVs actuated (safety function). (b) Group 2 MSRVs actuated (safety function). 2.83 Group 3 MSRVs actua ted (safety function). 3.17 Group 4 MSRVs actua ted (safety function).
a Approximately.
b Not used - out of serv ice for this analysis.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-09-007 15.2-38 Table 15.2-4
Sequence of Events for Figure 15.2-3
Turbine Trip, Trip Sc ram - Bypass and RPT On Original Rated Power
Time (sec) Event 0 Turbine trip initiates closure of main stop (throttle) valves.
0 Turbine trip initiates bypass operation. 0.01 Main turbine stop va lves reach 90% open pos ition and initiate reactor scram trip. 0.01 Main turbine stop va lves reach 90% open pos ition and initiate an RPT. 0.10 Turbine stop valves closed. 0.10 Turbine bypass va lves start to open to regulate pressure. 0.20 Recirculation pump motor circuit breakers open causing decrease in core flow to natural circulation. 1.63 Group 1 relief valves actuated. 1.78 Group 2 relief valves actuated. 1.94 Group 3 relief valves actuated. 2.14 Group 4 relief valves actuated. 2.50 Group 5 relief valves actuated. 4.67 Feedwater turbines trip on L8 high water level.
5.1 a Group 5 relief valves start to close.
7.2 a All relief groups closed. 31.0 Turbine bypass st arts to close.
32.3 a Turbine bypass closed. 39.7 Turbine bypass reopens on pressu re increase at turbine inlet. 45.3 Main steam line isolation b, HPCS system initiation, and RCIC system initiation on low level (L2) (not included in simulation). 50+ Group 1 relief valves cycl e open and close on pressure.
a Estimated.
b The analysis has not been update d for the change in MSIV isolation setpoint from Level 2 to Level 1 because the analysis is bounding and conclusions of th e analysis are not affected (Reference 15.2-9).
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-39 Table 15.2-5 Sequence of Events for Figure 15.2-4
Turbine Trip w ith Bypass Failure at 100% Power/106% Core Flow Time (sec) Event 0 Turbine trip initiates closure of main stop (throttle) valves.
0 Turbine bypass valves fail to operate. 0.01 Main turbine stop va lves reach 90% open pos ition and initiate reactor scram trip. 0.10 Turbine stop valves closed. 0.20 Recirculation pump motor circuit breakers open causing decrease in core flow to natural circulation.
(a) Group 1 re lief valves actuated.
(a) Group 2 re lief valves actuated. 2.83 Group 3 relief valves actuated. 3.18 Group 4 relief valves actuated.
a Not used - out of service for this analysis.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-40 Table 15.2-6 Sequence of Events for Figure 15.2-5
Main Steam Line Isolation Valve Closure 106.2% Uprated Power, 100% Rated Flow Time (sec) Event (-) 0.003 (approximately) Turbine-Generator de tection of loss of electrical load.
0 Initiate closure of all main steam line valves.
0.45 MSIVs reach 85% opening initiating a position valve scram.
2.0 Loss of feedwater begins as turbine loses steam supply.
2.53 Both recirculation pumps trip due to high pressure.
3.0 All MSIVs closed. 3.08 Groups 3 relief valves actuated.
3.16 Groups 4 relief valves actuated.
3.24 Groups 5 relief valves actuated. 11 (approx.) All pressure relieve valves closed.
18.23 Groups 3 relief valves begin to cycle.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-41 Table 15.2-7 Sequence of Events for Figure 15.2-6
Loss of Condenser Vacuum 104.1% Uprated Power, 100% Rated Flow Time (sec) Event (-)5.0 (approximately) Initiate simulated loss of condenser vacuum at 2 in. of Hg per second. 0.00 Low condenser vacuum main tu rbine trip and feedwater turbine trips initiated.
0.00 Main turbine trip in itiates turbine bypass operation.
0.01 Main turbine stop valves re ach 90% open pos ition and initiate reactor scram.
0.19 Both recirculation pumps trip.
2.15 Group 3 relief valves actuated.
2.28 Group 4 relief valves actuated.
2.42 Group 5 relief valves actuated. 2.91 Feedwater recirculation valve is tripped.
5.00 Low condenser vacuum initia tes turbine bypass valve closure and MSIV closure. 5.6 (approx.) All relief valves closed. 6.0 (approx.) Main steam is olation valves closed.
7.90 Group 3 relief valves reactuated.
8.35 Group 4 relief valves reactuated.
24.01 Group 5 relief valves reactuated.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-42 Table 15.2-8 Trip Signals Associated w ith Loss-of-Condenser Vacuum Vacuum a Protective Action Initiated 27 to 30 Normal vacuum range. 20 to 23 Main turbine trip and feedwate r turbine trip (stop valve closures). 7 to 10 Main steam line isolation valve closure and bypass valve closure.
a Inches of Hg.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-43 Table 15.2-9 Sequence of Events for Figure 15.2-7
Loss of Auxiliary Power Transformers 106.2% Uprated Power, 100% Rated Flow Time (sec) Event 0.00 Loss of auxiliary power transformers occurs. 0.00 Recirculation system pump motors are tripped.
2.00 Reactor scram due to loss of power to the scram solenoid. 2.00 Main steam line isolation valves begin to close due to loss of power to MSIV solenoids. 4.00 Feedwater pumps are tripped due to MSIV closure.
5.98 Group 3 relief valves actuated. 6.15 Group 4 relief valves actuated. 6.34 Group 5 relief valves actuated.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-44 Table 15.2-10 Sequence of Events for Figure 15.2-8
Loss of All Grid Connections 104.1% Uprated Power, 100% Rated Flow Time (sec) Event 0.00 Loss of grid causes turbine-generator to detect a loss of electrical load. 0.00 Turbine-generator PL U devices trip to initiate TCV fast closure and turbine bypass system operation.
0.00 Recirculation pumps trip. 0.00 Fast control valve closure initiates reactor scram.
2.00 Main steam line isolation is initiated due to loss of power to the solenoids. 2.12 Group 3 relief valves actuated. 2.25 Group 4 relief valves actuated. 2.37 Group 5 relief valves actuated. 4.00 Feedwater pump tripped due to MSIV closure.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-09-007 15.2-45 Table 15.2-11
Sequence of Events for Figure 15.2-9
Loss of All Feedwater Flow 106.2% Uprated Power, 100% Rated Flow Time (sec) Event 0 Initiate trip of all feedwater pumps. 3.91 Recirculation runback initiated with narrow range sensed level less than L4 and feedwater pumps off.
7.38 Vessel water level (L3) trip initiates scram trip. 32.32 Vessel water level (L2) trip initiates main steam line isolation a , recirculation pump trip and HP CS/RCIC system operation (not simulated).
32.51 Recirculation pump motor circuit breakers open causing decrease in core flow to natural circulation.
a The analysis has not been update d for the change in MSIV isola tion setpoint from Level 2 to Level 1 because the analysis is bounding and conclusions of th e analysis are not affected (Reference 15.2-8).
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-46 Table 15.2-12 Sequence of Events for Failu re of Residual Heat Removal Shutdown Cooling Original Rated Power Time a Event 0 Reactor is operating at 105% NBR steam flow when LOP transient occurs initiating plant shutdown.
0 Concurrently loss of division power occurs (i.e., loss of one diesel generator).
0 Initial suppression pool temperature at 95°F. 10 minutes Suppression pool cooling initiated to prevent overheating from SRV actuation. 10 minutes Controlled blowdown initiated. 2-3 hr Blowdown to 100 psi completed. 2-3 hr Personnel are sent in to open RHR shutdown co oling suction valve and fail. 2.5-3.5 hr Complete blowdown to suppression pool by opening SRVs. 2.5-3.5 hr Redirect RHR pump discharge from pool to vessel by means of the LPCI line. Alternate coo ling path now established. 7 hr Maximum suppression pool temperature attained.
a Approximately.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.2-47 Table 15.2-13 Evaluation of Failure of Residual Heat Removal Shutdown Cooling Parameter Value Initial power corresponding 105% original rated steam flow To suppression pool mass (lbm) 8.52 E6 Residual heat removal (KHX value) (Btu/sec/°F) 289 Initial vessel condition Pressure (psia) 1055 Temperature (°F) 550.7 Initial primary fluid inventory (lbm) 7.016 E5 Initial pool temperature (°F) 95 Service water temperature (°F) 87 Vessel heat capacity (Btu/lbm/°F) 0.123 High-pressure core spray on-off water level (ft)
HPCS ON 40.8 HPCS OFF 47 High-pressure core spray flow rate (lbm/sec) 868 Low-pressure coolant injection flow rate (lbm/sec) 982 Figure Form No. 960690 LDCN-07-011 Draw. No.Rev.060108.06 15.2-1 Pressure Regulator Failure - Down Scale Failure at 104.1% Uprated Power, 106% Flow Columbia Generating Station Final Safety Analysis Report Amendment 59 December 2007 Figure Form No. 960690 LDCN-07-011 Draw. No.Rev.020361.68 15.2-2.1 Columbia Generating Station Final Safety Analysis Report Generator Load Rejection with Bypass On -
Original Rated Power Amendment 59 December 2007
-25.0 25.0 75.0 125.0 175.0 225.0 275.00.01.02.03.04.05.06.07.0 Time (sec)% Rated Vessel Press Rise (psi)
Safety Valve Flow Relief Valve Flow Bypass Valve Flow
-50.0-25.0 0.0 25.0 50.0 75.0 100.0 125.00.01.02.03.04.05.06.07.0 Time (sec)% Rated Level - Inch above Sep Skirt Vessel Steam Flow Turbine Steam Flow Feedwater Flow
-2.0-1.5-1.0-0.5 0.0 0.5 1.0 1.5 2.0 2.50.01.02.03.04.05.06.07.0 Time (sec)
R eactivity Com ponents ($) Void Reactivity Doppler Reactivity Scram Reactivity Total Reactivity 0.0 50.0 100.0 150.0 200.0 250.0 300.00.01.02.03.04.05.06.07.0 Time (sec)% Rated Neutron Flux Average Surface Heat Flux Core Inlet Flow Core Inlet Subcooling 15.2-2.2Generator Load Rejection with BP Failure Figure Form No. 960690 LDCN-08-035 Draw. No.Rev.020361.69 Columbia Generating Station Final Safety Analysis Report Amendment 60 December 2009 Amendment 53 November 1998 Columbia Generating Station Final Safety Analysis Report Turbine Trip, Trip Scram, Bypass and RPT - On 020361.70 15.2-3 Rev.Figure Draw. No.Form No. 960690 LDCN-08-035 Amendment 60 December 2009 Columbia Generating Station Final Safety Analysis Report 020361.71 15.2-4 Turbine Trip with Bypass Failure
-25.0 25.0 75.0 125.0 175.0 225.0 275.00.01.02.03.04.05.06.07.0 Time (sec)% Rated Vessel Press Rise (psi)
Safety Valve Flow Relief Valve Flow Bypass Valve Flow
-50.0-25.0 0.0 25.0 50.0 75.0 100.0 125.00.01.02.03.04.05.06.07.0 Time (sec)% Rated Level - Inch above Sep Skirt Vessel Steam Flow Turbine Steam Flow Feedwater Flow
-2.0-1.5-1.0-0.5 0.0 0.5 1.0 1.5 2.0 2.50.01.02.03.04.05.06.07.0 Time (sec)
R eactivity Com ponents ($) Void Reactivity Doppler Reactivity Scram Reactivity Total Reactivity 0.0 50.0 100.0 150.0 200.0 250.0 300.00.01.02.03.04.05.06.07.0 Time (sec)% Rated Neutron Flux Average Surface Heat Flux Core Inlet Flow Core Inlet Subcooling Figure Form No. 960690 LDCN-07-011 Draw. No.Rev.020361.72 15.2-5 Main Steam Line Isolation Valve Closure at 106.2% Uprated Power, 100% Rated Flow Columbia Generating Station Final Safety Analysis Report Amendment 59 December 2007 Figure Form No. 960690 LDCN-07-011 Draw. No.Rev.15.2-6 Loss of Condenser Vacuum at 104.1% Uprated Power, 100% Rated Flow Columbia Generating Station Final Safety Analysis Report Amendment 59 December 2007 020361.73 Figure Form No. 960690 LDCN-07-011 Draw. No.Rev.15.2-7 Loss of Auxiliary Power Transformers -
at 106.2% Uprated Power, 100% Rated Flow Columbia Generating Station Final Safety Analysis Report Amendment 59 December 2007 020361.74 Figure Form No. 960690 LDCN-07-011 Draw. No.Rev.15.2-8 Loss of All Grid Connections -
104.1% Uprated Power, 100% Rated Flow Columbia Generating Station Final Safety Analysis Report Amendment 59 December 2007 020361.75 Figure Form No. 960690 LDCN-07-011 Draw. No.Rev.15.2-9 Loss of All Feedwater Flow -
106.2% Uprated Power, 100% Rated Flow Columbia Generating Station Final Safety Analysis Report Amendment 59 December 2007 020361.76 Figure Amendment 59 December 2007 Form No. 960690 LDCN-07-011 Draw. No.Rev.900547.63 15.2-10.1 Columbia Generating Station Final Safety Analysis Report Automatic Depressurization System/Residual Heat Removal Cooling LoopsADS Valve RHR Loops B and C (Division 2 and 3Available)
Division 2 Fails Division 1 Fails C1 C2 DepressurizeVessel Via ADS/ReliefValve Actuation RHR Suppression Pool Cooling AutomatedRelief Valve Actuation Loss ofOffsite PowerTransient B 1055 psia, 550°F to 100 psi, 330°F 100 psi, 330°F to 14.7 psi, 125°FNo Offsite PowerADS Valve RHR Loop A (Division 1 and 3 Available)
Normal Shutdown InitiatedP = 1055 psiaT = 550°F A Automatic Depressurization System/Residual Heat Removal Cooling Loops2.01-2.5177.875099 FigureForm No. 960690.veR.oN .warD Amendment 54 April 2000 ACTIVITY A Initial pressure =1055 psia Initial temperature = 550°F
For purpose of this analysis, the following worst-case conditions are assumed to exist:
- a. The reactor is assumed to be operating at 105% of original NBR steam flow, b. A loss of power transient occurs,
- c. A simultaneous loss of onsite power (Division 1 or Division 2), and
- d. Operator unable to open one of the RHR shutdown cooling line suction valves.
ACTIVITY B Initial system pressure =1055 psia Initial system temperature = 550°FOperator Actions During approximately the first 30 minutes, reactor decay heat is passed to the suppression pool by the automatic operation
of the reactor relief valves. Reactor water level will be returned to normal by the HPCS and RCIC systems automatic operation.
After approximately 10 minutes, the operator initiates depressurization of the reactor vessel to control vessel pressure.
Controlled depressurization procedure consists of controlling vessel pressure and water level by using the SRV or HPCS and/or RCIC systems. After approximately 15 minutes, it is assumed one RHR heat exchanger is placed in the suppression
pool cooling mode to remove decay heat. At this time, the suppression pool will be 121°F.
When the reactor pressure approaches 100 psig, the operator would normally prepare for operation of the RHR system in the shutdown cooling mode. At this time (121 minutes), the suppression pool will be 186°F.
ACTIVITY C1 (Division 1 fails, Division 2 available)
System pressure =100 psig System temperature = 330°FOperation Actions The operator establishes a closed cooling path as follows:
- a. A minimum of two ADS valves (dc Division 2) are powered open.
- b. Either of the following cooling paths are established:
- 1. Using RHR loop B, water from the suppression pool is pumped through the RHR heat exchanger (where a portion of the decay heat is removed) into the reactor vessel.
The cooled suppression pool water flows through the vessel (picking up a portion of the decay heat) out the ADS valves and back to the suppression pool.
This alternate cooling path is shown in Figure 15.2-12
.
- 2. Using RHR loops B and C together, water is taken from the suppression pool and pumped directly into the reactor vessel. The water passes through the vessel (picking up decay heat) and out the ADS valves returning to the suppression pool as shown in Figure 15.2-13. Suppression pool water is then cooled by operation of RHR loop B in the pool cooling mode (see Figure 15.2-14
). In this alternate cooling path, RHR loop C is used for injection and RHR loop B for cooling. Cold shutdown is achieved approximately 36 hr after the transient occurs.ACTIVITY C2 (Division 2 fails, Division 1 available) (Figure 15.2-15
)System pressure =100 psig
System temperature = 330°F Operator Actions The operator establishes a closed cooling path as follows:
- a. A minimum of two ADS valves (dc Division 1) are powered open, and
- b. Using RHR loop A instead of loop B, an alternate cooling path is established as shown in Activity C1. Cold shutdown is reached in approximately 15 hr.
NOTES Columbia Generating Station Final Safety Analysis Report Summary of Paths Available to Achieve Cold Shutdown 900547.6415.2-11 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Normal Shutdown InitiatedP = 1055 psiaT = 550 FLoss of OffsitePower Transient A 1055 psia, 550 F to 100 psi, 330 F 100 psi, 330 F to 14.7 psi, 125 F Normal ShutdownCooling (Offsite Power)CondenserNot Available AutomatedRelief Valve Actuation DepressurizeVessel Via Manual ReliefValve Actuation +
RHR Suppression Pool Cooling DepressurizeVessel to Main CondenserOffsite Power Available ADS/ReliefValve Alternate Path Normal Shutdown Cooling (Onsite Power)ADS/ReliefValve Alternate Path ADS/ReliefValve Alternate Path Failure of Division I DieselLoss of Offsite Power Failure of Div III Diesel Failure of Division III Diesel DepressurizeVessel Via Manual ReliefValve Actuation +
RHR Suppression Pool Cooling Failure of Shutdown CoolingSuction Valve Cold Shutdown Achieved(Vessel Head Removed)D Columbia Generating StationFinal Safety Analysis Report Activity C1 Alternate Shutdown Cooling Path Utilizing Residual Heat Removal Loop B 900547.65 15.2-12 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Suppression Area ReactorVessel MSLS/R Valve ServiceWater SystemService Water Discharge Heat Exchanger Columbia Generating StationFinal Safety Analysis Report Residual Heat Removal Loop C 900547.66 15.2-13 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Suppression Area ReactorVessel MSLS/R Valve B Columbia Generating StationFinal Safety Analysis Report Residual Heat Removal Loop A(B) (SuppressionPool Cooling/ Rated Pump Flow Test Mode) 900547.67 15.2-14 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.ServiceWater SystemService Water Discharge Heat Exchanger Columbia Generating StationFinal Safety Analysis Report Activity C2 Alternate Shutdown Cooling PathUtilizing Residual Heat Removal Loop A 900547.68 15.2-15 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Suppression Area ReactorVessel MSL ServiceWater SystemService Water Discharge Heat Exchanger DrywellRHR A S/RValve Columbia Generating StationFinal Safety Analysis Report Vessel Temperature and Pressure Versus Time(Activity C1.b.1 or C2) 900547.69 15.2-16 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Time (hr)Reactor Vessel Temperature (Reactor Vessel Pressure (psia) F)10 100 1000 0.1 1 10 100 10 100 1000Temperature Pressure Alternate Shutdown Mode Depressurization @ 100 F/hr 87F Service Water
95F Initial Pool Temperature LPCI Operates Through 1 Heat Exchanger Columbia Generating StationFinal Safety Analysis Report Vessel Temperature and Pressure Versus Time (Activity C1.b.2) 900547.70 15.2-17 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Time (hr)Reactor Vessel Temperature ( F)10 100 1000 0.1 1 10 100 10 100 1000Temperature Pressure Alternate Shutdown Mode Depressurization @ 100 F/hr 87F Service Water
95F Initial Pool Temperature LPCI Direct to the Vessel Columbia Generating StationFinal Safety Analysis Report Reactor Vessel Pressure (psia)
Suppression Pool Temperature Versus Time (with 87F Service Wate r Temperature)(Activity C1.b.1 or C2) 900547.71 15.2-18 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Time (hr)240 Alternate Shutdown Mode Depressurization @ 100 F/hr 87F Service Water
95F Initial Pool Temperature LPCI Operates Through 1 Heat Exchanger 200 160 120 80 40 0.1 1 10 100 1000 Columbia Generating StationFinal Safety Analysis Report Suppression Pool Temperature (°F)
Suppression Pool Temperature Versus Time (with 87 F Service Water Temperature) (Activity C1.b.2) 900547.72 15.2-19 Figure Amendment 53 November 1998 Form No. 960690 Draw. No.Rev.Time (hr)240 Alternate Shutdown Mode Depressurization @ 100 F/hr 87 F Service Water
95 F Initial Pool Temperature
LPCI Direct to the Vessel 200 160 120 80 40 0.1 1 10 100 1000 Columbia Generating Station Final Safety Analysis Report Suppression Pool Temperature (°F)
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-07-011 15.3-1 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE
15.3.1 RECIRCULATION PUMP TRIP
The events for two-recirculati on pump operation are not limiting, th erefore, the analyses have not been updated since the reactor power uprate analyses.
15.3.1.1 Identification of Causes and Frequency Classification 15.3.1.1.1 Identification of Causes
Recirculation pump motor operation can be tripped by design and by random operational failures. Design tripping will occur in response to:
- a. Reactor vessel water level L2 setpoint trip,
- b. Turbine control (governor) valve fast closure or stop (throttle) valve closure, c. Failure to scram high pressure setpoint trip, d. Motor branch circuit over-current protection, e. Motor overload protection, and f. Suction block va lve not fully open.
Random tripping will occur in response to:
- a. Operator error,
- b. Loss of electrical power source to the pumps, and
- c. Equipment or sensor failures and ma lfunctions which initiate the above intended trip response.
15.3.1.1.2 Frequenc y Classification
15.3.1.1.2.1 Trip of One Recirculation Pump. This event is categorized as an incident of moderate frequency.
15.3.1.1.2.2 Trip of Two Recirculation Pumps. This event is categori zed as an incident of moderate frequency.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-07-011 15.3-2 15.3.1.2 Sequence of Events and Systems Operation
15.3.1.2.1 Sequence of Events
15.3.1.2.1.1 Trip of One Recirculation Pump. Table 15.3-1 lists the sequence of events for Figure 15.3-1.
15.3.1.2.1.2 Trip of Two Recirculation Pumps. Table 15.3-2 lists the sequen ce of events for Figure 15.3-2.
15.3.1.2.2 Systems Operation
15.3.1.2.2.1 Trip of One Recirculation Pump. Tripping a single recirculation pump requires no protection system or safeguard system operati on. This analysis as sumes normal functioning of plant instrumentation and controls.
15.3.1.2.2.2 Trip of Two Recirculation Pumps. Analysis of this event assumes normal functioning of plant instrumentation and controls and plant and reactor protection systems.
Specifically, this transien t takes credit for vessel le vel (L8) instrumentation to trip the turbine. Reactor shutdown relies on scram trips from the turbine stop (throttle) valves. High system pressure is limited by the pressu re relief valve system operation.
15.3.1.2.3 The Effect of Single Failures and Operator Errors
15.3.1.2.3.1 Trip of One Recirculation Pump. None 15.3.1.2.3.2 Trip of Two Recirculation Pumps. Table 15.3-2 lists the vessel level (L8) trip event as the first response to in itiate corrective action in this transient and it is intended to prohibit moisture carryover to the main turbine.
Multiple level sensors are used to sense and detect when the water level reaches the L8 setpoint. At this point, a single failure will neither initiate nor impede a turbine trip signal. Turb ine trip signal transmission circuitry, however, is not built to single failure criterion. At this point the transient even t is functionally over.
Scram trip signals from the turbine are designed such that a single failure will neither initiate nor impede a reactor scram trip initiation.
15.3.1.3 Core and Sy stem Performance
15.3.1.3.1 Mathematical Model
The point-kinetics REDY model described in Section 15.0.3.3.1 is used to simulate this event.
C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 15.3-3 15.3.1.3.2 Input Paramete rs and Initial Conditions These analyses have been pe rformed, unless otherwise noted , with plant conditions in Table 15.0-2.
Pump motors and pump rotors are simulated with minimum specified rotating inertias.
15.3.1.3.3 Results
15.3.1.3.3.1 Trip of One Recirculation Pump. Figure 15.3-1 shows the response of the reactor system following the trip of one recirculation pump motor. Initially a recirculation pump is tripped in one loop, causing the core inlet flow to decr ease, while the other recirculation loop flow increases. Subsequently jet pump diffuser flow reverses in the tripped recirculation loop. At approximately 45 sec the reactor reaches a new equilibrium operating point, at approximately 75% power and 57% core flow. During th e transient, level swell is not sufficient to cau se turbine trip.
15.3.1.3.3.2 Trip of Two Recirculation Pumps. Figure 15.3-2 shows the response of the reactor system following the trip of both recirculation pump motors. Initially both
recirculation pumps are tripped, causing the core inlet flow to decrease, wh ile vessel level rises until both main and feedwater turbines trip on high level (L8). A reactor scram is subsequently initiated at 90% tu rbine stop valve position. Shortly after the scram is initiated the stop valves close and the bypass valves open to regulate pressure. At this point the transient event is functionally over.
15.3.1.3.4 Consideration of Uncertainties
Initial conditions chosen for these analyses are conservative and te nd to force anal ytical results to be more severe than expect ed under actual plant conditions.
Actual pump and pump-motor drive line rotating inertias are expected to be somewhat greater than the minimum design values assumed in this simulation. Actual plant deviations regarding inertia are expected to lessen the severity as analyzed.
Minimum design iner tias were used as well as the least negative void coefficient since these maximize the flow reduction.
15.3.1.4 Barrier Performance 15.3.1.4.1 Trip of One Recirculation Pump
Figure 15.3-1 results indicate a basic reduction in syst em pressures from the initial conditions. Therefore, the reactor cool ant pressure boundary (RCPB) barrier is not impacted.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-07-011 15.3-4 15.3.1.4.2 Trip of Two Recirculation Pumps
The results shown in Figure 15.3-2 indicate peak pressures stay well below the limit allowed by the applicable American Society of Mechanical Engineers (ASME) co de. Therefore, the RCPB barrier is not impacted.
15.3.1.5 Radiological Consequences
The consequence of this event does not result in fuel failure. It does result in the discharge of normal coolant activity to the suppression pool by means of safety/relief valve (SRV) operation, which is contained in the primary c ontainment. This even t does not result in an uncontrolled release to the environment, so the plant operator can choose to hold the activity in containment or discharge it when conditions permit. If purging of the containment is chosen, the release will be in accordance with established requirements.
15.3.2 RECIRCULATION FLOW CONTROL FAILURE - DECREASING FLOW
15.3.2.1 Identification of Causes and Frequency Classification
15.3.2.1.1 Identification of Causes
A postulated failure of the input demand signal, which is used in both loops, can decrease core
flow at the maximum ramp de mand rate established by the adjustable speed drive (ASD) control. Failure within either loop controller can result in a maximum ramp demand rate as
limited by the ASD control.
15.3.2.1.2 Frequenc y Classification
This event is categorized as an incident of mode rate frequency.
15.3.2.2 Sequence of Events and Systems Operation
15.3.2.2.1 Sequence of Events
15.3.2.2.1.1 Speed Decrease of One Recirculation Pump. Table 15.3-3 lists the sequence of events for Figure 15.3-3.
15.3.2.2.1.2 Speed Decrease of Two Recirculation Pumps. Table 15.3-4 lists the sequence of events for Figure 15.3-4.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-07-011 15.3-5 15.3.2.2.2 Systems Operation
15.3.2.2.2.1 Speed Decrease of One Recirculation Pump. The most severe control system disturbance is a failure that cau ses the ASD internal controller to move at its maximum rate.
Such transients may be obtained by instantaneous failure of a controller output into its upper or lower limits. Originally the recirculation flow was controlled by valve motion. For the current analysis the recirculation flow control valves have been locked at the full open position, and ASD units have been implemented to provide the necessary flow control.
15.3.2.2.2.2 Speed Decrease of Two Recirculation Pumps. The most severe control system disturbance is a failure that cau ses the ASD internal controller to move at its maximum rate.
Such transients may be obtained by instantaneous failure of a controller output into its upper or lower limits. The independent and simultaneous failu re of each indivi dual loop controller would be highly improbable.
Thus, for the two loop controller failure event, the ASD internal controller is assumed to move
at its maximum rate in both r ecirculation loops. Originally the recirculation flow was controlled by valve motion. For the current analysis the recirculation flow control valves have been locked at the full open position, and ASD units have been implemented to provide the necessary flow control.
15.3.2.2.3 The Effect of Single Failures and Operator Errors
The single failure and operator considerations for this event are essentially the same as in Section 15.3.1.2.3.2. The speed decrease of two instead of one recircula tion pump would be the envelope case for the additional singl e component failure or operator error.
15.3.2.3 Core and Sy stem Performance
15.3.2.3.1 Mathematical Model
The point-kinetics REDY model described in Section 15.0.3.3.1 is used to simulate these transient events.
15.3.2.3.2 Input Paramete rs and Initial Conditions
These analyses have been perf ormed, unless otherwise noted, w ith plant conditions listed in Table 15.0-2.
15.3.2.3.2.1 Speed Decrease of One Recirculation Pump. For the simulation of this event, a controller malfunction caus es a zero demand signal to be sent to one of the recirculation ASD units, while the plant is operating at 106% uprated power and 100% core flow. A control demand error (low) signal causes the ASD to adju st the recirculation pump speed demand rate
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-07-011 15.3-6 limit downward at an assumed rate of 25%/sec for one loop failure. The ensuing transient is similar to a recirculation pump trip.
15.3.2.3.2.2 Speed Decrease of Two Recirculation Pumps. For the simulation of this event, a controller malfunction causes a zero demand signal to be sent to both of the recirculation ASD units, while the plant is operating at 106% uprated power and 100% core flow. A control demand error (low) can cause the ASD units to adjust the recirculation pump speed downward in both loops at the 5%/sec pump speed rate limit.
15.3.2.3.3 Results
15.3.2.3.3.1 Speed Decrease of One Recirculation Pump. Figure 15.3-3 shows the response of the plant for this transient. Initially a negative recirculation pump speed demand is sent to the ASD due to a postulated controller failur
- e. The negative pump speed demand causes the diffuser flow to decrease, and eventually reverse, in the failed loop. At the same time the
active loop increases flow to comp ensate for the failed recircul ation loop. At approximately 45 sec the reactor reaches a new equilibrium op erating point, at approxi mately 74% power and 57% core flow. During the transi ent, level swell is not sufficie nt to cause turbine trip which would result in a reactor scram.
15.3.2.3.3.2 Speed Decrease of Two Recirculation Pumps. Figure 15.3-4 shows the response of the plant to this transient using the 5%/sec pump speed demand rate limit. Initially, a negative recirculation pump speed demand is sent to both AS D units due to a postulated controller failure. The negativ e pump speed demand causes the di ffuser flows to decrease in the failed loops. During the transient, level swell is not sufficie nt to cause turbine trip which would result in a reactor scram.
15.3.2.3.4 Consideration of Uncertainties
Initial conditions chosen for these analyses are conservative and te nd to force anal ytical results to be more severe than expect ed under actual plant conditions.
These analyses are unaffected by deviations in pump/pump motor and driveline inertias since it is the ASD controller that causes rapid recirculation decreases.
15.3.2.4 Barrier Performance 15.3.2.4.1 Speed Decrease of One Recirculation Pump
The pressure in the vessel dome is well below the vessel pressure limit.
The event does not result in a temperature or pressure transient in excess of the criteria for which the fuel, pressure vessel, or containment are designed. Therefore, barrier in tegrity and function is maintained.
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 15.3-7 15.3.2.4.2 Speed Decrease of Two Recirculation Pumps
The pressure in the vessel dome is well below the vessel pressure limit.
The event does not result in a temperature or pressure transient in excess of the criteria for which the fuel, pressure vessel, or containment are designed a nd these barriers maintain their integrity and function as designed.
15.3.2.5 Radiological Consequences
Since this event does not result in any fuel failures or any release of primary coolant to either the secondary containment or to the environment there are no radiological consequences associated with this event.
15.3.3 RECIRCULATION PUMP SEIZURE
The pump seizure accident for single loop operati on (SLO) is analyzed for the introduction of GE14 fuel into the Co lumbia reactor core.
15.3.3.1 Identification of Causes and Frequency Classification
The case of recirculation pump seizure represents the extremely unlikely event of instantaneous stoppage of the pump motor shaft of one recircul ation pump. This event produces a very rapid decrease of core flow as a result of the large hydraulic resistance introduced by the stopped rotor. The sudden decrea se in core coolant flow while the reactor is at full power results in a degradation of core heat transfer which c ould result in fuel damage.
The event is categorized as an infrequent incident when operating with two recirculation pumps in service. For single loop operation, this event is considered to be a limiting fault, but is analyzed as an incident of moderate frequency for Gl obal Nuclear Fuel reloads.
15.3.3.2 Sequence of Events and Systems Operation
15.3.3.2.1 Sequence of Events
Table 15.3-5 lists the sequence of events for Figure 15.3-5, for two loop operation.
Table 15.3-6 lists the sequence of events for the recirculation pump seizure accident during SLO.
Identification of Operator Actions
The operator must verify that the reactor scrams with the turbine trip resulting from reactor water level swell. The operato r should regain control of r eactor water level through RCIC
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-06-000,07-011 15.3-8 operation or by restart of a feedwater pump, and must monitor reactor water level and pressure control after shutdown.
15.3.3.2.2 Systems Operation
In order to properly simulate the expected sequence of events, the analysis of this event assumes normal functioning of plant instrumentati on and controls, plant pr otection, and reactor protection systems.
Operation of safe shutdown f eatures, including operation of the HPCS and RCIC systems though not included in this simulation, may be used to maintain adequate water level.
15.3.3.2.3 The Effect of Single Failures and Operator Errors
Single failures in the scram l ogic originating by means of the high vessel level (L8) trip are similar to the considerations in Section 15.3.1.2.3.2.
15.3.3.3 Core and Sy stem Performance
15.3.3.3.1 Mathematical Model
The point-kinetics REDY model described in Section 15.0.3.3.1 is used to simulate this event for two loop operation. The comput er model described in Reference 15.3-2 was used to simulate this event for SLO.
15.3.3.3.2 Input Paramete rs and Initial Conditions
This analysis has been performed for two loop operation, unless otherwise noted, with plant conditions tabulated in Table 15.0-2 , column "REDY (ASD Events
)". For the simulation of the event while in two loop ope ration, one recirculation pump was seized instantaneously (pump speed set to zero) while the plant is operating at 106%
uprated power and 100% core flow.
For single loop operation, the anal ysis has been performed, unle ss otherwise noted, with plant conditions tabulated in Table 15.0-2A , column "Original Rated Power". For the purpose of evaluating consequences to the fuel thermal limits, this transient event is assumed to occur as a consequence of an unspecified, instantaneous stoppage of the active recirculation pump shaft while the reactor is operating at 75% NBR power under SLO. Also, the reactor is assumed to be operating at thermally-limiting conditions. The void coefficien t is adjusted to the most conservative value, that is, the least negative value in Table 15.0-2A.
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 15.3-9 15.3.3.3.3 Results
Figure 15.3-5 presents the results of the ac cident for two loop operation.
Table 15.3-5 shows the sequence of events for this transient. Initially a recirculation pump is seized in one loop causing the flow in the seized l oop to reverse and the flow in the active loop to increase. As the flow in the seized loop decr eases, the vessel level rises until a turbine trip is initiated on high level, L8. Once L8 is reached, both feedwater pumps trip. A reactor scram is subsequently initiated due to 90% turbine stop (throttle) valve position. Shortly after the turbine trip is in itiated the stop valves close and the bypa ss valves open to regulate pressure. Simultaneously the active recirculation loop trip s due to the turbine trip. The MCPR does not decrease significantly before fuel surface heat flux begins dr opping enough to restore greater thermal margins. After the time at which MCPR occurs, heat flux decreases more rapidly than the rate at which heat is removed by the coolant and the CPR is less than 0.01.
Figure 15.3-6 presents the results of the event in SL O. Core coolant flow drops rapidly, reaching a minimum value of 25%
rated at about 2.0 sec.
The RRC pump seizure while in SLO is more limiting than the RRC pump seizure in two loop operation. See Table 15.0-1A.
15.3.3.3.3.1 Considerations of Uncertainties. Considerations of uncertainties are included in the analysis.
15.3.3.4 Barrier Performance
The bypass valves open to limit the pressure well within the range allowed by the ASME vessel code. The RCPB is not impacted by ove rpressure. Therefore, barrier integrity and function is maintained.
15.3.3.5 Radiological Consequences
Since this event does not result in any fuel failures or any release of primary coolant to either the secondary containment or to the environment there are no radiological consequences associated with this event.
15.3.4 RECIRCULATION PUMP SHAFT BREAK
15.3.4.1 Identification of Causes and Frequency Classification
The breaking of the shaft of a recirculation pump is considered a design ba sis accident event. It has been evaluated as a mild accident in re lation to other design basis accidents such as the loss-of-coolant accident.
The analysis has been conducted w ith consideration to a single or
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 15.3-10 two loop operation. Two loop operation represents the worst case since single loop operation is limited to approximately 75% power.
This postulated event is bounde d by the more limiting case of recirculation pump seizure.
15.3.4.1.1 Identification of Causes
The case of recirculation pump shaft breakage represen ts the unlikely event of rapid stoppage of the pump operation of one recirculation pump. This event produces a rapid decrease of core flow.
15.3.4.1.2 Frequenc y Classification
This event is categorized as an in cident of infre quent frequency.
15.3.4.2 Sequence of Events and Systems Operation
15.3.4.2.1 Sequence of Events
A postulated instantaneous break of the pump motor shaft of one recirculation pump as discussed in Section 15.3.4.1.1 will cause the core flow to decrease rapidly resulting in water level swell in the reactor vessel. When the vessel water level reaches the high water level setpoint (Level 8), a main turbine trip and feedwater pump trip will be initiated.
A reactor scram and the remaining recirculation pump trip will be initiated due to the turbine trip. Eventually the vessel water level will be controlled by HPCS and/or RCIC flow.
15.3.4.2.2 Systems Operation
Normal operation of plant instrumentation and control is assumed. This event takes credit for vessel water level (Level 8) inst rumentation to scram the reactor and trip the main turbine and feedwater pumps. High system pressure is limited by the pressure relief system operation.
Operation of HPCS and/or RCIC is expected in order to maintain adequate water level control.
15.3.4.2.3 The Effect of Single Failures and Operator Errors
Effects of single failures in the high vessel level (L8) trip are simi lar to the considerations in Section 15.3.1.2.3.2.
Assumption of single component failure or op erator error in other equipment has been examined and this has led to the conclusion that no other credible fa ilure exists for this event. Therefore, the bounding cas e has been considered.
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 15.3-11 15.3.4.3 Core and Sy stem Performance
The pump shaft break event is bounded by the pump seizure event. Sin ce this event is less
limiting than that event, only qualitative evaluation is provided.
Therefore, no discussion of mathematical model, input para meters, and consideration of un certainties, etc., is necessary.
15.3.4.3.1 Qualitative Results
If this unlikely event occurs, core coolant flow will drop rapidly. The level swell produces a trip of the main and feedwater tu rbines. A scram is initiated due to turbine trip. Since heat flux decreases more rapidly than the rate at whic h heat is removed by the coolant, there is no impact on thermal limits. Additionally, the bypass valves and the pot ential for a momentary opening of some of the SRVs limit the pressure well within the range allowed by the ASME vessel code. Therefore, the RCPB is not impacted by overpressure.
The severity of this pump shaft break event is bounded by the pump se izure event. In either of these two events, the re circulation drive flow of the af fected loop decreases rapidly.
In the case of the pump seizur e event, the loop flow decrease s faster than the normal flow coastdown as a result of the la rge hydraulic resistance introduced by the stopped rotor. For the pump shaft break event, the hydraulic resistance caused by the broke n pump shaft is less than that of the stopped rotor for the pump seizure ev ent. Therefore, the core flow decrease following a pump shaft break effect is slower than the pump seizure event. Thus, it can be concluded that the potential effects of the hypothetical pump shaft br eak accident are bounded by the effects of the pump seizure event.
15.3.4.4 Barrier Performance
The bypass valves and momentary opening of some of the SRVs limit the pressure well within the range allowed by the ASME vessel code. Therefore, the RCPB is not impacted by overpressure.
15.3.4.5 Radiological Consequences Since this event does not result in any fuel failures or any release of primary coolant to either the secondary containment or to the environment there are no radiological consequences associated with this event.
15.
3.5 REFERENCES
15.3-1 General Electric Company, "WNP-2 Power Uprate Tr ansient Analysis Task Report," GE-NE-208-08-039 3, September 1993.
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 15.3-12 15.3-2 NEDC-24154-P-A, "Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors," Volume s 1, 2, 3 and 4, February 2000.
15.3-3 Advanced Nuclear Fuels Corpora tion, "WNP-2 Single Loop Operation Analysis," ANF-87-11 9, September 1987.
C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 15.3-13 Table 15.3-1 Sequence of Events for Figure 15.3-1
Trip of One Recirculation Pump Motor Uprated Power Time (sec)
Event 0 Trip of one recirculation pump initiated.
9 Jet pump diffuser flow reverses in the tripped loop.
45 a Core flow and power level stab ilize at new equilibrium conditions.
a Approximately.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-07-011 15.3-14 Table 15.3-2
Sequence of Events for Figure 15.3-2
Trip of Both Recirculation Pump Motors Uprated Power
Time (sec) Event 0 Trip of both recirculation pumps initiated. 5.66 Vessel water level (L8) trip initiates turbine trip. 5.66 Feedwater pumps are tripped off. 5.67 Main turbine stop (throttle) valves reach 90% open position and initiate reactor scram trip. 5.76 Turbine bypass valves open.
C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 15.3-15 Table 15.3-3 Sequence of Events for Figure 15.3-3
Recirculati o n Flow Control Failure
Decreasing Flow in One Loop Uprated Power Time (sec)
Event 0 Initiate fast down scale of reci rculation pump speed in one loop.
4 a Jet pump diffuser flow reverses in the affected loop.
45 a Core flow and power level stab ilize at new equilibrium conditions.
a Approximately.
C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 15.3-16 Table 15.3-4 Sequence of Events for Figure 15.3-4
Recirculati o n Flow Control Failure
Decreasing Flow in Both Loops (5%/sec)
Uprated Power Time (sec)
Event 0 Initiate 5%/sec down scale of rec i rculation pump speed in both loops.
85 a Core flow and power level stab ilize at new equilibrium conditions.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-07-011 15.3-17 Table 15.3-5
Sequence of Events for Figure 15.3-5
One Recirculation Pump Seizure Uprated Power
Time (sec)
Event 0 Seizure of one recirc ulation pump initiated.
1 a Jet pump diffuser flow reverses in the seized loop. 4.40 Vessel water high level (L8) trip initiates a turbine trip. 4.40 Feedwater pumps are tripped off. 4.41 Main turbine stop (throttle) valves reach 90% open position and initiate reactor scram. 4.59 Active recirculation loop trips due to previous turbine trip.
a Approximately.
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 15.3-18 Table 15.3-6
Sequence of Events for Pump Seizure (for Single Loop Operation)
Time (sec)
Event 0.0 Recirculation pump motor trip off complete Single pump seizure was initiated; core flow decreases
~1.9 Reverse flow ceases in the idle loop
~6.0 Power and flow stabilize
910402.37 Amendment 60December 2009 LDCN-08-035 Columbia Generating StationFinal Safety Analysis Report Figure Form No. 960690Draw. No.Rev.SLO Recirculation Pump Seizure Results 15.3-6-50.0-25.0 0.0 25.0 50.0 75.0 100.0 125.0 0.0 2.0 4.0 6.0 8.010.012.0 Time (sec)% Rated Vessel Press Rise (psi)
Idle Loop Flow (%)
Pump Speed (% of Initial)
Active Loop Flow (%)
25.0 50.0 75.0 100.0 125.0 0.0 2.0 4.0 6.0 8.010.012.0 Time (sec)% Rated NR Level in ref. Sep Skirt Vessel Steam Flow Turbine Steam Flow Feedwater Flow 1.0 1.1 1.2 1.3 1.4 1.5 1.6 1.70.00.51.01.52.02.53.03.5 Time (sec)Cr itical Po we r Rati o GE14 ATRM10 0.0 25.0 50.0 75.0 100.0 0.0 2.0 4.0 6.0 8.010.012.0 Time (sec)% Rated Neutron Flux Average Surface Heat Flux Core Inlet Flow C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-07-011 15.4-1 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES
15.4.1 ROD WITHDRAWAL ERROR - LOW POWER
This transient is classified as a nonlimiting event for both original and uprated power conditions. Furthermore, the low power Rod Withdrawal Error (RWE) is not affected by power uprate and therefore, th e following qualitative analysis is valid for power uprate.
15.4.1.1 Control Rod Remova l Error During Refueling
15.4.1.1.1 Identification of Caus es and Frequency Classification
The event considered is inadvert ent criticality due to the complete withdrawal or removal of the most reactive rod during refueling. The prob ability of the initial causes alone is considered low enough to warrant it s being categorized as an infreque nt incident since there is no postulated set of circumstances which results in an inadvertent contro l rod withdrawal error (RWE) while in the refuel mode.
15.4.1.1.2 Sequence of Even ts and Systems Operation
15.4.1.1.2.1 Initial Control Rod Removal. During refueling operations, safety system interlocks provide assurance that inadvertent criticality does not occur because a control rod was removed or is withdrawn in coin cidence with another control rod.
15.4.1.1.2.2 Fuel Insertion With Control Rod Removed. To minimize the possibility of loading fuel into a cell containi ng no control rod, it is required that all control rods are fully inserted when fuel is being lo aded into the core. This requi rement is backed up by refueling interlocks on rod withdrawal a nd movement of the refueling platform. When the mode switch is in the "REFUEL" position, the interlocks prevent the platform from being moved over the core if a control rod is withdraw n and fuel is on the hoist. Likewise, if the refueling platform is over the core and fuel is on the hoist, control rod motion is blocked by the interlocks.
15.4.1.1.2.3 Second Control Rod Removal. When the platform is not over the core (or fuel is not on the hoist) and the mode switch is in the "REFUEL" position, only one control rod can be withdrawn. Any atte mpt to withdraw a second rod results in a rod block by the refueling interlocks.
Since the core is designed to meet shutdown requirements with the highest worth rod withdrawn, the core remains subcritical even with one rod withdrawn.
15.4.1.1.2.4 Contro l Rod Removal Without Fuel Removal. The design of the control rod, incorporating the velocity limiter, does not physi cally permit the upward removal of the control C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 15.4-2 rod without the simultaneous or prior removal of the four adjacent fuel bundles. This precludes any hazardous condition.
15.4.1.1.2.5 Effect of Single Failure and Operator Errors. If any one of the operations involved in initial failure or e rror is followed by any other singl e equipment failure or single operator error, the necessary safe ty actions are taken (e.g., rod block or scram) automatically prior to violation of any limits.
15.4.1.1.3 Core and System Performances
Since the probability of inadvertent criticality during refueli ng is precluded, the core and system performances were not analyzed. The withdrawal of the highest worth control rod during refueling will not result in criticality.
This is verified expe rimentally by performing shutdown margin checks. Add itional reactivity insertion is precluded by interlocks. As a result, no radioactive material is released from the fuel, making it unnecessary to assess any radiological c onsequences.
No mathematic models are involved in this event. The need for input parameters or initial conditions is not required as there are no results to report. C onsideration of uncertainties is not appropriate.
15.4.1.1.4 Barrier Performance
An evaluation of the barrier performance was not made for this even t since it is a highly localized event and does not result in any chan ge in the core pressu re or temperature.
15.4.1.1.5 Radiologi cal Consequences
An evaluation of the radiologi cal consequences was not made for this event since no radioactive material is released from the fuel.
15.4.1.2 Continuous Rod Withdraw al During Reactor Startup
15.4.1.2.1 Identification of Caus es and Frequency Classification This event is categorized as an infrequent incident. The proba bility of further development of this event is low because it is contingent upon the failure of the rod worth minimizer (RWM) system or failure of a second licensed operator (or technically qua lified member of the technical staff) observing the out-of-sequence rod selection concurrent w ith a high worth rod, out-of-sequence rod selection contrary to procedures, and operator disregard of continuous alarm annunciations prior to safety system actuation.
C OLUMBIA G ENERATING S TATION Amendment 55 F INAL S AFETY A NALYSIS R EPORT May 2001 LDC N-0 0-0 3 4 15.4-3 15.4.1.2.2 Sequence of Even ts and Systems Operation
15.4.1.2.2.1 Sequence of Events. Control RWEs are not consid ered credible in the startup and low power ranges. The RW M or second licensed operator (o r other technically qualified member of the technical staff) prevents the operator from selecting and withdrawing an out-of-sequence control rod.
Continuous control RWEs during reactor startup are precluded by the RWM or second qualified person. The RWM or second qualified person preven ts the withdrawal of an out-of-sequence control rod from 100% control rod density to 10% of rated thermal power.
15.4.1.2.2.2 Effects of Single Failure and Operator Errors. If any one of the operations involved in the initial failure or error is followed by another single component failure or single operator error, the necessary safe ty actions are automatically take n to preclude violation of any limits.
15.4.1.2.3 Core and System Performance
The performance of the RWM or second licensed operator (or technicall y qualified member of the technical staff) prevents e rroneous selection and withdrawal of an out-of-sequence control rod. Thus, core and system performance is not affected by such a single operator error.
No mathematical models are involved in this ev ent. The need for input parameters or initial conditions is not required as there are no results to report. C onsideration of uncertainties is not applicable.
15.4.1.2.4 Barrier Performance
An evaluation of the barrier performance was not performed for this event since there is no postulated set of circumstances fo r which this error could occur.
15.4.1.2.5 Radiologi cal Consequences
An evaluation of the radiologi cal consequences is not requi red for this event since no radioactive material is released.
C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December2005 LDCN-05-019 15.4-4 15.4.2 ROD WITHDRAWAL ERROR - AT POWER
15.4.2.1 Identification of Causes and Frequency Classifications
15.4.2.1.1 Identification of Causes
While operating in the power range in a normal mode of operation, the reactor operator makes a procedural error and withdraws the maximum worth control rod until the rod block monitor (RBM) system inhibits further withdrawal.
15.4.2.1.2 Frequenc y Classification
The probability of this event is considered low enough to warrant its be ing categorized as an infrequent incident. However, because of the lack of sufficien t frequency database, this event is considered an incident of moderate frequency.
15.4.2.2 Sequence of Events and Systems Operation
15.4.2.2.1 Sequence of Events
The sequence of events for this transient is presented in Table 15.4-1.
15.4.2.2.2 Systems Operation
The focal point of this event is localized to a small portion of the core; therefore, although reactor control and instrumentation is assumed to function normally, credit is taken only for the RBM system.
While operating in the power range in a norma l operational mode, the reactor operator makes a procedural error and withdraws the maxi mum worth control rod until the RBM system inhibits further withdrawal.
Under normal operating conditions the nearest local power range monitor (LPRM) would detect the peak linear power exceeding design limits and alarm. The operator would acknowledge the alarm and ta ke appropriate action.
If the RWE is severe, the RBM system woul d alarm, at which time the operator would acknowledge the alarm and take corrective action. Even fo r conditions such as highly abnormal control rod patterns, ope rator disregard of all alarms and warnings, a nd continuous control rod withdrawal, the RBM system will bl ock further withdrawal of the control rod before the fuel reaches the point of boiling tran sition or the 1% plastic strain limit imposed on the clad.
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 15.4-5 15.4.2.2.3 Effect of Single Fa ilure and Operator Errors
Operator errors do not impact the consequences of this event due to the single failure proof design of the RBM system.
15.4.2.3 Core and Sy stem Performance
15.4.2.3.1 Mathematical Model
The control RWE transient is classified as a "slow transient." A slow transient is a power increase transient that is sufficiently slow so that the assu mption that steady-st ate conditions are achieved at each time step is either realistic or conservative. Using this assumption, this transient is calculated using a steady state, three dimensi onal, coupled nuclear thermal hydraulics computer program PANACEA. All spatia l effects are included in the calculation.
A detailed discussion of the c ode is presented in Reference 15.4-4. The control RWE analysis has been performed to estimate the minimum critical power ratio (MCPR) and maximum linear heat generation rate (LHGR) in such a transient. A starting control rod pattern is established for the typical BWR reactor and a central control rod is withdrawn from the fully inserted position. R od withdrawal results in an increase of the LHGR and decrease of the cr itical power ratio (CPR). The computed maximum LHGR and minimum CPR are compared to values of other transients to establish operating limits for the reactor. The analysis determines the transient MCPR as a function of th e rod block monitor setpoint.
15.4.2.3.2 Input Paramete rs and Initial Conditions
The number of possible RWE transi ents is large due to the numb er of control rods and the wide range of exposures and power levels. In order to encompass all of the possible RWEs which could conceivably occur, a limiting anal ysis is defined such that a conservative assessment of the conse quences is provided. These conditions bound the effects of the RWE at lower power or flow conditions, including operation with only one reactor recirculation pump.
- a. The assumed error is a continuous withdrawal of the maximum worth rod at its maximum drive speed;
- b. The core is assumed to be operating at rated conditions;
- c. The reactor is presumed to be in its most reactive state and devoid of all xenon. This ensures that the amount of reactivity is a maximum;
- d. It is assumed that the operator has fully inserted the maximum worth rod prior to its removal and selected the remaining control rod pattern in such a way as to
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-029 15.4-6 approach thermal limits in the fuel bundl es in the vicinity of the rod to be withdrawn (this control rod configura tion would only be ach ieved by deliberate operator action or by numer ous operator errors);
- e. The operator is assumed to ignore all warnings during the transient;
- f. Of the four LPRM strings nearest to the control rod being withdrawn, the two highest reading LPRMs during the transient are assumed to have failed; and
- g. One of the two instrument channels is assumed to be bypassed and out-of-service. The A and C LPRM chambers i nput to one channel, while the B and D chambers input to the othe
- r. The channel with the greatest response is assumed to be bypassed.
15.4.2.3.2.1 Rod Block Monitor System Op eration. The RBM system minimizes the consequences of a RWE by blocking motion of th e control rod before the safety limits are exceeded.
The RBM has three trip levels (rod withdrawal pe rmissive removed). Th e trip levels may be adjusted and are nominally 8% of reactor power apart. The highest trip level is set so that the safety limit is not exceeded. Th e lower two trip levels are inte nded to provide a warning to the operator. Settings are 106%, 98%, and 90% of initial, steady-state, operating power at 100% flow. The trip levels are automa tically varied with reactor coolant flow to protect against fuel damage at lower flows. The va riation is set to ensure that no fuel damage will occur at any indicated coolant flow. The ope rator may encounter any number (up to three) of trip points depending on the starting power of a given control rod withdraw al. The lower two points may be passed up (reset) by manual ope ration of a push button. The reset permissive is actuated (and indicated by a light) when the RBM reaches 4% power less than the trip point. The operator would then assess his loca l power and either reset or se lect a new rod. The highest (power) trip point may not be reset. The reload licensing an alysis was performed for a high trip level of 108% to accommodate the hysteresis effect of the trip setpoint at 106%. The corresponding low trip levels ar e 100% and 92% based on the assumed 8% trip level power
difference stated above. All trip levels are in terms of percen t of initial steady state operating power at 100% flow.
15.4.2.3.3 Results
At certain core exposures and power/flow conditions, this limiti ng transient may be a control RWE. Results reflect GE14 fuel introduction, some of which are dependent on fuel design and core loading pattern. Complian ce with the event acceptance criteria is demonstrated by cycle-dependent analysis of po tentially limiting events just prior to the operation of that cycle. The results are reported in the Supplementa l Reload Licensing Report (Reference 15.4-16).
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 15.4-7 15.4.2.3.4 Considerations of Uncertainties
The conservative assumptions whic h ensure that this event has been conservatively analyzed have been previously discussed in Section 15.4.2.3.2.
15.4.2.4 Barrier Performance
An evaluation of the barrier performance was not made for this ev ent since this is a localized event with very little change in the gross core character istics. Typically, an increase in total core power is less than 6% and the changes in pressure are negligible.
15.4.2.5 Radiological Consequences
An evaluation of the radiologi cal consequences is not requi red for this event since no radioactive material is released from the fuel.
15.4.3 CONTROL ROD MALOPERATION (SYSTEM MALFUNCTION OR OPERATOR ERROR)
This event is covered with the evaluation cited in Sections 15.4.1 and 15.4.2.
15.4.4 STARTUP OF IDLE RECIRCULATION PUMP
15.4.4.1 Identification of Causes and Frequency Classification
15.4.4.1.1 Identification of Causes
This action results directly from the operator' s manual action to initiate pump operation. It assumes that the remaining loop is already operating.
15.4.4.1.2 Frequenc y Classification
15.4.4.1.2.1 Normal Restart of Recirculation Pump at Power. This event is categorized as an incident of m oderate frequency.
15.4.4.1.2.2 Abnormal Startup of Idle Recirculation Pump. This event is categorized as an incident of mode rate frequency.
15.4.4.2 Sequence of Events and Systems Operation
15.4.4.2.1 Sequence of Events
Table 15.4-2 lists the sequence of events for Figure 15.4-1.
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 15.4-8 15.4.4.2.2 Systems Operation
This event assumes and takes credit for normal functioning of plant instrumentation and controls. No protection systems action is anticipated. No engineered safety feature (ESF) action occurs as a result of the event.
15.4.4.2.3 The Effect of Single Failures and Operator Errors
Attempts by the operator to start the pump at higher power levels will result in a reactor scram on flux.
15.4.4.3 Core and Sy stem Performance
15.4.4.3.1 Mathematical Model
The point-kinetics REDY model described in Section 15.0.3.3.1 is used to simulate this event.
15.4.4.3.2 Input Paramete rs and Initial Conditions
This analysis has been performed while the plan t is operating with a single recirculation loop, at 58% uprated power and 34% core flow. Conservatively, the water in the idle loop is
assumed to have a minimum te mperature of 100°F. The average enthalpy is based on saturated water temperature at the suction inlet with a linear enthalpy gradient to the discharge outlet water temperature of 100°F.
The active recirculation loop is operating with a pump speed that produces about 45% of normal rated jet pump diffuser flow in the active jet pumps. The inactive recirculation loop jet pumps are forward flowing at about 2% of normal jet pump diffuser flow because of natural circulation affects. The core is receiving about 34% of its normal rated flow.
The idle recirculation pump suc tion and discharge block valves are open. Normal procedure requires leaving an idle loop in this condition to maintain the loop temperature within the required limits for restart.
15.4.4.3.3 Results
The transient response to the incorrect startup of a cold id le recirculation loop is shown in Figure 15.4-1. Shortly after the pump begins to m ove, the flow from the started jet pump diffusers causes the core inlet flow to increase. The pump startup demand is conservatively assumed to ramp at a rate of 3.3% until ma ximum pump speed is ach ieved. The diffuser flows on the started side of the reactor increas e ultimately to about 1 44% of rated while the flow rate of the opposite loop di ffusers decreases and eventually reverses to about -8% of
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 15.4-9 rated. As the inactive loop pump increases speed the cold fluid is pumped out of the recirculation loop piping and is mixed with hot downcomer fluid and the mixture flows to the core with a resulting increase of the core inlet subcooling.
A moderate-duration neutron flux peak to just above 124% of NB rated (NBR = 3486 MWt) is produced as the colder, increasing core flow reduces the void volume. Surface heat flux follows the slower respons e of the fuel and peaks at 110% of rated before decreasing after the cold water is washed out of the loop at about 30 sec.
No damage occurs to the fuel barrier as the MCPR remains substantia lly above the safety limit.
15.4.4.3.4 Consideration of Uncertainties
This particular transient is analyzed for a maximum pump speed demand signal causing the ASD to adjust the recirculation pump speed upward at a nominal speed demand rate limit.
A conservative idle loop temper ature is assumed and no othe r uncertainties were included.
15.4.4.4 Barrier Performance
No evaluation of barrier performance is required for this even t since no significant pressure increases are incurred duri ng this transient. See Figure 15.4-1.
15.4.4.5 Radiological Consequences
Since this event does not result in any fuel failures or any release of primary coolant to either the secondary containment or to the environment, there are no radiological consequences associated with this event.
15.4.5 RECIRCULATION FLOW CONTROL FAILURE WITH INCREASING FLOW
15.4.5.1 Identification of Causes and Frequency Classification
15.4.5.1.1 Identification of Causes
An upscale failure of the master manual setpoint station can cau se an increase in the core coolant flow rate. Upscale failure of an indi vidual remote manual setp oint station or manual demand loop can also cause an increase in core coolant flow rate.
15.4.5.1.2 Frequenc y Classification
This event is an incident of moderate frequency.
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 15.4-10 15.4.5.2 Sequence of Events and Systems Operation
The increase in recirculation flow results in an increase in core flow. The increase in core flow causes an increase in core power level and shifts the powe r toward the top of the core by reducing the void fraction in the top of the core.
The rate and magnitude of the power increase are dependent on th e rate and magnitude of the flow increase. The operator woul d be expected to control a sl ow or small increase through normal operating procedures. Howe ver, a rapid or significant incr ease in neutron flux could exceed the high flux scram set point and initiate a scram.
This analysis assumes a relatively gradual flow increase that ch allenges the thermal limits but does not initiate plant protective systems prior to operator action to terminate the transient.
The turbine control (governor) valves and possi bly the bypass valves open to control reactor pressure. Core power increases until a steady state power level is reached at the maximum recirculation flow. The operator then regains control of the fl ow control system and returns the plant to a normal operating condition.
The analysis of this event assumes and takes credit for normal functioning of plant instrumentation and controls, and the reactor protection system (RPS). Operation of ESF is not expected.
15.4.5.2.1 The Effect of Single Failures and Operator Errors
The greatest challenge to the thermal limits is the gradual flow increase without actuation of the RPS. The transient is terminated by opera tor action but not until the maximum core flow of 108.5% rated flow is reached. No actions, either automatic or manual, occur to mitigate the transient prior to event termina tion at the maximum core flow.
15.4.5.3 Core and Sy stem Performance
15.4.5.3.1 Mathematical Model
The core is assumed to be in a pseudo steady state condition in which all plant thermal hydraulics are in equilibrium. The feedwater inlet temperatur e is assumed to be at its equilibrium value at all power levels during the event. The flow control line used to define the power/flow points represents the steepest attainable during normal reactor operation. The core radial and axial peaking distribu tions are assumed not to change during the event. The MCPR hot channel analysis along the flow ascension path is calculated with ISCOR (References 15.4-4 and 15.4-14).
Only potentially MCPR limiting fuel is evaluated. Potentiall y limiting is defined as within 0.10 of the core MCPR for the nominal rated power rodded depl etion (typically only fresh and C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-029 15.4-11 once burnt fuel). ISCOR is run at the power/flow level corresponding to the plant/cycle specific maximum flow and iterations are performed on the potentially MCPR limiting fuel channel radial peaking factor(s) such that the hot channel MCPR is equa l to the MCPR safety limit (SLMCPR) +/-0.005. ISCOR is then run along the specified flow control line at a range of core flows from 30% to 100% to obtain the potentially MCPR limiting fuel MCPR value(s) for each case. The results of this analysis determine the flow-dependent MCPR limits to assure that the MCPR will not fall below the SLMCPR for a flow incr ease event. These analyses bound final feedwater temperature re duction (FFWTR) as well as normal feedwater temperature conditions.
15.4.5.3.2 Input Paramete rs and Initial Conditions
The gradual increase in recircula tion flow provides the greatest challenge to th ermal limits. The final point in the transien t is a power below the high fl ux scram setpoint at maximum flow. Maximum flow is the maximum flow that can be attained by a credible controller failure initiated from a given low flow st arting point. A spectrum of initial, low flow starting points is analyzed at various points through out the cycle. The analysis assumes that the event is quasi-steady state and that a flow bi ased scram does not occur.
Table 15.4-3 contains a listing of the important input parameters and initial conditions.
15.4.5.3.3 Results
The reduced flow MCPR was calculated at disc rete flow points. The reduced flow MCPR operating limit curve is shown in the cycle specific Core Operating Lim its Report (COLR) for all cycle exposures, including FFWTR operation.
15.4.5.3.4 Considerati ons of Uncertainties
The conservative nature of the an alysis approach bounds the uncer tainties in void reactivity and power distribution characteristics e xpected for actual plant conditions.
15.4.5.4 Barrier Performance
The reduced flow MCPR is established so that the event does not challenge the safety limit MCPR. Therefore, no fuel damage is predicted as a result of this event.
15.4.5.5 Radiological Consequences
An evaluation of the radiologi cal consequences is not requi red for this event since no radioactive material is released.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-029 15.4-12 15.4.6 CHEMICAL AND VOLUME CONTROL SYSTEM MALFUNCTIONS
This event is not applicable to boiling water reacto r (BWR) plants.
15.4.7 MISPLACED BUNDLE ACCIDENT
The fuel loading error considers the conse quences of either of two possible events:
misorientation or mislocation of a fuel assembly. Further, the assumption is made that the error is not discove red during core verification. The purpose of the analysis is to determine the change in the minimum CPR and increase in LHGR between the correc tly loaded core and the misloaded core. A combination of the misorientation and mislocation is not considered because of the very low probability of occurrence.
15.4.7.1 Identification of Causes and Frequency Classification
15.4.7.1.1 Identification of Causes
The event discussed in this section is the impr oper loading of a fuel bundle and subsequent operation of the core. For the mislocation of a fuel bundle, three errors must occur during the initial core loading. First, a bundle must be misloaded into a wrong position in the core.
Second, the bundle that was suppos ed to be loaded where the mi slocation occurred would have to be placed in an incorrect location. Th ird, the misplaced bundles would have to be overlooked during the core verification performed following core loading. For the
misorientation of a fuel bundle, the bundle is loaded 180° from the correct orientation and this error is overlooked during th e core verification.
A fuel loading error would place a fuel assembly in an incorrect location in the core, potentially placing several highly reactive assemblies in close pr oximity. If a relatively high reactivity assembly is placed in a location not directly monitored by the LPRM/core monitoring system (i.e., unmonito red location), this incorrectly located assembly will operate at higher powers with reduced thermal margins relative to the symmetric monitored assembly.
The incorrectly located assembly may violate operating limits if the symmetric, monitored assembly is operated close to limits. If the incorrect location is a directly monitored cell, the change in the local power readings may not be sufficient to either aler t the operators of the loading error or to completely account for the re duction in thermal margin. In this situation, the incorrectly located assembly may violate operating limits while being treated by the monitoring system as if it were correctly loaded.
15.4.7.1.2 Frequency of Occurrence
This event is categorized as an infrequent incident but is analyzed for GNF reloads consistent with Section 15.3.3.1 as an incident of moderate frequency.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-029 15.4-13 15.4.7.2 Sequence of Events and Systems Operation
A fuel bundle is misloaded (incorr ect location or orientation) into the core and the error is not identified during the core verification process.
The core is operated th rough the cycle at the conditions assumed for the reference core (the specified core load) with the control rod sequence developed for the reference core. At some point during the cycle, the control rod sequence places the limiting asse mbly (being monitored by the core monitoring system as a correctly loaded assembly) at the MCPR operating limit. Potentia lly, this causes the misloaded assembly to be operated below the MCPR ope rating limit curve and above the design LHGR limit curve. Because the operator may be unable to detect the error, the core operation continues throughout the cycle.
Fuel loading errors, undetected by in-core in strumentation following fueling operations, may result in undetected reductions in thermal margins during power operations. No detection is assumed, and therefore, no corrective operator action or automatic protection system functioning occurs.
15.4.7.2.1 Effect of Single Fa ilure and Operator Errors
This analysis already represents the worst case (i.e., operation with a misplaced bundle requires multiple equipment fa ilures or operator errors).
15.4.7.3 Core and Sy stem Performance
15.4.7.3.1 Mathematical Model
A three-dimensional BWR simulato r model is used to calculate the core performance resulting from this event. This model is described in detail in Reference 15.4-4 and Sections S.2.2.1.8 and S.2.2.1.9 of Reference 15.4-5.
15.4.7.3.2 Input Paramete rs and Initial Conditions
Initial input parameters and conditions are cycle specific. Sections S.2.2.1.8 and S.2.2.1.9 of Reference 15.4-5 describe the input parameters and initial conditions ap plied to the mislocated and misoriented bundle events. For both the mislocated and misoriented bundle events, the fuel loading error is undet ected throughout the cycle by the core monitoring system.
15.4.7.3.3 Results
The results of the analyses for the fuel loading errors show that the resulting MCPR does not challenge the SLMCPR. No rods are expected to exceed the LHGR lim its. Results reflect GE14 fuel introduction, some of which are dependent on fuel de sign and core loading pattern.
Compliance with the event acceptance criteria is demonstrat ed by cycle-dependent analysis of C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-029 15.4-14 potentially limiting events just prior to the operation of that cycl
- e. The results are reported in the Supplemental Reload Li censing Report (Reference 15.4-16).
15.4.7.3.4 Considerati ons of Uncertainties
A sufficient number of mislocated assembly cases are considered to assure that the limiting case is evaluated. The mislocated assemblies are loaded into positions that could produce limiting results. For the misoriented assembly case, th e gap sizes resulting from the rotation are selected to assure a conservative estimate of the impact on MCPR.
15.4.7.4 Barrier Performance
An evaluation of the barrier performance was not made for this event since it is a mild and highly localized event. No perceptible change in the core pressure would be observed.
15.4.7.5 Radiological Consequences
An evaluation of the radiologi cal consequences is not requi red for this event since no radioactive material is released from the fuel.
15.4.8 SPECTRUM OF ROD EJECTION ASSEMBLIES
This event is not applicable to BWR plants.
15.4.9 CONTROL ROD DROP ACCIDENT
15.4.9.1 Identification of Causes and Frequency Classification
15.4.9.1.1 Identification of Causes
The control rod drop accident (CRDA) is the re sult of a high worth control rod decoupled from the drive mechanism, dropping out of the core. The subsequent insertion of positive reactivity causes a localized power excursion. This is not an anticipated event because of the system failures and personnel errors that would have to occur in combin ation to presen t the reactivity required at the same time that the coupling failed. The contro l rod patterns are controlled in accordance with the banked position withdraw sequence (BPWS) to preclude situations in which rod drops would have sufficient reactivity to cause the damage postulated by the power excursion.
Detailed discussions of the rod drop anal ysis and BPWS are given in Reference 15.4-3 and 15.4-7.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.4-15 15.4.9.1.2 Frequenc y Classification
The CRDA is categorized as a limiting fault beca use it is not expected to occur during the lifetime of the plant. However, postulated consequences include the potential for the release of radioactive material.
For reactivity anomalies, the CRDA is the limiting event.
15.4.9.2 Sequence of Events and Systems Operation
The CRDA assumptions include:
- 1. At some time, a fully inserted contro l rod becomes decoupled from its drive and sticks in the fully inserted position.
- 2. During the start up sequence, the rod patterns employed are permitted by the constraints on rod movement s by technical specificati ons and the rod sequence control hardware, including the maximum allowable number of bypassed rods. At some time under critical reactor c onditions, the rod pattern causes the decoupled rod to have the maximum worth from fully inserted to the position of its drive. The rod worth minimizer is not functioning. The rod drops at that time.
- 3. The reactor goes on a positive period, and the fuel temperature reactivity feedback terminates the initial power burst.
- 5. All withdrawn rods, except for the decoupled rod, scram at the technical specification rate.
- 6. The scram terminates the accident.
- 7. If the mechanical vacuum pump (MVP) is maintaining condenser vacuum (e.g., the plant was operating at 5% power or le ss) and the main st eam line radiation (MSLR) monitors detected radiation levels above the setpoint, the MSLR monitors would trip the MVP to reduce the fission product release from the condenser.
Although other normal plant instrumentation and controls are assumed to function, no credit for their operation is taken in the analysis of this event. No operator actions are required to terminate this event. Subsequent to reactor scram which terminates the event, normal vessel inventory makeup systems will be used as available.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.4-16 15.4.9.2.1 Effect of Single Failures and Operator Errors
As discussed, the event is te rminated, and therefore mitigated, by the APRM high flux scram signal to RPS. The RPS design meets the single fa ilure criteria. The even t is further mitigated by an initial control rod confi guration that complies with the BPWS. The withdrawal (or insertion) sequence is implemented by the operator and enforced by the RWM. An operator error in control rod movement will be detected and stopped by the RWM. If the RWM system is not operable, rod mo vement can only continue with a backup for the operator verifying compliance with the BPWS sequence. Failure of the RWM concurrent with an operator error of moving an out-of-sequence rod, contrary to procedures woul d be required to result in a potentially more limiting event. Therefore, sufficient redundancy exists such that termination of this transient w ithin the limiting criteria is assured.
At low power levels, the MVP tr ip maintains the condenser leak rates within the analytical assumptions. The MSLR monitor design meets the single failure criteria and no active failure would prevent the trip signal (Section 11.5.2.1).
15.4.9.3 Core and Sy stem Performance
15.4.9.3.1 Mathematical Model
The analytical methods, assumpti ons and conditions for evaluating the excursion aspects of the CRDA are described in detail in References 15.4-1 , 15.4-3 , 15.4-13. To limit the worth of the postulated dropped rod, the rod pattern control systems are pr ogrammed to follow the BPWS, which is generically defined in Reference 15.4-7.
15.4.9.3.2 Input Paramete rs and Initial Conditions
The data presented in Reference 15.4-7 shows that BPWS reduces the control rod worths to the degree that the detailed analys es presented in References 15.4-1 , 15.4-3 , 15.4-13 or the bounding analyses presented in Reference 15.4-11 do not need to be performed each cycle.
15.4.9.3.3 Results
Control rod drop accident results from BPWS plants have been statistically analyzed and documented in Reference 15.4-12. The results show that, in all cases, the peak fuel enthalpy in a CRDA would be much less than the 280 cal/gm design limit even with a maximum incremental rod worth corresponding to 95% pr obability at the 95% conf idence level. Based on these results, it was proposed to the NRC, and subsequently found accep table, to delete the CRDA from the standard GE BWR re load package for the BPWS plants.
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 15.4-17 The US Supplement to Reference 15.4-5 reports the results of radi ological analyses for initial cores that are orders of magn itude below those identified in 10 CFR 100. The radiological consequences of the CRDA, assuming a full core of more recent GE/GNF fuel designs, are discussed in Reference 15.4-5. With implementation of Alternative Source Term (AST), the radiological release acceptance criterion beco mes 10 CFR 50.67. An evaluation of fuel damage was performed because the maximum deposited fu el rod enthalpy exceeded 170 cal/gm, which is the enthalpy limit assume d for eventual claddi ng perforation. The number of fuel rods pred icted to fail (Reference 15.4-15) are bounded by the number assumed in the radiological consequences analysis for this event.
The total energy deposited and the associated increase in reactor system pressure during a rod drop event is not high relative to other events such as turbine trip without bypass or main steam line isolation valve closure, both of which are quantitatively analyzed. As such, the increase in reactor system pressure is not anticipated to result in penetration of the stress limits defined in Section III of the ASME boiler and pressure code.
15.4.9.4 Barrier Performance
An evaluation of the barrier performance was not made for this accident since this is a localized event with no significant change in the gross core temp erature or pressure.
15.4.9.5 Radiological Consequences
The radiological analysis is based on the AST described in Reference 15.4-8. Specific models, assumptions, and the program used for computer evaluation are described in Reference 15.4-6. Specific parametric values used in the evaluation are presented in Table 15.4-4. The radiological consequenc es remain bounding because cycle sp ecific analyses have confirmed that the number of fuel rods with an enthalpy greater than the threshold for fuel failure are well below the number assumed in the analysis.
15.4.9.5.1 Fission Product Release from Fuel
The failure of 1.8% of the core was assumed for this analysis. The mass fraction of the fuel in the damaged rods which reaches or exceeds the initiation temperatur e of fuel melting (taken as 2804°C) is assumed to be 0.0077.
Fuel reaching melt condition is assumed to re lease 100% of the noble gas inventory and 50%
of the iodine inventory. The remaining fuel rods with clad damage only (no melting), will undergo a gap release which is assumed to release 10% of nobl e gases and 10% of iodine inventories.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-11-000 15.4-18 The core inventory of fission products is based on a plant-specific ORIGEN 2 run for pre-power uprate basis of 3489 MW with 1000 days of exposure, adjusted as follows:
A scale factor of 1.0528 to bound the dose impact of power level to 3556 MWt, A correction to increase by 25 percent short-lived krypton values (based on comparisons to other core inventory tables), and An increase by 60% in the activity of longer lived isotopes to bound longer plant operation at a higher burnup rate.
The assumed core power of 3556 MWt is the lic ensed power increased by 2% to account for power measurement uncertainties. These adjustments resulted in a conservative source term (in terms of activity available). A peaking factor of 1.7 is assumed and no delay time is considered between departure from that power condition and the initiation of the accident.
15.4.9.5.2 Fission Product Transport to the Environment
The transport pathway consists of a release from the core to the coolant, carryover with steam to the turbine condenser and l eakage from the condenser to the environment. The release fractions are given in Reference 15.4-6 and are consistent with Reference 15.4-8. No credit is taken for mixing in the turbine building or filtration by the CREF.
Of the activity reaching the condenser, 100% of the noble gases, and 10%
of the iodine (due to partitioning and plate-out) and 1% of the particulates rema in airborne and are available for release to the environment. The activity airborne in the condens er leaks to the environment as a ground level release at a rate of 1% of condenser volume per day. Release from the condenser is assumed to termin ate 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the onset of the accident. Radioactive decay is accounted for during residence in the condenser; however, it is neglected after release to the environment. If the condenser is not is olated from the offgas system, the activity is processed through the offgas system. In this case, radioactive decay is accounted for during the residence in the offg as system. Response of the offgas system to el evated radiation levels is described in Section 11.3.2.4.5.
The activity airborne in the condenser is presented in Table 15.4-5. The cumulative release of activity to the environment is presented in Table 15.4-6.
15.4.9.5.3 Results
The calculated exposures from the de sign basis analysis are presented in Table 15.4-7 and are within the limits of 10 CFR 50.67.
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 15.4-19 15.4.10 REFERENCES
15.4-1 Stirn, R. C., "Rod Drop Accident Analysis for Large BWRs," NEDO-10527, March 1976.
15.4-2 GE Nuclear Energy, "WNP-2 Power Uprate Transient Analysis Task Report,"
GE-NE-208-08-0393, Septem ber 1993 (Proprietary).
15.4-3 Stirn, R. C., "Rod Drop Accident Analysis for Large BWRs," Supplement 1, NEDO-10527, July 1972.
15.4-4 "Steady -State Nuclear Met hods," NEDE-30130-P-A, April 1985.
15.4-5 "General Electric Standard Appli cation for Reactor Fuel," NEDE-24011-P-A and "Supplement for United States,"
NEDE-24011-P-A-US (most recent approved version referenced in COLR).
15.4-6 Energy Northwest, "C olumbia Generating Station Alternative Source Term,"
CGS-FTS-0168, Revision 0, August 2007.
15.4-7 Paone, C. J., "Bank Position Withdrawal Sequence," NEDO-21231.
15.4-8 NRC Regulatory Guide 1.183, "Alternativ e Source Term for Evaluating Design Basis Accidents at Nuclear Po wer Reactors," July 2000.
15.4-9 Deleted.
15.4-10 General Electric, "Safety Evaluation for Eliminating the Boiling Water Reactor Main Steam Isolation Valve Closure F unction and Scram Func tion of the Main Steam Line Radiation Monitor," NEDO-31400A, Cla ss I, October 1992.
15.4-11 "GE BWR Generic Reload Applicati on for 8x8 Fuel," Supplement 3 to Revision 1, NEDO-20360.
15.4-12 Letter from R. E. Engel (GE) to D.
M. Vassallo (NRC), "Elimination of Control Rod Drop Accident Analysis for Banked Position Withdrawal Sequence Plants," February 24, 1982.
15.4-13 Stirn, R. C., "Rod Drop Accident Anal ysis for Large BWRs," Supplement 2, NEDO-10527, Ja nuary 1973.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-029 15.4-20 15.4-14 Letter from D. G. Eisenhut (NRC) to R. L. Gridley (GE), "Safety Evaluation for the General Electric Topical Report, Generic Reload Fuel Application (NEDE-24011-P)," May 12, 1978, MFN-212-78.
15.4-15 "General Electric Fuel Bundle De signs," NEDE-31152P, Revision 8, April 2001.
15.4-16 Supplemental Reload Licensing Report for Columbia (mos t recent version referenced in COLR).
C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 15.4-21 Table 15.4-1 Sequence of Events - Rod Wi t hdra w al Error in Power Range
Time (sec) a Event 0 Core is assumed to be o p erating at rated conditions.
0 Operator selects and withdraws the maximum w o rth cont r o l rod. 1 The total core power and the local po wer in the vicinity of the control rod increase. 5 The LPRM system indicates excessive localized peaking. 5 The operator ignores warni ng and continues withdrawal. 15 The RBM system indicates excessive localized peaking. 15 The operator ignores warni ng and continues withdrawal. 20 The RBM system initiates a rod bl ock inhibiting further withdrawal. 40 Reactor core stabilizes at higher core power level.
60 Operator re-inserts control r od to reduce core power level. 80 Core stabilizes at rated conditions.
a Approximately.
C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 15.4-22 Table 15.4-2 Sequence of Events for an Abnormal Startup of an Idle Recirculation Loop
Time (sec)
Event 0.00 Plant operating with one recirculation loop only. 5.00 Start idle recirculation loop pump motor. 29.6 Peak value of core inlet subcooling. 29.6 Peak thermal power. Estimated APRM thermal power approximately 1% below APRM thermal power setpoint. 45.7 Pump motor at full speed.
80+ Reactor reaches new e q uilibrium condition.
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 15.4-23 Table 15.4-3
Reactor Recirculation Pump Flow Increase Input Parameters and Initial Conditions Parameter Value Reactor Power/Core Flow Flow increase is initiated from several power/flow points and terminates at 111% rated power, 108.5%rated flow Power Distribution The MCPR equals th e safety limit at th e final power/flow condition Reactivity The results are applicable from Beginning of Cycle to End of Cycle for nominal and reduced feedwater temperatures Control Rod Configuration The control rod patte rn is the same at the initial and final points.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 15.4-24 Table 15.4-4
Control Rod Drop Accide nt Evaluation Parameters I. Data and assumptions used to estimate radioactive source from postulated accidents.
A. Power level Section 15.4.9.5.1 B. Burnup Section 15.4.9.5.1 C. Fuel damaged 1.8% of core D. Release of activity by nuclide Table 15.4-6 E. Iodine fractions (1) Organic 0.0015 (2) Elemental 0.0485 (3) Particulate 0.95 F. Reactor coolant activity before the accident. N/A II. Data and assumptions used to estimate activity released. A. Condenser leak rate (%/day) 1.0 B. Turbine building leak rate (%/day) N/A C. Valve closure time (sec) N/A D. Adsorption and filtration efficiencies (1) Organic iodine N/A (2) Elemental iodine N/A (3) Particulate iodine N/A (4) Particulate fission products N/A E. Recirculation system parameters (1) Flow rate N/A (2) Mixing efficiency N/A (3) Filter efficiency N/A F. Containment spray parameters (flow rate, drop size, etc.) N/A C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 LDCN-03-003 15.4-25 Table 15.4-4
Control Rod Drop Accident Ev aluation Paramete rs (Continued)
G. Containment volumes N/A H. All other pertinent da ta and assumptions None III. Dispersion data Table 15.0-4 IV. Dose data A. Method of dose calculation Reference 15.4-6 B. Dose conversion assumptions Reference 15.4-6 C. Peak activity concentrations in condenser Table 15.4-5 D. Doses Table 15.4-7
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 15.4-26 Table 15.4-5 Control Rod Drop Accident Activity Airborne in the Condenser (Curies) 3556 MWth 1 m 30 m 2 hrs 8 hrs 12 hrs 24 hrs Kr83m 4.12E+04 3.43E+04 1.96E+04 2.06E+03 4.61E+02 5.13E+00 Kr85m 8.50E+04 7.88E+04 6.21E+04 2.40E+04 1.27E+04 1.90E+03 Kr85 4.77E+03 4.77E+03 4.76E+03 4.75E+03 4.74E+03 4.72E+03 Kr87 1.54E+05 1.18E+05 5.20E+04 1.95E+03 2.18E+02 3.05E-01 Kr88 2.19E+05 1.95E+05 1.34E+05 3.03E+04 1.12E+04 5.72E+02 Kr89 2.05E+05 3.71E+02 1.14E-06 6.96E-18 1.01E-16 7.70E-18 Xe131m 3.24E+03 3.23E+03 3.22E+03 3.16E+03 3.13E+03 3.02E+03 Xe133m 1.93E+04 1.91E+04 1.88E+04 1.74E+04 1.65E+04 1.41E+04 Xe133 6.30E+05 6.28E+05 6.23E+05 6.01E+05 5.87E+05 5.47E+05 Xe135m 1.23E+05 3.40E+04 6.24E+02 7.12E-05 1.68E-09 2.52E-16 Xe135 1.52E+05 1.46E+05 1.31E+05 8.30E+04 6.13E+04 2.47E+04 Xe137 4.52E+05 2.62E+03 2.99E-04 4.13E-18 3.26E-17 2.41E-17 Xe138 4.00E+05 1.22E+05 3.11E+03 1.30E-03 7.24E-08 2.47E-18 Total noble gases 2.49E+06 1.39E+06 1.05E+06 7.68E+05 6.97E+05 5.96E+05 I-131* 3.06E+05 2.64E+05 1.83E+05 8.50E+04 6.15E+04 2.47E+04 I-132 6.71E+05 1.98E+05 1.34E+05 3.03E+04 1.12E+04 5.72E+02 I-133 6.05E+05 1.22E+05 3.11E+03 1.30E-03 7.24E-08 1.02E-17 I-134 3.24E+03 3.23E+03 3.22E+03 3.16E+03 3.13E+03 3.02E+03 I-135 2.51E+06 1.41E+06 1.07E+06 7.85E+05 7.14E+05 6.10E+05 Total Iodine 4.09E+06 1.99E+06 1.39E+06 9.03E+05 7.90E+05 6.38E+05
- The isotopic iodine activity is the sum of the elemental and organic iodines with 97% elemental and 3% organic.
The particulate iodine comprise 0%.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 15.4-27 Table 15.4-5 Control Rod Drop Accident Activity Airborne in the Conde nser (Curies) (Continued) 3556 MWth 1 m 30 m 2 hrs 8 hrs 12 hrs 24 hrs Rb86 5.87E-02 5.86E-02 5.84E-02 5.77E-02 5.73E-02 5.60E-02 Cs134 8.23E+00 8.23E+00 8.22E+00 8.20E+00 8.18E+00 8.14E+00 Cs136 1.82E+00 1.82E+00 1.81E+00 1.79E+00 1.77E+00 1.71E+00 Cs137 6.63E+00 6.63E+00 6.62E+00 6.61E+00 6.59E+00 6.56E+00 Sb127 1.38E-02 1.38E-02 1.36E-02 1.30E-02 1.26E-02 1.14E-02 Sb129 3.95E-02 3.66E-02 2.88E-02 1.10E-02 5.84E-03 8.61E-04 Te127m 1.95E-03 1.95E-03 1.94E-03 1.94E-03 1.93E-03 1.91E-03 Te127 1.38E-02 1.38E-02 1.38E-02 1.36E-02 1.34E-02 1.25E-02 Te129m 5.81E-03 5.80E-03 5.79E-03 5.75E-03 5.72E-03 5.63E-03 Te129 3.72E-02 3.74E-02 3.43E-02 1.52E-02 8.10E-03 1.20E-03 Te131m 1.75E-02 1.73E-02 1.67E-02 1.45E-02 1.32E-02 9.97E-03 Te132 1.67E-01 1.66E-01 1.64E-01 1.55E-01 1.49E-01 1.33E-01 Ba137m 1.54E+00 6.62E+00 6.62E+00 6.61E+00 6.59E+00 6.56E+00 Ba139 7.82E-02 6.14E-02 2.90E-02 1.43E-03 1.94E-04 4.75E-07 Ba140 7.65E-02 7.64E-02 7.61E-02 7.49E-02 7.41E-02 7.17E-02 Mo99 1.02E-02 1.02E-02 1.00E-02 9.39E-03 8.99E-03 7.90E-03 Tc99m 9.06E-03 9.12E-03 9.27E-03 9.45E-03 9.33E-03 8.55E-03 Ru103 9.81E-03 9.81E-03 9.79E-03 9.72E-03 9.68E-03 9.55E-03 Ru105 7.21E-03 6.69E-03 5.33E-03 2.14E-03 1.16E-03 1.87E-04 Ru106 4.26E-03 4.26E-03 4.26E-03 4.24E-03 4.23E-03 4.21E-03 Rh105 6.83E-03 6.83E-03 6.80E-03 6.42E-03 6.04E-03 4.87E-03 Y90 3.50E-05 6.38E-05 1.52E-04 4.90E-04 7.02E-04 1.28E-03 Y91 4.56E-04 4.66E-04 4.95E-04 5.82E-04 6.22E-04 6.87E-04 Y92 6.39E-04 4.76E-03 1.26E-02 1.28E-02 7.81E-03 1.10E-03 Y93 5.94E-04 5.74E-04 5.18E-04 3.42E-04 2.59E-04 1.13E-04 Zr95 7.13E-04 7.13E-04 7.12E-04 7.08E-04 7.06E-04 6.98E-04 Zr97 7.23E-04 7.09E-04 6.66E-04 5.21E-04 4.42E-04 2.70E-04 C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 15.4-28 Table 15.4-5 Control Rod Drop Accident Activity Airborne in the Conde nser (Curies) (Continued) 3556 MWth 1 m 30 m 2 hrs 8 hrs 12 hrs 24 hrs Nb95 7.13E-04 7.13E-04 7.13E-04 7.11E-04 7.10E-04 7.06E-04 La140 8.08E-04 1.43E-03 3.33E-03 1.04E-02 1.46E-02 2.54E-02 La141 7.26E-04 6.66E-04 5.10E-04 1.75E-04 8.58E-05 1.01E-05 La142 6.91E-04 5.55E-04 2.81E-04 1.84E-05 3.00E-06 1.29E-08 Pr143 6.31E-04 6.32E-04 6.35E-04 6.44E-04 6.49E-04 6.58E-04 Nd147 2.86E-04 2.85E-04 2.84E-04 2.79E-04 2.76E-04 2.66E-04 Am241 1.28E-07 1.28E-07 1.28E-07 1.28E-07 1.27E-07 1.27E-07 Cm242 2.91E-05 2.91E-05 2.90E-05 2.89E-05 2.89E-05 2.87E-05 Cm244 2.36E-06 2.35E-06 2.35E-06 2.35E-06 2.34E-06 2.33E-06 Ce141 1.85E-03 1.85E-03 1.85E-03 1.83E-03 1.82E-03 1.80E-03 Ce143 1.67E-03 1.66E-03 1.60E-03 1.40E-03 1.28E-03 9.85E-04 Ce144 1.36E-03 1.36E-03 1.36E-03 1.35E-03 1.35E-03 1.34E-03 Np239 2.93E-02 2.91E-02 2.85E-02 2.64E-02 2.51E-02 2.15E-02 Pu238 3.99E-06 3.99E-06 3.99E-06 3.98E-06 3.97E-06 3.95E-06 Pu239 7.89E-07 7.89E-07 7.89E-07 7.87E-07 7.85E-07 7.82E-07 Pu240 1.30E-06 1.30E-06 1.30E-06 1.29E-06 1.29E-06 1.29E-06 Pu241 3.70E-04 3.70E-04 3.69E-04 3.68E-04 3.68E-04 3.66E-04 Sr89 3.37E-02 3.37E-02 3.37E-02 3.35E-02 3.33E-02 3.29E-02 Sr90 5.58E-03 5.58E-03 5.57E-03 5.56E-03 5.55E-03 5.52E-03 Sr91 4.32E-02 4.17E-02 3.74E-02 2.42E-02 1.81E-02 7.54E-03 Sr92 5.01E-02 4.41E-02 2.97E-02 6.14E-03 2.15E-03 9.15E-05 C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 15.4-29 Table 15.4-6 Control Rod Drop Accident Activity Airborne to the Environment (Curies) 3556 MWth 1 m 0.5 hr 2 hrs 8 hrs 12 hrs 24 hrs Kr83m 2.78E-01 7.86E+00 2.43E+01 4.38E+01 4.56E+01 4.61E+01 Kr85m 5.74E-01 1.71E+01 6.09E+01 1.61E+02 1.91E+02 2.19E+02 Kr85 3.21E-02 9.92E-01 3.97E+00 1.59E+01 2.38E+01 4.75E+01 Kr87 1.04E+00 2.83E+01 7.87E+01 1.17E+02 1.18E+02 1.18E+02 Kr88 1.48E+00 4.31E+01 1.45E+02 3.20E+02 3.52E+02 3.69E+02 Kr89 1.54E+00 8.07E+00 8.08E+00 8.08E+00 8.08E+00 8.08E+00 Xe131m 2.18E-02 6.73E-01 2.69E+00 1.07E+01 1.59E+01 3.13E+01 Xe133m 1.30E-01 4.00E+00 1.58E+01 6.10E+01 8.92E+01 1.66E+02 Xe133 4.24E+00 1.31E+02 5.22E+02 2.05E+03 3.04E+03 5.88E+03 Xe135m 8.48E-01 1.48E+01 2.00E+01 2.01E+01 2.01E+01 2.01E+01 Xe135 1.02E+00 3.10E+01 1.18E+02 3.80E+02 5.00E+02 7.01E+02 Xe137 3.32E+00 2.09E+01 2.10E+01 2.10E+01 2.10E+01 2.10E+01 Xe138 2.75E+00 5.00E+01 7.03E+01 7.08E+01 7.08E+01 7.08E+01 Total Noble 1.73E+01 3.58E+02 1.09E+03 3.28E+03 4.50E+03 7.70E+03 Gases I-131* 4.08E-01 1.19E+01 4.01E+01 1.05E+02 1.35E+02 2.12E+02 I-132 4.81E+00 1.48E+02 5.83E+02 2.21E+03 3.23E+03 6.10E+03 I-133 8.80E-01 1.58E+01 2.40E+01 3.60E+01 4.39E+01 6.76E+01 I-134 2.06E+00 5.93E+01 1.97E+02 4.97E+02 6.18E+02 8.19E+02 I-135 4.80E+00 6.40E+01 1.66E+02 3.41E+02 3.73E+02 3.90E+02 Total Iodine 1.30E+01 2.99E+02 1.01E+03 3.19E+03 4.40E+03 7.59E+03
- The isotopic iodine activity is the sum of the elemental and organic iodines with 97% elemental and 3% organic.
The particulate iodine comprise 0%.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 15.4-30 Table 15.4-6 Control Rod Drop Accident Activity Airborne to the Environment (Curies) (Continued) 3556 MWth 1 m 0.5 hr 2 hrs 8 hrs 12 hrs 24 hrs Rb86 3.95E-07 1.22E-05 4.88E-05 1.94E-04 2.90E-04 5.73E-04 Cs134 5.54E-05 1.71E-03 6.86E-03 2.74E-02 4.11E-02 8.19E-02 Cs136 1.23E-05 3.79E-04 1.52E-03 6.02E-03 8.99E-03 1.77E-02 Cs137 4.47E-05 1.38E-03 5.52E-03 2.21E-02 3.31E-02 6.60E-02 Sb127 9.31E-08 2.87E-06 1.14E-05 4.47E-05 6.60E-05 1.26E-04 Sb129 2.66E-07 7.92E-06 2.83E-05 7.46E-05 8.82E-05 1.01E-04 Te127m 1.31E-08 4.05E-07 1.62E-06 6.48E-06 9.70E-06 1.93E-05 Te127 9.31E-08 2.88E-06 1.15E-05 4.58E-05 6.83E-05 1.33E-04 Te129m 3.91E-08 1.21E-06 4.83E-06 1.93E-05 2.88E-05 5.72E-05 Te129 2.50E-07 7.77E-06 3.04E-05 9.08E-05 1.10E-04 1.28E-04 Te131m 1.18E-07 3.63E-06 1.43E-05 5.33E-05 7.65E-05 1.34E-04 Te132 1.12E-06 3.46E-05 1.38E-04 5.35E-04 7.88E-04 1.49E-03 Ba137m 5.43E-06 1.21E-03 5.35E-03 2.19E-02 3.29E-02 6.58E-02 Ba139 5.29E-07 1.45E-05 4.15E-05 6.44E-05 6.55E-05 6.56E-05 Ba140 5.15E-07 1.59E-05 6.36E-05 2.53E-04 3.77E-04 7.41E-04 Mo99 6.89E-08 2.12E-06 8.44E-06 3.27E-05 4.80E-05 9.02E-05 Tc99m 6.11E-08 1.89E-06 7.65E-06 3.12E-05 4.69E-05 9.18E-05 Ru103 6.61E-08 2.04E-06 8.17E-06 3.26E-05 4.88E-05 9.69E-05 Ru105 4.86E-08 1.45E-06 5.19E-06 1.39E-05 1.66E-05 1.93E-05 Ru106 2.87E-08 8.87E-07 3.55E-06 1.42E-05 2.13E-05 4.24E-05 Rh105 4.60E-08 1.42E-06 5.68E-06 2.23E-05 3.27E-05 6.00E-05 Y90 2.33E-10 1.02E-08 7.78E-08 8.86E-07 1.88E-06 6.88E-06 Y91 3.07E-09 9.60E-08 3.97E-07 1.75E-06 2.76E-06 6.06E-06 Y92 3.78E-09 5.63E-07 6.33E-06 4.33E-05 6.04E-05 7.81E-05 Y93 4.00E-09 1.22E-07 4.63E-07 1.52E-06 2.02E-06 2.90E-06 C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 15.4-31 Table 15.4-6 Control Rod Drop Accident Activity Airborne to the Environment (Curies) (Continued) 3556 MWth 1 m 0.5 hr 2 hrs 8 hrs 12 hrs 24 hrs Zr95 4.81E-09 1.48E-07 5.94E-07 2.37E-06 3.55E-06 7.06E-06 Zr97 4.87E-09 1.49E-07 5.79E-07 2.06E-06 2.86E-06 4.60E-06 Nb95 4.81E-09 1.48E-07 5.94E-07 2.38E-06 3.56E-06 7.10E-06 La140 5.37E-09 2.31E-07 1.72E-06 1.91E-05 4.00E-05 1.41E-04 La141 4.90E-09 1.45E-07 5.11E-07 1.29E-06 1.50E-06 1.68E-06 La142 4.67E-09 1.30E-07 3.81E-07 6.23E-07 6.37E-07 6.40E-07 Pr143 4.25E-09 1.32E-07 5.28E-07 2.13E-06 3.21E-06 6.48E-06 Nd147 1.92E-09 5.94E-08 2.37E-07 9.42E-07 1.40E-06 2.76E-06 Am241 8.63E-13 2.67E-11 1.07E-10 4.27E-10 6.39E-10 1.28E-09 Cm242 1.96E-10 6.05E-09 2.42E-08 9.67E-08 1.45E-07 2.89E-07 Cm244 1.59E-11 4.90E-10 1.96E-09 7.84E-09 1.18E-08 2.34E-08 Ce141 1.25E-08 3.85E-07 1.54E-06 6.15E-06 9.20E-06 1.83E-05 Ce143 1.13E-08 3.47E-07 1.37E-06 5.12E-06 7.36E-06 1.30E-05 Ce144 9.15E-09 2.83E-07 1.13E-06 4.52E-06 6.77E-06 1.35E-05 Np239 1.97E-07 6.07E-06 2.41E-05 9.28E-05 1.36E-04 2.52E-04 Pu238 2.69E-11 8.31E-10 3.33E-09 1.33E-08 1.99E-08 3.98E-08 Pu239 5.32E-12 1.64E-10 6.58E-10 2.63E-09 3.94E-09 7.86E-09 Pu240 8.75E-12 2.70E-10 1.08E-09 4.33E-09 6.48E-09 1.29E-08 Pu241 2.49E-09 7.69E-08 3.08E-07 1.23E-06 1.85E-06 3.68E-06 Sr89 2.27E-07 7.02E-06 2.81E-05 1.12E-04 1.68E-04 3.34E-04 Sr90 3.76E-08 1.16E-06 4.65E-06 1.86E-05 2.78E-05 5.55E-05 Sr91 2.91E-07 8.84E-06 3.36E-05 1.09E-04 1.44E-04 2.05E-04 Sr92 3.38E-07 9.81E-06 3.26E-05 7.00E-05 7.63E-05 7.96E-05 C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 15.4-32 Table 15.4-7
Control Rod Drop Accident Radiological Effects (rem)
Area Time TEDE Dose Exclusion area (1950 m) 2 hr 0.03 Low population zone (4827 m) 30 days 0.03 Control Room 30 days 0.7 Amendment 53 November 1998 Columbia Generating Station Final Safety Analysis Report Abnormal Startup of Id le Recirculation Loop at 57.9% Uprated Power, 34.1% Flow 020002.03 15.4-1.1 Amendment 53 November 1998 Columbia Generating Station Final Safety Analysis Report Abnormal Startup of Idle Recirculation Loop at 57.9% Uprated Power, 34.1% Flow 020002.04 15.4-1.2 Amendment 53 November 1998 Columbia Generating Station Final Safety Analysis Report Abnormal Startup of an Idle Recirculation Loop at 57.9% Uprated Power, 34.1% Flow 020002.05 15.4-1.3 Amendment 53 November 1998 Columbia Generating Station Final Safety Analysis Report Abnormal Startup of Idle Recirculation Loop at 57.9% Uprated Power, 34.1% Flow 020002.06 15.4-1.4 Amendment 53 November 1998 Columbia Generating Station Final Safety Analysis Report Abnormal Startup of Idle Recirculation Loop at 57.9% Uprated Power, 34.1% Flow 020002.07 15.4-1.5 Columbia Generating StationFinal SafetyAnalysis ReportDraw. No.Rev.Figure Amendment 59December 2007 Form No. 960690FH LDCN-05-009, 06-000Leakage Path Model for Rod DropAccident 960222.71 15.4-2 Core Coolant Condenser TURBINE BUILDING EnvironmentNoc redit wastakenfor mixingor holdupin theTurbine Building. Therefore the TB was not included in the model.Des ignBasis Model C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 15.5-1 15.5 INCREASE IN REACTOR COOLANT INVENTORY
15.5.1 INADVERTENT HIGH-PRE SSURE CORE SPRAY STARTUP
This transient is classified as a nonlimiting event for both original and uprated power conditions. The transient is not analyzed for each reload, but was analyzed for the GE14 New Fuel Introduction. Inadvertent startup of the high-pressure core spray (HPCS) system was analyzed because it provides the greatest auxiliary source of cold water into the vessel.
15.5.1.1 Identification of Causes and Frequency Classification 15.5.1.1.1 Identification of Causes
Manual startup (i.e., operator erro r) and continued injection of the HPCS system is postulated for this analysis.
15.5.1.1.2 Frequenc y Classification
This transient disturbance is categorized as an incident of moderate frequency.
15.5.1.2 Sequence of Events and Systems Operation
The HPCS system is manually initiated and injects into the reactor vessel, reaching full flow in approximately one second. The addition of the cooler wate r to the upper plenum causes a reduction in steam flow. This ca uses some reactor pressure d ecrease as the turbine control system responds to the event.
As the steam flow decreases, the feedwater system responds by decreasing flow. In less than a minute, the reac tor and the auxiliary st eam systems stabilize at a new, lower power level. The analysis assumes normal functioning of plant instrumentation and controls. No engineered safety feature (E SF) function is expected in response to this transient. Plant paramete r responses are shown in Figure 15.5-1.
15.5.1.2.1 The Effect of Single Failures and Operator Errors
Inadvertent operation of the HPCS system results in a mild depressurization. Level control
and pressure regulator actuation are expected to establish a new stable operating state. The effect of a single failure in the DEH control system will have no effect on the transient because of its redundant design.
The effect of a single failure in the level control system has rather straightforward consequences including level rise or fall by improper control of the feedwater system.
Increasing level will trip the tu rbine and automatically stop inj ection by the HPCS system.
Decreasing level will automatically initiate a scram at the L3 level trip.
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 15.5-2 15.5.1.3 Core and Sy stem Performance
15.5.1.3.1 Mathematical Model
The one-dimensional ODYN transient analysis model described in Reference 15.5-1 was used to simulate this transient.
15.5.1.3.2 Input Parameter and Initial Conditions The important parameters are shown in Table 15.5-1. 15.5.1.3.3 Results
The calculated uncorrected CPR for the simulated bundles is less than 0.01 (Reference 15.5-2). A summary of transient key peak values is found in Table 15.0-1.
15.5.1.3.3.1 Consider ation of Uncertainties. Important analytical factors including reactivity coefficient and feedwater temperat ure change have been assumed to be at the worst conditions so that any deviations in th e actual plant parameters will produce a less severe transient.
15.5.1.4 Barrier Performance
Figure 15.5-1 indicates only a slight pressure reduction from in itial conditions; therefore, reactor coolant pressure boundary pressure margins are not impacted.
15.5.1.5 Radiological Consequences
Since this event does not result in any fuel failures or any release of primary coolant to either the secondary containment or to the environment, there are no radiological consequences associated with this event.
15.5.2 CHEMICAL VOLUME CONTROL SYSTEM MALFUNCTION (OR OPERATOR ERROR)
This event is not applicable to boiling water reactor plants.
15.5.3 BOILING WATER REACTOR TRANS IENTS WHICH INCREASE REACTOR COOLANT INVENTORY
These events are discussed in Sections 15.1 and 15.2.
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 15.5-3 15.
5.4 REFERENCES
15.5-1 "Qualification of the One-Dimensional Core Transi ent Model for Boiling Water Reactors," Volumes 1, 2, 3 and 4, NEDC-24154-P-A, February 2000.
15.5-2 GE Hitachi Nucear Energy, "Inadver tent High Pressure Core Spray Startup Analysis," 0000-0098-5369-R0, March 2009.
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 15.5-5 Table 15.5-1
Input Parameters and Initial Conditions HPCS Injection
Parameter Value Reactor power 102% (3556 MWth)
Core flow 106% ICF and 88% ELLLA HPCS water source Suppression pool HPCS source pressure 14.7 psia HPCS source temperature 40° F HPCS source enthalpy 11.0 Btu/lbm HPCS pump flow 12.6 % of rated feedwater flow (3800 gal/min)
Vessel-to-suppression pool differential pressure 1020 psid
Rev.Figure Draw. No.Form No. 960690 LDCN-08-035 Amendment 60 December 2009 Columbia Generating Station Final Safety Analysis Report 020002.18 15.5-1 Inadvertent Start of High Pressure Core Spray Pump at 102% Uprated Power, 88% Flow 0.0 25.0 50.0 75.0 100.0 125.0 150.00.05.010.015.020.025.030.035.0 Time (sec)% Rated Level - Inch above Sep Skirt Vessel Steam Flow Turbine Steam Flow Feedwater Flow HPCS Flow (% Rated FW) 0.0 50.0 100.0 150.0 200.0-25.0 0.0 25.0 50.0 75.0 100.0-0.5-0.3 0.0 0.3 0.50.05.010.015.020.025.030.035.00.05.010.015.020.025.030.035.00.05.010.015.020.025.030.035.0 Time (sec)
Time (sec)
Time (sec)% Ra% Rated Reactivity Components ($)ted Neutron Flux Average Surface Heat Flux Core Inlet Flow Core Inlet Subcooling Vessel Press Rise (psi)
Safety Valve Flow Relief Valve Flow Bypass Valve Flow Void Reactivity Doppler Reactivity Scram Reactivity Total Reactivity C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-06-043 15.6-1 15.6 DECREASE IN REACTOR COOLANT INVENTORY
15.6.1 INADVERTENT SAFETY/RELIEF VALVE OPENING
This event is discussed in Section 15.1.4.
15.6.2 INSTRUMENT LINE PIPE BREAK
This faulted condition is not a limiting event for either original or uprated power conditions. Therefore, the uprated power analysis for the accident has not been updated.
This event involves the postulated pipe break in a small steam or liquid line that is connected to the reactor coolant pressure boundary (RCPB) and is located in the reactor building. If the break were inside primary containment, the event would be bou nded by the steam line break inside containment (LOCA) (see Section 15.6.5). That event is bounding because the instrument line sizes are bounded by the spectrum of breaks considered in the LOCA analysis.
15.6.2.1 Identification of Causes and Frequency Classification
15.6.2.1.1 Identification of Causes
There is no specific event or circumstance identified whic h results in the failure of an instrument line. These lines are designed to specific engineering specifi cations and standards, and seismic and environmental requirements. However, for the purpose of evaluating the consequences of this event, the failure of an instrument line is assumed to occur.
15.6.2.1.2 Frequenc y Classification
This event is categorized as a limiting fault.
15.6.2.2 Sequence of Events and Systems Operation
The instrument line ruptures (com plete circumferential break) and releases reactor coolant into the secondary containment. The analysis assume s that the reactor cool ant activity is at the Technical Specification limit corresponding to an iodine spike of 4µCi/g dose equivalent 131 I and that the break cannot be isol ated. The operators have a vari ety of methods to detect the leak such as monitoring plant area temperatures and radiation levels, system pressures, or sump inventories, or during operator rounds. It is assumed that the re actor operators identify the break after 20 minutes and initiate a reactor scram.
Using available plant systems, the operators maintain reactor water le vel and cool down and depressurize the reactor within 5 hr, at a rate less than or equal to the 100°F/hr limit in the Technical Specifications. Examples of plant sy stems or components the operators can use for C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 15.6-2 inventory and temperature contro ls include HPCS, RCIC, SRVs, RHR, or the condensate feed system. No credit is taken for the au tomatic initiation of the RPS or ESF.
15.6.2.2.1 The Effect of Single Failures and Operator Errors
The event is handled by operator actions. Assuming additional single equipment failure or single operator error occurrences, adequate equipment would be available to respond to the loss of reactor coolant.
15.6.2.3 Core and Sy stem Performance The inventory loss is within the capacity of the make up systems available and the shutdown and the cool down are controlled evolutions. Therefore, no fuel damage will occur and no other barriers are challenged.
15.6.2.3.1 Qualitative Summary - Results
Since instrument line break s result in a slower rate of cool ant loss and are bounded, the results are qualitative rather than quan titative. Since the rate of coolant loss is slow, an orderly reactor system depressurization follows reactor scram and the primary system is cooled down and maintained without ECCS actua tion. No fuel damage or co re uncovery occurs as a result of this event.
15.6.2.4 Barrier Performance
15.6.2.4.1 General
The release of primary coolant through the orificed instrument line would not result in an increase in secondary containment pressure.
15.6.2.5 Radiological Consequences
The radiological consequen ces are based on the following assumptions and methods:
- a. The broken instrumentation line contains a 0.5-in. diameter flow restricting orifice inside the drywell.
- b. Flow is critical at the orifice and is determined using the GOTHIC computer program (Reference 15.6-1) that employs the Henry model for subcooled liquid and the Moody model for saturated and superheated vapors (Reference 15.6-4). c. The total integrated mass of fluid released by means of the break during the blowdown is 121,000 lb. Of this total, 29,800 lb flash to steam.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-06-043 15.6-3 d. The specific models, assumptions and the program used for the radiological analysis is in Reference 15.6-3. Specific values of parameters used in the evaluation are presented in Table 15.6-1. The leakage path used in these calculations is shown in Figure 15.6-1. e. The activity released from the instrument line break is based on the iodine spike concentration of 4µCi
/g dose equivalent 131I and is assumed to not mix or be held up within the secondary containment a nd is released to the environment at a flow rate of 80,000 cfm without SGT filtration.
- f. The activity rele ased to the secondary contai nment and the environment is presented in Table 15.6-2.
15.6.2.5.1 Results
The calculated exposur es are presented in Table 15.6-3.
15.6.3 STEAM GENERATOR TUBE FAILURE
This event is not applicable to boiling water reactor (BWR) plants.
15.6.4 STEAM SYSTEM PIPING BREAK OUTSIDE CONTAINMENT
This event involves the postulation of a larg e steam line pipe break outside the primary containment. The analysis assumes that a main stea m line instantaneously and circumferentially breaks at a location downstream of the outboard isolation valve. The plant is designed to immediately detect such an occurrence, initiate isolation of the broken line, and actuate the necessary protective features. The main steam line was selected for analysis because the postulated event envelops evaluation of steam line failures outside containment.
15.6.4.1 Identification of Causes and Frequency Classification
15.6.4.1.1 Identification of Causes
A main steam line break is postulated without the cause being identifie
- d. These lines are designed to specific engineering codes and standards, and seismi c and environmental requirements. However, for the purpose of evaluating the conse quences of a postulated large steam line rupture, the failure of a ma in steam line is as sumed to occur.
15.6.4.1.2 Frequenc y Classification
This event is categorized as a limiting fault.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 15.6-4 15.6.4.2 Sequence of Events and Systems Operation
15.6.4.2.1 Sequence of Events
When the steam line breaks, steam flow through the failed line will rapidly increase, initiating the high main steam line flow trip that initiates the signal to close the main steam isolation valves (MSIVs). The closing MSIVs initiate a reactor protection (RPS) signal, scramming the reactor. The analysis assumes that the MSIVs are fully closed 6 seconds after the break.
Performance of the engineered sa fety feature (ESF) systems in response to the loss of coolant is discussed in Chapter 6.
The sequence of events and approximate time required to reach the event is given in Table 15.6-4.
15.6.4.2.2 Systems Operation
A postulated guillotine break of one of the four main steam lines outside the containment results in mass loss from both ends of the break. The flow from the upst ream side is initially limited by the flow restrictor upstream of the inboard isolation valve. Flow from the downstream side is initially limited by the total area of the flow restrictors in the three unbroken lines. Initially, only steam will issue from the broken end of the steam line. The flow in each line is limited by cr itical flow at the limiter to a maximum of 200% of rated flow for each line. Rapid depressurization of the RPV causes the water level to rise resulting in a steam-water mixture flowing from the break until the valves are closed. Mass loss (steam and water) is reduced and finally terminated (except for leakage) as the MSIVs close.
15.6.4.2.3 The Effect of Single Failures and Operator Errors
The effect of single failures has been considered in analyzing this event. All of the protective sequences for this event are capable of single equipment failure or single operator error accommodation and yet still complete the necessary safety action.
15.6.4.3 Core and Sy stem Performance The temperature and pressure transients resulting as a conseq uence of this accident are insufficient to cause fuel damage.
15.6.4.3.1 Input Paramete rs and Initial Conditions
See Section 6.3 for initial conditions.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 15.6-5 15.6.4.3.2 Results
There is no fuel damage as a consequence of this accident.
See Section 6.3 for ECCS analysis.
15.6.4.3.3 Considerations of Uncertainties
Discussions of the uncertainties associated with the ECCS pe rformance and the containment isolation systems are di scussed in Sections 6.3 and 7.3, respectively.
15.6.4.4 Barrier Performance
Since this break occurs outside the primary containment, barrier performance within the containment envelope is not applicable. There are sufficient vent openings in the steam tunnel to ensure that the secondary containment structure will not be damaged.
15.6.4.5 Radiological Consequences
The radiological analysis is based on NRC Regulatory Guide 1.183 (Reference 15.6-5). The dispersion of the plume is based on the pu ff model given in Regulatory Guide 1.194 (Reference 15.6-6).
The specific models, assumptions, and the program used for computer evaluation are described in Reference 15.6-3. Specific values of parameters used in the evaluation are presented in Table 15.6-5. There is no fuel damage as a result of this accident. The only activity available for release from the break is that which is pres ent in the reactor coolant and steam lines prior to the break. The iodine inventories and the subsequent exposur es are based on the equilibrium conditions and maximu m reactor coolant activity for an iodine spiking event as allowed by the Technical Specifications. The analys is assumes all the activity in this discharge becomes airborne and released directly and unfiltered to the e nvironment. The release of activity to the environment is presented in Table 15.6-6.
The following assumptions and conditions are used in determining the mass loss from the primary system from the inception of the break to full closure of the MSIVs:
- a. The reactor is operating at the power level associated with maximum mass release,
- b. Nuclear system pressure is 1060 psia and remains consta nt during closure,
- c. An instantaneous circumferential break of the main steam line occurs, C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 15.6-6 d. Isolation valves start to close at 0.5 sec on high flow signal and are fully closed at 6 sec,
- e. The Moody critical fl ow model (Reference 15.6-4) is applicable, and
- f. The flow limiters allow up to 200% of rated flow through the MSIVs, and
- g. Level rise time is conservatively assume d to be one second.
Mixture quality is conservatively taken to be a constant 7% (steam weight percentage) during mixture flow.
The total integrated mass leav ing the RPV through the steam lin e break is 130,000 lb of which 105,000 lb is liquid and 25,000 lb is steam. Only the liquid por tion of the discharged coolant is assumed to carry the iodine activity of 4
µCi/gm dose-equivalent of I-131. The entire amount of activity in the liquid is assume d to be released to the environment.
The transport pathway is a direct unfiltered release to the environment and an unfiltered entrance to the control room as presented in Figure 15.6-2.
15.6.4.5.1 Results
The calculated doses for the design basis analysis are presented in Table 15.6-7. The doses are within the limits of 10 CFR 50.67.
15.6.5 LOSS-OF-COOLANT ACCIDENTS (RESULTING FROM SPECTRUM OF POSTULATED PIPING BREAKS WITHIN THE REACTOR COOLANT PRESSURE BOUNDARY) - INSIDE CONTAINMENT
Accidents that could result in the release of radioactive fission produc ts directly into the containment are the results of postulated break s in the reactor coolant system pressure boundary (RCPB). The analysis postulates that the most severe pressurization tr ansient to the primary containment is caused by a complete circumferential break of the suction line of one of the two recirculation loops. Flow through the break transports the reactor vessel contents to the suppression pool.
The loss-of-coolant accident (LOCA) postulates the break of a ny of the spectrum of piping systems that form the RCPB. The plant and operator responses to the spectrum of breaks are presented in Chapter 6. Chapter 6 demonstrates that fuel, co re, and barrier performance requirements are met for the spect rum of breaks. The bounding ra diological analysis for the LOCA event detailed in this section reflects an inadequate core cooling accident that degrades to complete core damage. The event assumed for the analysis is the break inside containment of one of the main steam lines. The radi ological analysis assu mptions presented in C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 15.6-7 Section 15.6.5 are separate and distinct and are not mechanistically tied to the pipe break analyzed in Chapter 6.
15.6.5.1 Identification of Causes and Frequency Classification
15.6.5.1.1 Identification of Causes
There are no realistic, identifia ble events that would result in a pipe break inside the containment of the magnitude required to cause the fuel damage sufficient to release the source terms assumed in this section. The piping is designed to specific engineering codes and standards and for severe seis mic and environmental conditions. However, since such an accident provides an upper limit estimate of th e dose consequences for the limiting faulted condition, it is evaluated without the causes bein g identified.
15.6.5.1.2 Frequenc y Classification
This event is categorized as a limiting fault.
15.6.5.2 Sequence of Events and Systems Operation
The sequence of events and syst em operations ar e discussed in Chapter 6. The effect of single failures and operator errors is discussed in Chapter 6.
15.6.5.3 Core and Sy stem Performance
For the plant response to the LOCA and the eval uation of the system an d core performance, see Chapter 6.
15.6.5.4 Radiological Consequences
The radiological conse quences are based on the guidance provided in Regulatory Guide 1.183 (Reference 15.6-5) for the purpose of determining ade quacy of the plant design to meet 10 CFR 50.67 limits.
A schematic of the transport pathway is shown in Figure 15.6-3.
15.6.5.4.1 Design Basis Analysis
The specific models, assumptions, and computer code used to evaluate the radiological
consequences of the bounding LOCA based on the above criteria are presented in Reference 15.6-3. Specific values of parameters used in this evaluation are presented in Table 15.6-8.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 15.6-8 15.6.5.4.1.1 Fission Pr oduct Release from Fuel. It is assumed that 100% of the noble gases and 30% of the iodine are released from an equilibrium core operati ng at a power level of 3556 MWt for 1000 days prior to the accident. Of this release, 100% of the noble gases become airborne. Some of the iodine is removed by plate-out and filtrati on; therefore, it is not available for airborne release to the environment. The activity airborne in the containment is presented in Table 15.6-9.
15.6.5.4.1.2 Fission Product Tr ansport to the Environment. The fission product transport to the environment consists of tw o basic pathways. One transpor t pathway consists of leakage from the containment to the secondary containment by several differen t mechanisms and is discharged to the environment through the SGT system at an elevated location. Of the secondary containment flow, 50 cfm bypasses the SGT filters. The second transport pathway consists of leakage from the containment directly to the environment through piping systems that originate in containment and terminate outside the reactor build ing. The individual mechanisms for leakage from the primary containment are:
- a. Containment leakage -
Leakage from primary cont ainment to the secondary containment. Prior to the completion of secondary containment drawdown this leakage is assumed to be a direct release to the environment; however, when drawdown is complete, 20 minutes post accident, this leakage is treated by the SGT system before it is released to the environment. No credit is taken for forced mixing and holdup within the secondary containment.
- b. ESF leakage - Leakage from engin eered safety feature (ESF) components outside the primary containment (all ES F equipment which circulates primary coolant or suppression pool water during the course of the postulated accident) to the secondary containment. This leak age is treated in a manner similar to the containment leakage described above.
- d. MSIV leakage - Leakage from the primary containment (or the RPV) through the main steam isolation valves (MSIVs) to the turbine building. Iodine is assumed to plate-out in th ree of the four main steam lines. Plate-out in the broken fourth main steam line is not cred ited. In that line it is assumed one MSIV fails open (single failure), wherea s all MSIVs in the other three steam lines are assumed to close.
- e. Bypass leakage -
Leakage from the primary co ntainment that bypasses the secondary containment and is released, unt reated, directly to the environment.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-09-007 15.6-9 Fission product activities in secondary containment and those released to the environment based on the above assu mptions are given in Tables 15.6-10 and 15.6-11, respectively.
15.6.5.4.1.3 Suppression Pool pH Control. The suppression pool pH is maintained above 7.0 for the duration of the accident as a result of the standby liquid contro l system injection of sodium pentaborate solution. The solution is assumed to be injected a nd fully mixed with the suppression pool water with in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> post-LOCA.
15.6.5.4.1.4 Results. The calc ulated doses for the design basis analysis are presented in Table 15.6-12 and are within the limts of 10 CFR 50.67.
15.6.6 FEEDWATER LINE BREAK - OUTSIDE CONTAINMENT
In order to evaluate large liquid process line pipe breaks outside containment, the failure of a feedwater line is assumed to evaluate the response of the plant de sign to this postulated event.
The postulated break of the feedwater line, re presenting the largest liquid line outside the containment, provides the envelope evaluation relative to this type of occurrence. The break is assumed to be instantaneous, circ umferential, and outboard of the outermost isolation valve.
The analysis has not been update d for the change in MSIV isola tion setpoint from Level 2 to Level 1 because the analysis remains bounded by the reci rculation line break LOCA (Reference 15.6-9).
15.6.6.1 Identification of Causes and Frequency Classification
15.6.6.1.1 Identification of Causes
A feedwater line break is assumed without the cau se being identified.
The subject piping is designed to specific engineer ing codes and standards.
15.6.6.1.2 Frequenc y Classification
This event is categorized as a limiting fault.
15.6.6.2 Sequence of Events and Systems Operation 15.6.6.2.1 Sequence of Events
The sequence of events is shown in Table 15.6-13.
15.6.6.2.2 Systems Operation
It is assumed that the normally operating plant instrument and controls are functioning. Credit is taken for the actuation of the reactor isolation system and ECCS. The RPS (SRVs, ECCS, C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.6-10 and control rod drives) and plant protection system (RHR heat exchanger) are assumed to function. The ESF system is assumed to ope rate normally. Although not an ECCS and not credited nor required for mitigation of this event, RCIC will al so be used if available for maintaining vessel level as it initiates at approximate ly the same low reac tor vessel level as HPCS.
15.6.6.2.3 The Effect of Single Failures and Operator Errors
The feedwater line outside the containment is a special case of the general LOCA break spectrum considered in Section 6.3. The general single-failur e analysis for LOCAs is discussed in Section 6.3.3.3. For the feedwater line break out side the containment, since the break is isolable, the HPCS can provide adequate flow to the vessel to maintain core cooling and prevent fuel rod cladding failure. A single failure of the HPCS w ould require actuation of ADS and the low-pressure core cooling system s to keep the core covered with water.
15.6.6.3 Core and Sy stem Performance
15.6.6.3.1 Qualitative Summary
The accident evaluation qualitatively considered in this section is consider ed to be conservative and to envelope assessment of the consequences of the postu lated failure of one of the feedwater piping lines external to the containment. The accident is postulated to occur at the input parameters and in itial conditions given in Table 6.3-2.
15.6.6.3.2 Qualitative Results
The feedwater line break outside the containmen t is less limiting than the steam line break outside the containment or the LOCA inside the containment.
The reactor vessel is isolated on low-low water level and the HP CS would restore the reactor water level to the normal elevati on. The fuel is c overed throughout the event and there are no pressure or temperature transients sufficient to cause fuel damage.
15.6.6.3.3 Consideration of Uncertainties
This event was conservatively analyzed and uncertainties were adequately considered (see Section 6.3 for details).
15.6.6.4 Barrier Performance
A break spectrum analysis for the complete range of reactor conditions indicates that the limiting fault event for breaks outside the containment is a complete severance of one of the C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.6-11 main steam lines. The feedwater system piping br eak is less severe than the main steam line break.
15.6.6.5 Radiological Consequences
The specific models, assumptions, and the program used for computer evaluation are described in Reference 15.6-2. Specific values of parameters used in the evaluation are presented in Table 15.6-14. A diagram of the leakage path for this accident is shown in Figure 15.6-4. 15.6.6.5.1 Fission Product Release
Fission product release is assumed to occur from two pathways: activity being pumped from the condenser hotwell and activity returning to the feedwater system fr om the reactor water cleanup (RWCU) system. The activity in both of these sources is based on the Technical Specification coolant limit.
Noble gas activity in the condens ate is negligible si nce the air ej ectors remove most of the noble gas from the condenser.
15.6.6.5.2 Fission Product Transport to the Environment
The transport pathway consists of liquid release from the break, carryover to the turbine building atmosphere due to flas hing and partitioning and unfiltered release to the environment through the turbine build ing ventilation system.
Of the 860,000 lb of condensate released from the break, 86,000 lb flashes to steam with assumed iodine carryover of 100%.
Of the activity remaining in the unflashed liquid, 5% is assumed to become airborne.
Normally, all feedwater reaching the break location will have passed through condensate demine ralizers which have a 90% i odine removal efficiency.
However, as a result of the increased feedwater flow caused by the break, differential pressure across the demineralizers is assumed to initiate flow through the demineralizer bypass line. This bypass line then carries 15%
of the total flow resulting in an effective iodine removal efficiency for all flow of 76.5%. In additi on, it is also assumed that 2771 lb of liquid returning from the RWCU are released prior to isolation of the RWCU. The activity concentration in this return steam is 1% of the RPV coolant concentration.
Taking no credit for holdup, decay, or plate-out dur ing transport through the turbine building, the release of activity to the environment is presented in Table 15.6-15. The release is assumed to take place within 2 hr of the occurrence of the break.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-09-007 15.6-12 15.6.6.5.3 Results
The calculated exposures for the re alistic analysis are presented in Table 15.6-16 and are a small fraction of 10 CFR 100 guidelines.
15.
6.7 REFERENCES
15.6-1 GOTHIC Containment Analysis Pack age, Technical Manual, Version 4.0, Numerical Applications, In c., NA18907-06, Revision 3.
15.6-2 Nguyen, D., "Realistic Accident Analysis for Gene ral Electric Boiling Water Reactor - The RELAC Code and User's Guide," (NEDO-21142).
15.6-3 Energy Northwest, "C olumbia Generating Station Alternative Source Term,"
CGS-FTS-0168, Revision 0, August 2007.
15.6-4 Moody, F. J, "Maximum Two-Phase Vessel Blowdown from Pipes," ASME Paper Number 65-WA/HT-1, March 15, 1965.
15.6-5 Regulatory Guide 1.183 , Revision 0, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July, 2000.
15.6-6 Regulatory Guide 1.194, Revision 0, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," July, 2003.
15.6-7 Federal Guidance Report 11, "Limiti ng Values of Radionuc lide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion," Oak Ridge National Laboratory, 1988.
15.6-8 Federal Guidance Report 12, "External Exposure to Radionuclides in Air, Water and Soil," Oak Ridge National Laboratory, 1993.
15.6-9 GE Hitachi Nuclear Energy, "Li cense Amendment Request for Proposed Changes to Columbia Technical Spec ifications: Changing Group 1 Isolation Valves' Low Reactor Water Level Isolation Signal from the Current Level 2 to Level 1," 0000-0081-6730-R1, July 2008.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-06-043 15.6-13 Table 15.6-1
Instrument Line Break Accident - Parameters Tabulated for Postulated Accident Analyses Parameters Design Basis Assumptions I. Data and assumptions used to estimate radioactive source from postulated accidents A. Power level N/A B. Burnup N/A C. Fuel damaged None D. Airborne activity by nuclide Table 15.6-2 E. Iodine fractions (1) Organic 0.15% (2) Elemental 4.85% (3) Particulate 95% F. Initial reactor coolant activity with iodine spike 4 µci/g 131 I 1.6 µci/g 132 I 14.8 µci/g 133 I 11.0 µci/g 134 I 30.0 µci/g 135 I 16.0 µci/g II. Data and assumptions used to estimate activity released A. Primary containment leak rate (%/day) N/A B. Secondary containment effluent rate (cfm) 80,000 a C. Valve movement times N/A D. Adsorption and filtr ation efficiencies (1) Organic iodine N/A (2) Elemental iodine N/A (3) Particulate iodine N/A (4) Particulate fission products N/A E. Recirculation system parameters (1) Flow rate N/A (2) Mixing efficiency N/A (3) Filter efficiency N/A F. Containment spray parameters (flow rate, drop size, etc.) N/A G. Containment volumes N/A H. All other pertinent da ta and assumptions None C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-06-043 15.6-14 Table 15.6-1
Instrument Line Break Accident - Parameters Tabulated for Postulated A ccident Analyses (Continued)
Parameters Design Basis Assumptions III. Dispersion data A. Offsite See Table 15.0-4 B. Control Room See Table 15.0-5 IV. Dose data A. Method of dose calculation Reference 15.6-5 B. Dose conversion assumptions Reference 15.6-7 C. Peak activity released from secondary containment Table 15.6-2 D. Doses Table 15.6-3 a No forced mixing in secondary containment is considered.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-06-043 15.6-15 Table 15.6-2
Instrument Line Failure
Activity Airborne in Secondary Containment (Ci) 2 hr 5 hr 8 hr 1 day 30 days 133 Xe 7.02E-07 2.60E-07 0.00E+00 0.00E+00 0.00E+00 135 Xe 1.22E-05 3.67E-06 0.00E+00 0.00E+00 0.00E+00 Total 1.29E-05 3.93E-06 0.00E+00 0.00E+00 0.00E+00 131 I 2.00E-02 8.09E-03 0.00E+00 0.00E+00 0.00E+00 132 I 1.03E-01 1.72E-02 0.00E+00 0.00E+00 0.00E+00 133 I 1.29E-01 4.80E-02 0.00E+00 0.00E+00 0.00E+00 134 I 7.60E-02 2.80E-03 0.00E+00 0.00E+00 0.00E+00 135 I 1.64E-01 4.93E-02 0.00E+00 0.00E+00 0.00E+00 Total 4.92E-01 1.25E-01 0.00E+00 0.00E+00 0.00E+00 Activity Airborne in the Environment (Ci) 2 hr 5 hr 8 hr 1 day 30 days 133 Xe 1.90E-03 3.93E-03 4.91E-03 6.06E-03 6.08E-03 135 Xe 3.56E-02 6.76E-02 8.02E-02 9.06E-02 9.06E-02 Total 3.75E-02 7.15E-02 8.51E-02 9.67E-02 9.67E-02 131 I 4.87E+01 8.46E+01 8.72E+01 8.72E+01 8.72E+01 132 I 3.51E+02 4.81E+02 4.86E+02 4.86E+02 4.86E+02 133 I 3.26E+02 5.51E+02 5.65E+02 5.65E+02 5.65E+02 134 I 4.91E+02 5.52E+02 5.53E+02 5.53E+02 5.53E+02 135 I 4.46E+02 7.06E+02 7.22E+02 7.22E+02 7.22E+02 Total 1.66E+03 2.37E+03 2.41E+03 2.41E+03 2.41E+03
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-06-043 15.6-16 Table 15.6-3
Instrument Line Failure Radiological Effects
Area Total Effective Dose Equivalent (rem) Limit (rem TEDE)
Control Room 1.58 5 Exclusion area (1950 m) (2 hr) 0.36 2.5 Low population zone (4827 m) (30 days) 0.16 2.5 C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 LDCN-03-003 15.6-17 Table 15.6-4
Sequence of Events for Steam Line Break Outside Containment
Time Event 0 Guillotine break of one main steam line outside primary containment.
0.5 a High steam line flow signal initiates closure of MSIV. <1.0 Reactor begins to scram.
>6.0 Main steam line isola tion valves fully closed. 10 Safety/relief valves open on high ve ssel pressure. The valves open and close to maintain vessel pressu re at approxima tely 1100 psi. 600 Operator initiates ADS or manually controls relief valves. Vessel depressurizes rapidly. 750 High-pressure core spray initiates on low water level. 1270 Core effectively refloode
- d. No fuel rod failure.
a Approximately.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 15.6-18 Table 15.6-5
Steam Line Break Accident - Parameters Tabulated for Postulated Accident Analyses
Parameters Design Basis Assumptions I. Data and assumptions used to estimate radioactive source from postulated accidents.
A. Power level N/A B. Burnup N/A C. Fuel damaged None D. Release of activity by nuclide Table 15.6-6 E. Iodine fractions 1 (1) Organic N/A (2) Elemental N/A (3) Particulate N/A F. Reactor coolant activity before the accident corresponds to the iodine spike of 4
µci/gm dose-equivalent I-131 4 µci/gm II. Data and assumptions used to estimate activity released. A. Primary containment leak rate (%/day) N/A B. Secondary containmen t leak rate (%/day)
N/A C. Isolation valve closure time (sec) 6 D. Adsorption and filtration efficiencies (1) Organic iodine 1 N/A (2) Elemental iodine N/A (3) Particulate iodine N/A (4) Particulate fission products N/A 1 Since no filtration is credited, speci ation of iodines is not applicable.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 15.6-19 Table 15.6-5
Steam Line Break Accident - Parameters Tabulated for Postulated A ccident Analyses (Continued)
Parameters Design Basis Assumptions E. Recirculation system parameters N/A (1) Flow rate N/A (2) Mixing efficiency N/A (3) Filter efficiency N/A F. Containment spray parameters (flow rate, drop size, etc.) N/A G. Containment volumes N/A H. All other pertinent da ta and assumptions None III. Dispersion data (1) Offsite Table 15.0-4 (2) Control Room 8.19 E-4 sec/m 3 IV. Dose data A. Method of dose calculation Reference 15.6-3 B. Dose conversion assumptions Reference 15.6-7 C. Peak activity concentrations in containment N/A D. Doses Table 15.6-7 C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 15.6-20 Table 15.6-6
Steam Line Break Accident Activity Release to Environment (Curies)
Isotope Activity Released 131 I 1.91E 02 C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 15.6-21 Table 15.6-7
Steam Line Break Accident Radiological Effects of a Puff Release
Exclusion area (1950 m) 0.40 Low population zone (4827 m) 0.11 Control Room 1.8 C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 15.6-22 Table 15.6-8
Loss-of-Coolant Acci dent - Parameters
Tabulated for Postulated Accident Analysis Parameters Design Basis Assumptions I. Data and assumptions used to estimate radioactive source from postulated accidents A. Power level 3556 B. Burnup N/A C. Fuel damaged 100% D. Airborne activity by nuclide Table 15.6-10 and 15.6-11 E. Iodine fractions (1) Organic 0.0015 (2) Elemental 0.0485 (3) Particulate 0.95 F. Reactor coolant activity before the accident N/A II. Data and assumptions used to estimate activity released A. Primary containment leak rate includes MSIV leakage (% volume/day) 0 - 24 hrs 0.5 24 - 720 hrs 0.25 B. Secondary containment leak rate (%/day) N/A C. Drawdown period (minutes) 20 D. Adsorption and filtration efficiencies (%) (1) Organic iodine 98% (2) Elemental iodine 98% (3) Particulate iodine 98% (4) Particulate fission products 98%
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 15.6-23 Table 15.6-8
Loss-of-Coolant Acci dent - Parameters
Tabulated for Postulated A ccident Analysis (Continued)
Parameters Design Basis Assumptions E. Secondary containment volumetric flow rate bypassing SGT filters 1 , cfm 50 F. Secondary containment bypass leakage 0 - 24 hrs 0.04% volume per day 24 - 720 hrs 0.02%
volume per day G. Recirculation system parameters (1) Flow rate (cfm) N/A (2) Mixing efficiency N/A (3) Filter efficiency N/A H. Containment spray removal rates Time (hr) Removal Rate (1/hr) 0 0.0 0.25 6.20 2.44 0.62 24.0 0.0 I. Containment volumes Table 6.2-1 J. MSIV leak rate per steam line 0 - 24 hrs 16 scfh 24 - 720 hrs 8 scfh K. ESF leakage into secondary containment 2 gpm L. CREF bypass leakage 50 cfm III. Dispersion data (1) Offsite Table 15.0-4 (2) Control room Table 15.0-5
1 SGT filter bypass will reduce SGT filt er efficiency from 99% to 98%.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 15.6-24 Table 15.6-8
Loss-of-Coolant Acci dent - Parameters
Tabulated for Postulated A ccident Analysis (Continued)
Parameters Design Basis Assumptions IV. Dose data A. Method of dose calculation Reference 15.6-5 B. Dose conversion assumptions Reference 15.6-7 , 15.6-8 C. Peak activity concentrations in containment Table 15.6-9 D. Doses Table 15.6-12
C OLUMBIA G ENERATING S TATION Amendment59 F INAL S AFETY A NALYSIS R EPORTDecember 2007LDCN-05-009 15.6-25 Table 15.6-9 Loss-of-Coolant Accident Primary Containment 1 Activity (Curies)
Isotope 0.25 hr 0.5 hr 0.8 hr 1.0 hr 2.0 hr 4.0 hr 8.0 hr 24.0 hr 30 day 131 I 2.15E+06 1.72E+06 2.42E+06 2.63E+06 2.69E+06 3.93E+05 1.25E+05 4.09E+04 3.03E+03 132 I 2.82E+06 2.10E+06 2.75E+06 2.79E+06 2.15E+06 2.17E+05 3.57E+04 1.12E+02 1.03E-01 133 I 4.16E+06 3.30E+06 4.62E+06 4.97E+06 4.95E+06 6.82E+05 1.92E+05 3.92E+04 3.25E-06 134 I 3.81E+06 2.49E+06 2.87E+06 2.55E+06 1.18E+06 3.47E+04 4.52E+02 4.13E-04 0 135 I 3.78E+06 2.94E+06 4.05E+06 4.29E+06 3.97E+06 4.77E+05 1.02E+05 6.80E+03 0 Total iodines 1.67E+07 1.26E+07 1.67E+07 1.72E+07 1.49E+07 1.80E+06 4.55E+05 8.71E+04 3.03E+03 83m Kr 2.51E+05 4.92E+05 1.80E+06 3.02E+06 5.87E+06 2.84E+06 6.34E+05 1.58E+03 0 85m Kr 5.45E+05 1.13E+06 4.35E+06 7.71E+06 1.86E+07 1.39E+07 7.36E+06 5.84E+05 0 85 Kr 3.17E+04 6.82E+04 2.74E+05 5.05E+05 1.43E+06 1.46E+06 1.46E+06 1.45E+06 1.29E+06 87 Kr 9.02E+05 1.69E+06 5.92E+06 9.53E+06 1.56E+07 5.33E+06 5.97E+05 9.35E+01 0 88 Kr 1.38E+06 2.79E+06 1.05E+07 1.82E+07 4.03E+07 2.51E+07 9.29E+06 1.76E+05 0 89 Kr 6.47E+04 5.31E+03 8.12E+02 5.71E+01 3.42E-04 1.60E-15 3.36E-38 5.95E-129 0 131m Xe 2.15E+04 4.63E+04 1.86E+05 3.42E+05 9.66E+05 9.81E+05 9.71E+05 9.30E+05 1.56E+05 133m Xe 1.28E+05 2.74E+05 1.10E+06 2.01E+06 5.63E+06 5.60E+06 5.33E+06 4.33E+06 6.17E+02 133 Xe 4.19E+06 9.00E+06 3.61E+07 6.64E+07 1.87E+08 1.89E+08 1.86E+08 1.72E+08 3.52E+06 135m Xe 4.40E+05 4.87E+05 1.00E+06 9.51E+05 1.88E+05 9.29E+02 2.18E-02 0 0 135 Xe 1.02E+06 2.23E+06 8.57E+06 1.56E+07 4.20E+07 4.21E+07 3.82E+07 1.79E+07 0 137 Xe 2.50E+05 3.75E+04 1.05E+04 1.35E+03 8.99E-02 5.09E-11 1.59E-29 1.52E-103 0 138 Xe 1.50E+06 1.75E+06 3.81E+06 3.82E+06 9.34E+05 7.13E+03 3.98E-01 3.87E-18 0 Total noble gases 1.07E+07 2.00E+07 7.36E+07 1.28E+08 3.19E+08 2.86E+08 2.50E+08 1.97E+08 4.97E+06 Alkali metals 9.84E+05 7.85E+05 9.21E+05 9.62E+05 9.72E+05 1.29E+05 3.05E+04 9.57E+01 7.73E+01 Te-group 2.83E+05 2.31E+05 2.04E+06 2.49E+06 2.41E+06 3.09E+05 6.56E+04 1.72E+02 5.34E+01 Noble metals 0.00E+00 0.00E+00 1.55E+05 1.97E+05 1.99E+05 2.57E+04 5.82E+03 1.60E+01 3.99E+00 La-group 0.00E+00 0.00E+00 2.40E+04 3.19E+04 3.17E+04 9.64E+03 3.11E+03 1.33E+01 1.00E+01 Ce-group 0.00E+00 0.00E+00 1.15E+05 1.46E+05 1.52E+05 1.98E+04 4.49E+03 1.18E+01 1.05E+00 ______________________________
1Primary Containment includes the Dry Well & the Wet Well Air-Space
C OLUMBIA G ENERATING S TATION Amendment59 F INAL S AFETY A NALYSIS R EPORTDecember 2007LDCN-05-009 15.6-26 Table 15.6-10 Loss-of-Coolant Accident Secondary Containment Activity (Curies) - 20 Minute Drawdown Isotope 0.25 hr 0.5 hr 0.75 hr 1 hr 2 hr 4 hr 8 hr 24 hr 30 day 131 I 0 6.89E+00 9.90E+00 1.15E+01 1.50E+01 7.11E+00 6.08E+00 5.47E+00 4.11E-01 132 I 0 8.38E+00 1.13E+01 1.22E+01 1.19E+01 3.26E+00 8.84E-01 6.86E-03 1.79E-07 133 I 0 1.32E+01 1.89E+01 2.17E+01 2.75E+01 1.23E+01 9.36E+00 5.24E+00 0 134 I 0 9.98E+00 1.18E+01 1.12E+01 6.54E+00 6.29E-01 2.20E-02 5.53E-08 0 135 I 0 1.18E+01 1.65E+01 1.88E+01 2.21E+01 8.61E+00 4.95E+00 9.08E-01 0 Total iodines 0 5.02E+01 6.84E+01 7.54E+01 8.29E+01 3.19E+01 2.13E+01 1.16E+01 4.11E-01 83m Kr 0 1.65E+00 5.88E+00 1.02E+01 2.02E+01 9.84E+00 2.20E+00 5.47E-03 0 85m Kr 0 3.77E+00 1.42E+01 2.59E+01 6.40E+01 4.81E+01 2.55E+01 2.03E+00 0 85 Kr 0 2.28E-01 8.97E-01 1.70E+00 4.91E+00 5.06E+00 5.06E+00 5.03E+00 2.24E+00 87 Kr 0 5.66E+00 1.94E+01 3.21E+01 5.36E+01 1.85E+01 2.07E+00 3.25E-04 0 88 Kr 0 9.33E+00 3.44E+01 6.13E+01 1.38E+02 8.69E+01 3.22E+01 6.10E-01 0 89 Kr 0 1.78E-02 2.66E-03 1.92E-04 1.17E-09 5.56E-21 1.17E-43 0 0 131m Xe 0 1.55E-01 6.08E-01 1.15E+00 3.32E+00 3.40E+00 3.37E+00 3.22E+00 2.70E-01 133m Xe 0 9.17E-01 3.59E+00 6.78E+00 1.93E+01 1.94E+01 1.85E+01 1.50E+01 1.07E-03 133 Xe 0 3.01E+01 1.18E+02 2.24E+02 6.43E+02 6.59E+02 6.49E+02 6.06E+02 6.47E+00 135m Xe 0 1.63E+00 3.29E+00 3.20E+00 6.44E-01 3.22E-03 7.58E-08 0 0 135 Xe 0 7.70E+00 2.87E+01 5.35E+01 1.50E+02 1.61E+02 1.57E+02 8.24E+01 0 137 Xe 0 1.25E-01 3.43E-02 4.53E-03 3.09E-07 1.77E-16 5.52E-35 0 0 138 Xe 0 5.87E+00 1.25E+01 1.28E+01 3.21E+00 2.47E-02 1.38E-06 0 0 Total noble gases 0 6.72E+01 2.42E+02 4.33E+02 1.10E+03 1.01E+03 8.95E+02 7.14E+02 8.98E+00 Alkali metals 0 2.75E+00 3.18E+00 3.33E+00 3.38E+00 4.51E-01 1.07E-01 3.34E-04 1.34E-04 Te-group 0 8.69E-01 6.92E+00 8.69E+00 8.45E+00 1.08E+00 2.29E-01 5.99E-04 9.28E-05 Noble metals 0 0.00E+00 5.18E-01 6.76E-01 6.92E-01 8.96E-02 2.03E-02 5.60E-05 6.92E-06 La-group 0 0.00E+00 8.11E-02 1.12E-01 1.12E-01 3.36E-02 1.09E-02 4.65E-05 1.74E-05 Ce-group 0 0.00E+00 3.82E-01 5.04E-01 5.30E-01 6.92E-02 1.57E-02 4.11E-05 1.81E-06 C OLUMBIA G ENERATING S TATION Amendment59 F INAL S AFETY A NALYSIS R EPORTDecember 2007LDCN-05-009 15.6-27 Table 15.6-11 Loss-of-Coolant Accident Activity Released to th e Environment (Curies) - 20 Minute Drawdown Isotope 0.25 hrs 0.5 hrs 0.75 hrs 1 hrs 2 hrs 4 hrs 8 hrs 24 hrs 30 d 131 I 6.59E+01 1.39E+02 1.80E+02 2.31E+02 4.48E+02 5.49E+02 6.37E+02 8.35E+02 3.03E+03 132 I 8.83E+01 1.82E+02 2.31E+02 2.89E+02 5.27E+02 7.45E+02 1.13E+03 2.52E+03 1.15E+04 133 I 1.28E+02 2.69E+02 3.47E+02 4.44E+02 8.49E+02 1.03E+03 1.17E+03 1.42E+03 1.67E+03 134 I 1.23E+02 2.45E+02 2.99E+02 3.53E+02 4.98E+02 5.25E+02 5.28E+02 5.28E+02 5.28E+02 135 I 1.17E+02 2.44E+02 3.13E+02 3.97E+02 7.35E+02 8.74E+02 9.63E+02 1.05E+03 1.06E+03 Total iodines 5.22E+02 1.08E+03 1.37E+03 1.71E+03 3.06E+03 3.72E+03 4.43E+03 6.35E+03 1.78E+04 83m Kr 1.09E+01 4.59E+01 1.47E+02 3.80E+02 2.19E+03 4.92E+03 6.81E+03 7.35E+03 7.35E+03 85m Kr 2.34E+01 1.02E+02 3.43E+02 9.21E+02 6.15E+03 1.67E+04 2.99E+04 4.36E+04 4.42E+04 85 Kr 1.34E+00 6.00E+00 2.09E+01 5.82E+01 4.33E+02 1.37E+03 3.25E+03 1.07E+04 1.64E+05 87 Kr 3.98E+01 1.62E+02 5.03E+02 1.25E+03 6.46E+03 1.27E+04 1.55E+04 1.59E+04 1.59E+04 88 Kr 5.95E+01 2.55E+02 8.42E+02 2.22E+03 1.40E+04 3.49E+04 5.53E+04 6.71E+04 6.73E+04 89 Kr 9.01E+00 1.13E+01 1.16E+01 1.16E+01 1.16E+01 1.16E+01 1.16E+01 1.16E+01 1.16E+01 131m Xe 9.12E-01 4.07E+00 1.42E+01 3.94E+01 2.93E+02 9.28E+02 2.18E+03 7.06E+03 5.55E+04 133m Xe 5.42E+00 2.41E+01 8.40E+01 2.33E+02 1.72E+03 5.38E+03 1.24E+04 3.71E+04 9.18E+04 133 Xe 1.77E+02 7.92E+02 2.76E+03 7.66E+03 5.68E+04 1.80E+05 4.22E+05 1.35E+06 6.45E+06 135m Xe 2.28E+01 6.80E+01 1.40E+02 2.36E+02 4.35E+02 4.58E+02 4.59E+02 4.59E+02 4.59E+02 135 Xe 4.33E+01 1.98E+02 6.85E+02 1.86E+03 1.34E+04 4.24E+04 9.96E+04 2.67E+05 3.33E+05 137 Xe 2.69E+01 3.79E+01 3.99E+01 4.04E+01 4.04E+01 4.04E+01 4.04E+01 4.04E+01 4.04E+01 138 Xe 7.66E+01 2.35E+02 5.02E+02 8.78E+02 1.74E+03 1.87E+03 1.87E+03 1.87E+03 1.87E+03 Total noble gases 4.97E+02 1.94E+03 6.09E+03 1.58E+04 1.04E+05 3.02E+05 6.49E+05 1.81E+06 7.23E+06 Alkali metals 2.99E+01 6.16E+01 7.76E+01 9.54E+01 1.68E+02 1.96E+02 2.10E+02 2.14E+02 2.16E+02 Te-group 6.96E+00 1.64E+01 3.87E+01 8.25E+01 2.70E+02 3.38E+02 3.71E+02 3.81E+02 3.82E+02 Noble metals 0 0 1.54E+00 4.94E+00 2.01E+01 2.56E+01 2.84E+01 2.93E+01 2.94E+01 La-group 0 0 2.30E-01 7.72E-01 3.23E+00 4.56E+00 5.82E+00 6.30E+00 6.65E+00 Ce-group 0 0 1.13E+00 3.66E+00 1.51E+01 1.94E+01 2.15E+01 2.22E+01 2.22E+01 C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 15.6-28 Table 15.6-12
Loss-of-Coolant Accident (Design Basis Analysis)
Radiological Effects
Total Effect TEDE (rem) Exclusion area (1950 m) (2 hr) 4.1 Low population zone (4827 m) (30 days)
4.0 Control
Room (30 days) 3.5 C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-09-007 15.6-29 Table 15.6-13
Sequence of Events for Feedwater Line Break Outside Containment
Time Event 0 One feedwater line breaks.
0+ Feedwater line check valves isolate the reactor from the break. <30 sec At low reactor water level, reactor scram would initiate and, at low-low reactor water level, HPCS and MSIV closure b would initiate and recirculation pumps would trip.
2 minutes a The SRVs open and close and maintain the reactor vessel pressure at approximately 1100 psig. 1 to 2 hr Normal reactor cooldown established.
a Approximately.
b The analysis has not been updated for the change in MSIV isolation setpoint from Level 2 to Level 1 because it remains bounded by the r ecirculation line break LOCA (Reference 15.6-9).
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 15.6-30 Table 15.6-14
Feedwater Line Break Accident - Parameters Tabulated for Postulated Accident Analysis
Parameter Value I. Data and assumptions used to estimate radioactive source from postulated accidents A. Power level N/A B. Burnup N/A C. Fuel damaged None D. Release of activity by nuclide Table 15.6-15 E. Iodine fractions (1) Organic 0 (2) Elemental 1% (3) Particulate 0 (4) Reactor coolant activity before the accident Section 15.6.6.5.1 II. Data and assumptions used to estimate activity released A. Primary containment leak rate (%/day)
N/A B. Secondary containmen t leak rate (%/day)
N/A C. RWCU total isolation valve closure time (sec) 75 D. Adsorption and filtr ation efficiencies (1) Organic iodine N/A (2) Elemental iodine N/A (3) Particulate iodine N/A (4) Particulate fission products N/A E. Recirculation system parameters N/A (1) Flow rate N/A (2) Mixing efficiency N/A (3) Filter efficiency N/A F. Containment spray parameters (flow rate, drop size, etc.) N/A G. Containment volumes N/A H. All other pertinent da ta and assumptions None C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 15.6-31 Table 15.6-14
Feedwater Line Break Accident - Parameters Tabulated for Postulated A ccident Analysis (Continued)
Parameter Value III. Dispersion data A. Boundary and LPZ distance (m) 1950/4827 B. /Qs for time intervals of 0-2 hr - EAB/LPZ 2.62 x 10-4/1.06 x 10
-4 IV. Dose data A. Method of dose calculation Reference 15.6-2 B. Dose conversion assumptions Reference 15.6-2 C. Peak activity concentrations in containment N/A D. Doses Table 15.6-16 C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 15.6-32 Table 15.6-15
Feedwater Line Break Accident Activity Release to Environment (Curies)
Isotope Activity 131 I 2.22 x 10-2 132 I 2.05 x 10-1 133 I 1.52 x 10-1 134 I 4.45 x 10-1 135 I 2.22 x10-1 Total 1.04 x 10
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 15.6-33 Table 15.6-16
Feedwater Line Break Accident Biological Effects of a Puff Release
Area Whole Body Dose (rem) Thyroid Dose (rem) Exclusion area (1950 m) 1.37 x 10-4 5.47x10-3 Low population zone (4827 m) 5.53 x 10-5 2.21x10-3
Steam Flow Schematic for Steam Break Outside Containment 900547.74 15.6-2 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Main Condenser MainTurbine Bypass LineSteam Tunnel Feedwater Pump CheckValve Flow Limiter FlowP ReactorVessel Containment Break Main Steam LineIsolation ValvesTurbine BuildingTurbine AdmissionValveTurbineStop Valve Columbia Generating StationFinal Safety Analysis Report 900547.75 Columbia Generating StationFinal Safety Analysis Report Leakage Path for LOCADraw. No.Rev.Figure Amendment 59December 2007 15.6-3 Form No. 960690FH LDCN-05-009 Drywell Secondary Containment Core SGT MSIV leakage from failed steamlineMSIV leakage from the 3 intact steamlines Suppression Pool Wet Well Air-SpaceESF leakage (1 gpm) t < 20 min plateout PC leakage for t < 20 min SC bypass leakage Primary Containment leakage PC leakage t > 20 min KK-doors + RB Walls Environment t > 20 min Leakage Path for Feedwater Line Break Outside Containment 900547.76 15.6-4 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Turbine Break CheckValve ReactorVessel (Steam)Pumps/Heaters/
Controls Pumps/Demin/
Controls Bypass Main Condenser Hotwell Flow CheckValve CheckValve Containment Columbia Generating StationFinal Safety Analysis Report C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 15.7-1 15.7 RADIOACTIVE RELEASE FROM SUBSYSTEMS AND COMPONENTS These events are classified as nonlimiting events for both original and uprated power conditions. Therefore, no further analysis has been performed.
15.7.1 RADIOACTIVE GAS WASTE SYSTEM LEAK OR FAILURE
Not applicable.
15.7.2 LIQUID RADIOACTIVE SYSTEM FAILURE
Not applicable.
15.7.3 POSTULATED RADIOACTIVE RELEASES DUE TO LIQUID RADWASTE TANK FAILURE
15.7.3.1 Identification of Causes and Frequency Classification
15.7.3.1.1 Identification of Causes
The liquid radwaste tanks are constructed to specific engineering codes and standards and to the uniform building code seismi c requirements. Thes e tanks operate at at mosphere pressure and low temperatures. A positive action interlock system is pr ovided to prevent inadvertent opening of a drain valve because of operator error. Accordingly, the possibility of a complete tank failure or drainage is considered small.
An unspecified event is postulated to cause the complete release of the average radioactivity inventory in the tank containing the largest quantities of signifi cant radionuclides in the liquid radwaste system. The tank postulated to rupture is one of th e two decontamination solution concentrated waste tanks (see Figure 11.2-1
).
15.7.3.1.2 Frequenc y Classification
This accident is categorized as a limiting fault.
15.7.3.2 Sequence of Events and Systems Operation 15.7.3.2.1 Sequence of Events
The sequence of events expected to occur is as follows:
C OLUMBIA G ENERATING S TATION Amendment 55 F INAL S AFETY A NALYSIS R EPORT May 2001 LDC N-0 0-0 3 4 15.7-2 Sequence of Events - Liquid Radwaste Tank Failure Events Time Event begins-failure occurs 0 Area radiation alarms alert plant personnel 1 minute Operator action begins 10 minute 15.7.3.2.2 Systems Operation Failure of a concentrated wast e tank does not require a shutdow n nor does it impair a safe shutdown. It will lead to limited operation of the concentrated waste system using the
remaining tank.
The liquid contents of this tank will also be contained by the building walls and an unlined, 18-in. high concrete dike around the radwaste tank area. Floor drain sump pumps would receive a high water level alarm, activate automatically, and remove the spilled liquid.
15.7.3.2.3 The Effects of Single Failures and Operator Errors This event has been an alyzed without taking credit for any expected operator action or system operation; therefore, a discussi on of single equipment failure or single operator error is not applicable.
15.7.3.3 Core and Sy stem Performance
The failure of this liquid radwas te system component does not dir ectly affect th e nuclear steam supply system (NSSS). It will lead to d ecoupling of NSSS with the subject system.
This failure has no applicable effect on the reactor core or the NSSS safety performance. Specific assumptions and parameters are presented in Table 15.7-1.
15.7.3.4 Barrier Performance
This event does not involve any containment barrier integrity except the tank itself and the radwaste building. The dike and walls of the radwaste building surrou nding the tanks are built to Seismic Category I criteria.
In the analysis of spill conseque nces, no credit is taken for the dike or radwaste building in recontaining the spilled liquid.
C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 LDCN-03-003 15.7-3 15.7.3.5 Radiological Consequences
The entire volume (700 gal) of concentrator waste tank assumed to spill with isotope inventory given in Table 11.2-1. Tritium concentration is assumed to be 0.01
µCi/ml (Environmental Report, Operating Lice nse Section 3.5.1).
The hypothetical radwaste tank fa ilure was evaluated using cons ervative assumpti ons such as no containment in the radwaste building and unim peded flow vertically through 50-60 ft of sand and gravel.
The following offsite con centration data for the radionuclides of interest are provided for the WNP-1/4 wells and at the Columbia River:
Radionuclide Concentration at WNP-1/4 Wells
(µCi/ml) Concentration at Columbia River
(µCi/ml) Concentration Limit (µCi/ml) 3 H 1.0 x 10-7 1.3 x 10-8 1 x 10-3 90 Sr 1.7 x 10-4 4.2 x 10-7 5 x 10-7 137 Cs 2.2 x 10-10 1.4 x 10-27 1 x 10-6 The calculations show the strontium concentration exceeding the unrestricted area limitation at the WNP-1/4 wells. These wells are a temporary water supply and are under the control of Energy Northwest. Should a spill occur there w ill be ample time to as sess the severity and extent of contamination.
Concentration at the river bank will be diluted by the river flow. The nearest surface water users are several miles downstream.
15.7.4 FUEL HANDLING ACCIDENT
15.7.4.1 Identification of Causes and Frequency Classification
15.7.4.1.1 Identification of Causes
The fuel handling accident is assumed to occur as a consequence of a failure of the fuel assembly lifting mechanism resulting in dropping a raised irradiated fu el assembly onto other fuel bundles seated in the reactor pressure ve ssel (RPV). The event was selected as the bounding event because it cons iders the maximum height a nd weight, while assuming a minimum water level above the damaged fuel.
15.7.4.1.2 Frequenc y Classification
This event is categorized as a limiting fault.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 15.7-4 15.7.4.2 Sequence of Events and Systems Operation
The following sequence of events is assumed in the analysis.
No fuel movement will take place in the first 24 hr following shutdown. At 24 hr post-shutdown fuel movement starts and the fuel handling equipment is assumed to fail dropping the fuel grapple and an irradiated fuel bundle onto the irradiated fuel bundles seated in the RPV.
Fuel is damaged and fission pro ducts are released to the reacto r coolant, then to the reactor building atmosphere, and finally to the environmen t over a 2-hr period.
No credit is taken for holdup or mixing in the reactor building, nor is credit taken for filtration by the standby gas treatment (SGT).
15.7.4.2.1 The Effects of Single Failures and Operator Errors
No systems or operator actions are credited to mitigate a fuel handling accident.
15.7.4.3 Core and Sy stem Performance
15.7.4.3.1 Mathematical Model
Because of the complex nature of the impact and the resulting damage to fuel assembly components, a rigorous prediction of the number of failed rods is not possible. For this reason, a simplified energy appr oach was taken and numerous c onservative assumptions were made to ensure a conservative estimat e of the number of failed rods.
The kinetic energy acquired by a falling fuel assembly may be dissipated in one or more impacts. The energy absorption on successive impacts is estimated by considering a plastic impact.
The energy transferred in the dropped assembly is considered in two phases. First, the fuel assembly is expected to impact on the reactor core at a small angle from the vertical, inducing a bending mode of failure on the fuel rods of th e dropped assembly. The kinetic energy of the fall is dissipated in the impact. The analysis assumes that the energy of the dropped assembly is absorbed by only the cladding and other core structures. The assu mption that no energy is absorbed by the fuel material results in considerable c onservatism in the mass-energy calculations. Half of the energy is dissipated in the structure of the dropped assembly, failing all the rods in the assembly. The remaining half is allocated evenly across the structural mass of the impacted assemblies.
The energy dissipated by the claddi ng is calculated by multiplying the impacted assembly energy by the cladding mass fraction and dividing by the energy required to fail a rod (based on 1%
uniform plastic deformation).
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-11-004 15.7-5 The second phase considers the ki netic energy developed by the i rradiated fuel assembly and lifting mechanism tipping over and impacting th e core horizontally. The kinetic energy developed is equal to the initial potential energy of the assembly re lative to the top of the core.
Again, half of the energy is absorbed by the dropped assemb ly and half by the impacted assemblies. The number of failed rods in the impacted assemblies is determined using the cladding mass fraction and the energy required to fail a rod.
15.7.4.3.2 Input Paramete rs and Initial Conditions The parameters and cond itions used to determine the number of failed rods are listed below:
- a. The fuel assembly is dropped from a height of 34 ft. The maximum height allowed by the fuel handling equipment is less than 34 ft;
- b. The dropped mass consists of a fuel assembly (586 lb bo unding analyzed wet weight for GE14 10 x 10 fuel , 617 lb wet weight for GE 8 x 8 fuel and 665 lb dry weight for ATRIUM-10) and the fuel grapple (350 lb wet weight);
- c. The energy required to fail a fuel rod is approximately 175 ft-lb for GE14 10 x 10 fuel, 250 ft-lb for GE 8 x 8 fu el and 205 ft-lb for ATRIUM-10; see Reference 15.7-3 for SVEA-96.
15.7.4.3.3 Results
Based on a core of GE 8 x 8 fuel, the calculation predicts 124 failed fuel rods; 62 rods in the dropped assembly, 43 rods in the first impact, and 19 additional rods in the second impact.
Westinghouse analysis pr edicts a maximum of 123 failed rods (Reference 15.7-3) and AREVA NP calculated that up to 156 rods could fail (Reference 15.7-4). Analysis of the GE14 10 x 10 fuel estimates that a total of 151 fuel rods will fail (Reference 15.7-5).
15.7.4.4 Barrier Performance
The reactor coolant pressure boundary, primar y containment and sec ondary containment are open at the time of the accident. However, a similar event could occur in the spent fuel pool (SFP), during spent fuel transfer from the RPV to, or handling in, the SFP. Assuming a drop height of 4 ft, the number of failed rods as a result of a GE 8 x 8 bundle (uncha nneled) drop in the SFP was calculated and fou nd to be 90 rods; the number for channeled GE 8 x 8 is less than that. The dose analysis for a drop of a bundle over the core, which assumes 250 failed rods, bounds a drop of a bundle in the SFP.
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 15.7-6 15.7.4.5 Radiological Consequences The fission product inventory is based on a plan t-specific ORIGEN 2 run for pre-power uprate basis of 3489 MW with 1000 days exposure adjusted as described in Section 15.4.9.5.1. The release is based on da mage to 250 fuel rods. A 24-hr period for decay from the power condition is assumed.
Figure 15.7-1 indicates the leakage flow path for this accident.
15.7.4.5.1 Design Basis Analysis
Specific values of parameters used in the evaluation are presented in Table 15.7-2. The dispersion coefficients used to determine offsite doses are presented in Table 15.0-4. 15.7.4.5.1.1 Fission Product Release From Fuel. The fission product inventory of a core average exposure fuel rod is adju sted by a peaking factor of 1.7 to establish the inventory of each damaged rod. Five percent of the noble gases inventory (10% for 85 Kr) and 5% of the iodine inventory (8% for 131I), and 12% of the alkali metals inventory are assumed to be released to the reactor well. The activity airborne in the seconda ry containment is presented in Table 15.7-3.
15.7.4.5.1.2 Fission Product Tr ansport to the Environment. The transport pathway consists of mixing in the reactor well water, migration from the r eactor well to the secondary containment atmosphere, and release to the environment without passing through the SGT. All of the noble gas, 0.5% of the iodines, and 0% of the alkali metals are assumed to become airborne in the secondary containment (Reference 15.7-1).
From the activity airborne in th e reactor building, 99% is releas ed to the environment in 2 hr.
The release of activity to the environment is presented in Table 15.7-4.
15.7.4.5.1.3 Results. The calculated doses for the design basis analysis are presented in Table 15.7-5 and are within the limits of 10 CFR 50.67.
15.7.5 SPENT FUEL CASK DROP ACCIDENT
The spent fuel cask is equippe d with ANSI N14.6 (Reference 15.7-2) compliant lifting lugs and a lifting yoke compatible with the reactor building crane main hook. The reactor building crane is provided with sufficient redundancy such that no credib le postulated failure of any crane component required to lift, hold, and move loads, will result in the dropping of the fuel cask. Therefore, an analysis of the spent fuel cask dr op is not required.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-029 15.7-7 15.
7.6 REFERENCES
15.7-1 Energy Northwest, "C olumbia Generating Station Alternative Source Term,"
CGS-FTS-0168, Revision 0, August 2007.
15.7-2 "Special Lifting Devices for Shippi ng Containers Weighing 10,000 Pounds (4500 kg) or More," ANSI N14.6-1993, June 1993.
15.7-3 ABB/Combustion Engineering, "Fuel Assembly Mechanical Design Report for WNP-2," CE NPSD-792-P, May 1996.
15.7-4 AREVA NP, "Columbi a Generating Station Cycle 19 Reload Analysis,"
ANP-2602, Revision 0, March 2007.
15.7-5 GEH-0000-0075-4920, "G E14 Fuel Design Cycle-I ndependent Analyses for Energy Northwest Columbia Generati ng Station" (most recent version referenced in the COLR).
C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 15.7-9 Table 15.7-1
Liquid Radwaste Tanks Failure - Parameters and Concentrations Parameter Value I. Data and assumptions used to estimate radioactive source Entire volume (700 gal) of concentrator waste tank assumed to spill with isotope inventory
given in Table 11.2-1. Tritium concentration assumed to be
0.01 mCi/ml from the CGS ER-OL. II. Data and assumptions used to estimate activity released A. Containment leak rate (%/day) N/A B. Secondary containment leak rate (%/day) N/A C. Valve movement times N/A D. Absorption and filtration efficiencies N/A (1) Organic iodine N/A (2) Elemental iodine N/A (3) Particulate iodine N/A (4) Particulate fission products N/A E. Recirculation system parameters N/A (1) Flow rate N/A (2) Mixing efficiency N/A (3) Filter efficiency N/A F. Containment spra y parameters (flow rate, drop size, etc.)
N/A G. Containment volumes N/A H. Other pertinent data and assumptions See Section 2.4.13.3 III. Concentration data
@ WNP-1/4 Wells Radionuclide
@ Col. R (µCi/ml) Conc. Limit a (µCi/ml) (µCi/ml) 3 H 1.0 x 10-7 1.3 x 10-8 1 x 10-3 90 Sr 1.7 x 10-4 4.2 x 10-7 5 x 10-7 137 Cs 2.2 x 10-10 1.4 x 10-27 1 x 10-6 a From 10 CFR Part 20.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 15.7-10 Table 15.7-2 Fuel Handling Accident Parameters Tabulated for Postulated Accident Analysis
Parameters Design Basis Assumptions I. Data and assumptions used to estimate radioactive source from postulated accidents A. Power level Fuel decay period 3556 24 hrs B. Radial peaking factor 1.7 C. Assumed fuel damaged Bundles in the core Rods per bundle 250 rods 764 62 D. Release of activity fr om the gap to the reactor well water Figure 15.7-1 E. Iodine species fractions released Figure 15.7-1 (1) Organic (2) Elemental (3) Particulate F. Reactor coolant activity before the accident N/A II. Data and assumptions used to estimate activity released A. Primary containment leak rate (%/day) N/A B. Secondary containment release rate 99% of the activity in 2 hr with a flow rate of 2.3 SC volumes per hr C. Valve movement times N/A D. SGT filtration N/A E. Scrubbing by reactor well water Figure 15.7-1 (1) Organic iodine (2) Elemental iodine (3) Particulate iodine (4) Particulate alkali metals
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 15.7-11 Table 15.7-2
Fuel Handling Accident Parameters Tabulated for Postulated Accident Analysis (Continued)
Parameters Design Basis Assumptions F. Recirculation system parameters (1) Flow rate N/A (2) Mixing efficiency N/A (3) Filter efficiency N/A G. Containment spray parameters (flow rate, drop size, etc.)
N/A H. Containment volumes N/A I. Other pertinent data and assumptions (1) SGT filtration None (2) CREF filtration None (3) Holdup in reactor building None (4) Mixing in reactor building None III. Dispersion data (for dura tion of release, 0 - 2 hr)
(1) Offsite Table 15.0-4 (2) Control room 8.69E-4 sec/m 3 IV. Dose data A. Method of dose calculation RegulatoryGuide1.183 B. Dose conversion assumptions RegulatoryGuide1.183 C. Peak activity concentrations in containment N/A D. Doses Table 15.7-5 C OLUMBIA G ENERATING S TATION Amendment59 F INAL S AFETY A NALYSIS R EPORTDecember 2007LDCN-05-009 15.7-12 Table 15.7-3 Fuel Handling Accident Activity Airborne in Seconda ry Containment (Curies)
Isotope 6 minutes 12 minutes 0.5 hr 1 hr 2 hr 4 hr 8 hr 1 day 4 days 30 days 131 I 2.64E+02 2.10E+02 1.05E+02 3.31E+01 3.29E+00 3.26E+00 3.22E+00 3.04E+00 2.35E+00 2.51E-01 132 I 1.95E-01 1.50E-01 6.87E-02 1.87E-02 1.39E-03 7.65E-04 2.33E-04 1.98E-06 9.73E-16 6.23E-20 133 I 1.58E+02 1.25E+02 6.19E+01 1.93E+01 1.86E+00 1.74E+00 1.52E+00 8.96E-01 8.21E-02 8.31E-11 134 I 1.54E-06 1.13E-06 4.44E-07 9.40E-08 4.21E-09 8.45E-10 3.40E-11 8.99E-17 6.50E-19 1.25E-36 135 I 2.72E+01 2.14E+01 1.04E+01 3.11E+00 2.80E-01 2.28E-01 1.51E-01 2.91E-02 1.75E-05 1.10E-22 Total iodine 4.49E+02 3.56E+02 1.77E+02 5.55E+01 5.43E+00 5.23E+00 4.90E+00 3.97E+00 2.43E+00 2.51E-01 83m Kr 5.55E-01 4.25E-01 1.90E-01 4.98E-02 3.42E-03 1.62E-03 3.62E-04 9.06E-07 1.78E-18 7.63E-20 85m Kr 2.10E+02 1.64E+02 7.85E+01 2.29E+01 1.95E+00 1.42E+00 7.57E-01 6.04E-02 6.91E-07 5.85E-21 85 Kr 1.06E+03 8.42E+02 4.22E+02 1.33E+02 1.33E+01 1.33E+01 1.33E+01 1.33E+01 1.33E+01 1.32E+01 87 Kr 3.24E-02 2.43E-02 1.03E-02 2.49E-03 1.44E-04 4.81E-05 5.39E-06 8.50E-10 9.87E-22 1.21E-30 88 Kr 6.26E+01 4.85E+01 2.26E+01 6.30E+00 4.91E-01 2.99E-01 1.11E-01 2.11E-03 3.80E-11 4.03E-22 89 Kr 6.66E-19 1.43E-19 1.54E-21 1.49E-23 1.56E-24 2.38E-20 3.88E-19 4.10E-20 1.19E-19 4.03E-51 131m Xe 3.39E+02 2.70E+02 1.35E+02 4.26E+01 4.24E+00 4.22E+00 4.18E+00 4.02E+00 3.38E+00 7.55E-01 133m Xe 1.58E+03 1.25E+03 6.26E+02 1.97E+02 1.94E+01 1.89E+01 1.80E+01 1.47E+01 5.95E+00 2.34E-03 133 Xe 6.73E+04 5.34E+04 2.68E+04 8.50E+03 9.24E+02 1.30E+03 1.98E+03 3.78E+03 4.60E+03 1.62E+02 135m Xe 1.92E-20 1.17E-20 2.64E-21 2.22E-22 1.88E-24 1.85E-26 4.61E-28 1.38E-20 1.72E-21 7.32E-53 135 Xe 1.67E+04 1.32E+04 6.58E+03 2.14E+03 3.64E+02 1.02E+03 1.71E+03 1.39E+03 1.22E+01 5.37E-20 137 Xe 6.42E-21 1.76E-21 3.71E-23 2.48E-25 6.80E-27 4.31E-24 1.39E-19 1.11E-20 1.01E-20 8.74E-51 138 Xe 1.73E-20 1.07E-20 2.58E-21 2.42E-22 2.50E-24 2.92E-26 4.76E-28 1.83E-21 2.69E-22 3.34E-52 Total noble gases 8.73E+04 6.92E+04 3.47E+04 1.10E+04 1.33E+03 2.36E+03 3.73E+03 5.20E+03 4.63E+03 1.76E+02 C OLUMBIA G ENERATING S TATION Amendment59 F INAL S AFETY A NALYSIS R EPORTDecember 2007LDCN-05-009 15.7-13 Table 15.7-4 Fuel Handling Accident Activity Released to th e Environment (Curies)
Isotope 6 minutes 12 minutes 0.5 hr 1 hr 2 hr 4 hr 8 hr 1 day 4 days 30 days 131 I 6.84E+01 1.23E+02 2.28E+02 2.98E+02 3.28E+02 3.28E+02 3.29E+02 3.29E+02 3.29E+02 3.29E+02 132 I 5.12E-02 9.07E-02 1.63E-01 2.07E-01 2.23E-01 2.23E-01 2.23E-01 2.23E-01 2.23E-01 2.23E-01 133 I 4.09E+01 7.33E+01 1.35E+02 1.77E+02 1.95E+02 1.95E+02 1.95E+02 1.95E+02 1.95E+02 1.95E+02 134 I 4.16E-07 7.21E-07 1.23E-06 1.49E-06 1.56E-06 1.56E-06 1.56E-06 1.56E-06 1.56E-06 1.56E-06 135 I 7.07E+00 1.26E+01 2.31E+01 3.01E+01 3.28E+01 3.28E+01 3.28E+01 3.28E+01 3.28E+01 3.28E+01 Total iodine 1.16E+02 2.09E+02 3.86E+02 5.06E+02 5.56E+02 5.56E+02 5.57E+02 5.57E+02 5.57E+02 5.57E+02 83m Kr 1.47E-01 2.59E-01 4.61E-01 5.81E-01 6.21E-01 6.21E-01 6.21E-01 6.21E-01 6.21E-01 6.21E-01 85m Kr 5.49E+01 9.78E+01 1.78E+02 2.30E+02 2.50E+02 2.50E+02 2.50E+02 2.50E+02 2.50E+02 2.50E+02 85 Kr 2.75E+02 4.93E+02 9.13E+02 1.20E+03 1.32E+03 1.32E+03 1.32E+03 1.32E+03 1.32E+03 1.32E+03 87 Kr 8.63E-03 1.51E-02 2.64E-02 3.28E-02 3.47E-02 3.47E-02 3.47E-02 3.47E-02 3.47E-02 3.47E-02 88 Kr 1.64E+01 2.92E+01 5.26E+01 6.73E+01 7.26E+01 7.26E+01 7.26E+01 7.26E+01 7.26E+01 7.26E+01 89 Kr 3.64E-19 4.43E-19 4.64E-19 4.64E-19 4.64E-19 4.64E-19 4.65E-19 4.78E-19 4.38E-16 1.27E-13 131m Xe 8.80E+01 1.58E+02 2.92E+02 3.85E+02 4.23E+02 4.23E+02 4.23E+02 4.23E+02 4.23E+02 4.23E+02 133m Xe 4.10E+02 7.35E+02 1.36E+03 1.79E+03 1.96E+03 1.96E+03 1.96E+03 1.96E+03 1.96E+03 1.96E+03 133 Xe 1.74E+04 3.13E+04 5.79E+04 7.63E+04 8.40E+04 8.40E+04 8.40E+04 8.40E+04 8.40E+04 8.40E+04 135m Xe 5.74E-21 9.23E-21 1.34E-20 1.46E-20 1.47E-20 1.47E-20 1.47E-20 1.47E-20 9.44E-18 1.02E-13 135 Xe 4.34E+03 7.78E+03 1.44E+04 1.89E+04 2.10E+04 2.10E+04 2.10E+04 2.10E+04 2.10E+04 2.10E+04 137 Xe 3.03E-21 3.86E-21 4.16E-21 4.17E-21 4.17E-21 4.17E-21 4.31E-21 2.95E-20 8.44E-16 1.27E-13 138 Xe 5.10E-21 8.26E-21 1.22E-20 1.34E-20 1.35E-20 1.35E-20 1.35E-20 1.35E-20 5.69E-18 1.05E-13 Total noble gases 2.26E+04 4.06E+04 7.51E+04 9.89E+04 1.09E+05 1.09E+05 1.09E+05 1.09E+05 1.09E+05 1.09E+05 C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 15.7-14 Table 15.7-5
Fuel Handling Accident (Design Basis Analysis)
Radiological Effects
Exclusion area (1950 m) (2 hr) 1.0 Low population zone (4827 m) (30 day)
0.3 Control
room (30 day) 3.7 900547.77 Columbia Generating StationFinal Safety Analysis Report Leakage Path for Fuel Handling AccidentDraw. No.Rev.Figure Amendment 59December 2007 15.7-1 Form No. 960690FH LDCN-05-009 Core Gap Reactor Well Water (Scrubbing)
Secondary Containment Environment:
EAB and LPZ Control Room Only noble gases, elemental and organic iodines get released to the environment and to the Control Room No CREF ~ltration(1)No SGT Filtration(2)No Holdup in Reactor Building (3)No Mixing in Reactor Building Particulate*
Elemental Organic% Scrubbed 100.000 99.714 0.286% Released 0.000 0.286 99.714 DF 350 1 Scrubbing by Reactor Well Water Iodines I-131 Noble Gas Kr-85 Alkali Metals (Cs & Rb)Particulate 0.95 0.95 0.00 0.00 1.00 Elemental 0.0485 0.0485 1.0000 1.0000 0.0000 Organic 0.0015 0.0015 0.0000 0.0000 0.0000 Species Fractions Released
% Activity Released 5 8
5 10 12*All Alkali Metals are particulates and get 100% scrubbed, All particulate iodines convert to elemental.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009,07-011 15.8-1 15.8 ANTICIPATED TRANS IENTS WITHOUT SCRAM
15.8.0 CAPABILITIES OF PRES ENT DESIGN TO ACCOMMODATE ANTICIPATED TRANSIENTS WITHOUT SCRAM
The anticipated transien ts without scram (ATWS) events de scribed in this section are not design basis events for Colu mbia Generating Station (CGS
). A proposed method for minimizing the effects of failure-to-scram is describe d in References 15.8-1 and 15.8-2. The recirculation pump trip (RPT), alternate rod insertion (ARI), and two pump standby liquid control (SLC) system operation features are utilized at CGS to provide protection against failure to scram. Due to th e CGS design feature utilizing SLC system injection through the high-pressure core spray (HPCS) header, a plant-unique analysis was performed to demonstrate ATWS protection and mitig ation at pre-power uprate conditions.
The ATWS acceptance criteria are establis hed in Reference 15.8-3 as: a. The reactor coolant pressure boun dary (RCPB) remains below emergency pressure limits,
- b. The containment pressure remains below design limits. The suppression pool temperature remains below local saturation temperature limits as defined in Reference 15.8-3 ,
- c. A coolable geometry is maintained,
- d. Radiological releases are maintained within 10 CFR 100 allowable limits. With implementation of Alternate Source Te rm (AST), the radiological release acceptance criterion becomes 10 CFR 50.67, and
- e. Equipment necessary to mitigate the postulated ATWS event are evaluated to provide a high degree of assurance (assurance of function) that it will function in the environment (pressure, temperatur e, humidity, and radiation) predicated to occur as a result of the ATWS event.
The ATWS analysis, performed in conformance with NEDE-2422 2, did not include a SLCS pump suction valve delay in the SLCS injection time. To determine the impact of the 35 sec opening time for the suction valves upstream of the SLCS pumps, the limiting ATWS event for peak suppression pool temperatur e (i.e., MSIVC) was analyzed with the 35 sec delay in SLC system injection time. The re sults presented for the hot shutdown time, peak suppression pool temperature, and peak contai nment pressure for MSIVC in Table 15.8-3 include the effects of the 35 sec delay.
C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 15.8-2 Section 15.8.9 shows that for the ATWS ev ent with the most severe heat flux transient, fuel related applicable limits were met with considerable margin
- s. In addition, Reference 15.8-3 concludes that maximum peak cla dding temperature will not exceed 2200
°F and the maximum local oxidation will be much less than 17%. Thus, criteria 3 and 4 are shown to be satisfied by the plant specific and generic analyses. Sections 15.8.7 and 15.8.9 show that resulting primary system pressures will be less than emergency pressure limit an d that suppression pool temperature increase and peak pressure are within design limits. Reference 15.8-3 concludes that the safety/relief valve (SRV) air clearing loads will be bounde d by the design loads. Thus criteria 1 and 2 are satisfied. In Reference 15.8-5 , Energy Northwest concluded that ATWS equipment had been determined to be qualified by (a) materials analysis of agreeable components including test reports when available, (b) existing qualificat ion to other accident profiles (LOCA, HELB) that encompass the ATWS profile, or (c) location in a mild environment that is not affected by the ATWS acci dent environment. This satisfies criterion 5.
Power Uprate Evaluation The ATWS events were analy zed at power uprate operati ng conditions to demonstrate protection and mitigation of the consequences of these events. These analyses were performed at 3629 MWt power level and bound operation at uprate power level of 3486 MWt. The selection of critical even ts which were analyzed were guided by Reference 15.8-3.
For power uprate evaluation, it was conservatively assumed that ARI has failed, thus, requiring SLC system injection to achieve reactor shutdown.
The analysis presented herein ar e applicable to application of flow control valve (FCV) or adjustable speed drive to reactor recirculation system (RRC). A summary of ATWS results are shown in Table 15.8-3. The analysis results presente d in this section are based on a representative reload core at the time of the analysis (Cycle 8). The Power Uprate ATWS Evaluation was confirmed for th e introduction of GE14 (Reference 15.8-8).
15.8.1 INADVERTENT CONTROL ROD WITHDRAWAL
This transient is bounded by assumptions in the GE licensing topical reports and the other transients analyzed in this section.
15.8.2 LOSS OF FEEDWATER
15.8.2.1 Identification of Causes and Frequency Classification
15.8.2.1.1 Identification of Causes Section 15.2.7 provides identification of causes for loss of feedwater event. The loss of feedwater event with failure to scram will initiate an ATWS event.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-07-011 15.8-3 15.8.2.1.2 Frequenc y Classification
This event is of extremely low probability and is categorized as a limiting fault.
15.8.2.2 Sequence of Events and System Operation
15.8.2.2.1 Sequence of Events
Table 15.8-4 lists the sequence of events for Figure 15.8-1.
15.8.2.2.1.1 Identification of Operator Actions. For the simulation purpose, the following operator actions have been assumed.
- a. Allow automatic operation of the HPCS and reactor core isolation cooling (RCIC),
- b. Begin boron injection at two minutes fo llowing ATWS high-pressure trip or at boron injection initiation temperature (BIIT), whichever is later, and
- c. Switch residual heat removal (RHR) to suppression pool cooling mode 11 minutes following initiation of the transient.
The emergency operating procedures provide operator actions for an ATWS event.
15.8.2.2.2 System Operation
For the loss of feedwater ATWS ev ent, a complete failure to sc ram is postulated to occur for all reactor protection system (R PS) scram signals. All other pl ant control systems maintain normal operation. The relief valves are all assumed to function at the specified setpoints. Loss of feedwater flow results in a proportional reduction of vessel inventory causing the vessel water level to drop. The first correctiv e action is the initiation of HPCS and RCIC on Level 2. For this event, a complete failure of ARI is postula ted. The operator must manually initiate SLC system to inject boron into the reactor vessel for reactor shutdown.
15.8.2.2.3 The Effect of Single Failure and Operator Errors
This ATWS event is based on the assumed complete failure of all control r ods to scram. This is a multiple equipment failure. For the conservative assumption of failure of the ARI system, the ATWS event is terminated by boron injec tion through operator activation of the SLC system. This event is less limiting compared with other ATWS events analyzed at power uprate condition.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-09-007 15.8-4 15.8.2.3 Core and Sy stem Performance
15.8.2.3.1 Mathematical Model
Reference 15.8-7 describes the generic ev aluation methodology for the ATWS event evaluated at uprated power conditions. Additional plant sp ecific analyses were pe rformed for a bounding 10% power uprate using the same methodology.
15.8.2.3.2 Input Paramete rs and Initial Conditions
The initial operating conditions and equipment performance characteristics are given in Tables 15.8-1 and 15.8-2, respectively. MSIV closure occurs on low-low-low water level (L1) but is analyzed based on low-low water level (L2
), conservatively overp redicting suppression pool heatup. The HPCS/RCIC fl ow rates are conservatively high and water level setpoints represent nominal values. The ATWS high pressu re setpoint was set at the upper analytical limit. The SRV setpoints were set using a statistical spread of the analytical setpoint limits for the first opening of each value and reset to a statistical spread of the no minal setpoints for all remaining SRV openings during the transient event.
15.8.2.3.3 Results
The results of this ATWS ev ent simulation are shown in Figure 15.8-1. Feedwater pump trip is assumed to occur at the onset of the event.
Upon the loss of the feedwater flow, reactor pressure, water level, and neutron flux begin to fall. Once reactor water level reaches low-low water level (L2), the protection system trips th e recirculation pumps, in itiates HPCS and RCIC and signals closure of main steam line isolation valves (MSIVs).
Reactor pressure begins to rise due to closure of MSIVs. The relief valves begin to open due to reactor pressure increase. It is conservatively assumed the operator manually initiates SLC system 2 minutes after the ATWS setpoint has been reached.
15.8.2.3.4 Consideration of Uncertainties
Uncertainties in these an alyses involve protection system setpoints, system capacities, and system response times. For ATWS transient analyses, best estim ated values are used when possible. Examples of conser vative bounding values which were used to cover uncertainties are as follows:
- a. For conservatism, the analysis assumed the highest probable ATWS high-pressure trip setpoint, and
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 15.8-5 15.8.2.4 Barrier Performance The calculated peak vessel bottom head pressure is 1202 psig, which is below the American Society of Mechanical Engin eers (ASME) Code Limit of 1375 psig for the RCPB and well below the ASME service level C of 1500 psig. The consequences of this event do not result in any temperature or pressure transient in excess of the criteria for which the fuel, pressure vessel, or containment are designed. Therefore, barrier integrity and function is maintained.
15.8.2.5 Radiological Consequences
While this event does not result in fuel failure it does result in the discharge of normal coolant activity to the suppression pool by means of SRV operation. Since this ac tivity is contained in the primary containment, there will be no uncontrolled release to the environment.
15.8.3 LOSS OF ALTERNATE CURRENT POWER
This transient is bounded by the other tran sients analyzed in this section.
15.8.4 LOSS OF ELECTRICAL LOAD
This transient is bounded by assumptions in the GE licensing topical reports and the other transients analyzed in this section.
15.8.5 LOSS OF CONDENSER VACUUM
This transient is bounded by assumptions in the GE licensing topical reports and the other transients analyzed in this section.
15.8.6 TURBINE TRIP
This event was analyzed at pre-power uprate condition for low power and full power (corresponding to 3323 MWt) ope ration. At uprated conditions, the event is bounded by the other transients analyzed in th is section. The selection of critical events were guided by Reference 15.8-3.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 15.8-6 15.8.7 CLOSURE OF MAIN STEAM LINE ISOLATION VALVES
15.8.7.1 Identification of Causes and Frequency Classification
15.8.7.1.1 Identification of Causes
Various steam line and nuclear system malfunctions, or operator actions, can initiate MSIV closure. These are detailed in Section 15.2.4. The MSIV closure even t with failure to scram will initiate an ATWS event.
15.8.7.1.2 Frequenc y Classification
This event is of extremely low probability and is categorized as a limiting fault.
15.8.7.2 Sequence of Events and System Operation
15.8.7.2.1 Sequence of Events
Table 15.8-5 lists the sequence of events for Figure 15.8-2.
15.8.7.2.1.1 Identification of Operator Actions. For the simulation purpose, the following operator actions have been assumed:
- b. Begin boron injection at 2 minutes following ATWS high-pressure trip or at BIIT, whichever is later, and
Emergency Operating Procedures provide operator actions for an ATWS event.
15.8.7.2.2 System Operation
For the MSIV closure ATWS event, a complete failure to scram is postulated to occur for all RPS scram signals. All other plant control systems maintain normal operation. The relief valves are all assumed to functi on at the specified setpoints.
The RPT occurs at the ATWS high pressure trip setpoint. For this event, a complete failure of AR I is postulated. The operator must manually initiate SLC system to inject boron into the reactor vessel for reactor shutdown.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-07-011 15.8-7 15.8.7.2.3 The Effect of Single Failures and Operator Errors
For the conservative assumption of failure of ARI system, the ATWS event is terminated by boron injection through operator activation of the SLC system. Relief valves operate to limit system pressure. All of these aspects are designed to single failure criterion. Two sensitivity analysis were performed. One to determine the impact of four SRVs inoperable and the other to determine the impact of delay in SLC system injection time.
15.8.7.3 Core and Sy stem Performance
15.8.7.3.1 Mathematical Model
Reference 15.8-7 describes the generic ev aluation methodology for the ATWS event evaluated at uprated power conditions. Additional plant sp ecific analyses were pe rformed for a bounding 10% power uprate using the same methodology.
15.8.7.3.2 Input Paramete rs and Initial Conditions
The initial operating conditions and equipment performance characteristics are given in Tables 15.8-1 and 15.8-2, respectively. The HPCS/RCIC fl ow rates are conservatively high and water level setpoints represent nominal values. The ATWS high pressure setpoint was set at the upper analytical limit. The SRV setpoint s were set using a stat istical spread of the analytical setpoint limits for the first opening of each value and reset to a statistical spread of the nominal setpoints for all remaining SRV openings during the transient event.
15.8.7.3.3 Results
The results of this ATWS ev ent simulation are shown in Figure 15.8-2. The MSIVs close within a nominal 4 sec stroke time. Once the MSIVs reach th e 85% open position, a reactor scram is initiated. The scram was assumed to fail to insert any control rods. The rapid increase in reactor pressure generates rapid incr ease in reactor core power due to collapsing core voids. The relief valves begin to open responding to reactor pressure rise. Upon reaching the ATWS high-pressure setpoint, the RPT occurs and reduces core power. It is conservatively assumed the operator manually initiates SLC system 2 minutes after the ATWS setpoint has been reached.
The MSIV closure (MSIVC) ATWS event was analyzed with four SRVs inoperable for maximum reactor vessel pressure determination. The MSIVC and PREGO event described in Section 15.8.9 were selected based upon previous ATWS analyses that indicated these two events are most limiting with re spect to vessel pressure. Th e results of an ATWS event simulation with four SRVs inoperable are shown in Figure 15.8-3. The sequence of events for this event are shown in Table 15.8-6. The peak calculated vesse l bottom head pressure is 1467 psig, which is below the ASME Service Level C limit of 1500 psig.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-07-011 15.8-8 The calculated peak suppression pool temperature of the event with delayed SLC system injection time is less than 175
°F, which is below the containment design limit. The calculated peak containment pressure for this event is 8.46 psig, which is also below the containment design limit.
15.8.7.3.4 Consideration of Uncertainties
Uncertainties in these an alyses involve protection system setpoints, system capacities, and system response times. For ATWS transient analyses, best estim ated values are used when possible. Examples of conser vative bounding values which were used to cover uncertainties are as follows:
- a. For conservatism, the analyses assumed the highest probable ATWS high-pressure trip setpoint, and
- c. A 35 sec opening time for the suction va lues upstream to the SLCS pumps was modeled increasing the SLCS initiation time.
15.8.7.4 Barrier Performance
The calculated peak vessel bottom head pressure is 1310 psig and 1467 psig for MSIV with full complement of SRVs and with four SRVs inoperable respectively.
The calculated peak vessel bottom head pressure is below the ASME Service Level C limit of 1500 psig, thus, meeting criterion 1. The cons equences of this event do not result in any temperature or pressure transient in excess of the criteria for which the fuel, pressure vessel, or containments are designed. Therefore, these barrier integrity and function is maintained.
15.8.7.5 Radiological Consequences
While this event does not result in fuel failure it does result in the discharge of normal coolant activity to the suppression pool by means of SRV operation. Since this ac tivity is contained in the primary containment, there will be no uncontrolled release to the environment.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-09-007 15.8-9 15.8.8 INADVERTENT OPENING OF RELIEF VALVE
15.8.8.1 Identification of Causes and Frequency Classification
15.8.8.1.1 Identification of Causes
This event assumes that a SRV may "open" a nd stick in the "open" position. These events are detailed in Section 15.1.4. The inadvertent opening of relief valve (IORV) event with failure to scram will initiate an ATWS event.
15.8.8.1.2 Frequenc y Classification
This event is of extremely low probability and is categorized as a limiting fault.
15.8.8.2 Sequence of Events and System Operation
15.8.8.2.1 Sequence of Events
The analysis has not been update d for the change in MSIV isola tion setpoint from Level 2 to Level 1 because the analysis is bounding and conclusions of th e analysis are not affected (Reference 15.8-9).
Table 15.8-7 lists the sequence of events for Figure 15.8-4.
15.8.8.2.1.1 Identification of Operator Actions. For the si mulation purpose, the following operator actions have been assumed:
- a. Initiate boron injection 2 minutes after BIIT,
- b. Disable HPCS, RCIC, and low level MSIV closure, c. Use feedwater to manually control the water level at the top of active fuel, and d. Manually trip the recirculation pumps.
15.8.8.2.1.2 System Operation. For the IORV ATWS event, a complete failure to scram is postulated to occur for all RPS scram signals. All other plant control systems maintain normal operation. For this event, a complete failure of ARI is also postulate
- d. The operator must manually initiate SLC system to inject boron into the reactor vessel for reactor shutdown.
15.8.8.2.2 The Effect of Single Failures and Operator Errors
For the conservative assumption of failure of ARI system, the ATWS event is terminated by boron injection through operator activation of the SLC system. This is a multiple equipment failure event. All of these aspects are designed to single failure criterion.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.8-10 The instrumentation, which detects and audibly alarms the resulting suppression pool temperature rise, and the RHR containment heat removal system are designed to meet the single failure criteria. Th e operator must, however, manua lly initiate suppression pool cooling.
15.8.8.3 Core and Sy stem Performance
15.8.8.3.1 Mathematical Model
Reference 15.8-7 describes the generic ev aluation methodology for the ATWS event evaluated at uprated power conditions. Additional plant sp ecific analyses were pe rformed for a bounding 10% power uprate using the same methodology.
15.8.8.3.2 Input Paramete rs and Initial Conditions
The initial operating conditions and equipment performance characteristics are given in Tables 15.8-1 and 15.8-2, respectively. The HPCS/RCIC fl ow rates are conservatively high and water level setpoints represent nominal values. The ATWS high pressure setpoint was set at the upper analytical limit. The SRV setpoint s were set using a stat istical spread of the analytical setpoint limits for the first opening of each value and reset to a statistical spread of the nominal setpoints for all remaining SRV openings during the transient event.
15.8.8.3.3 Results
The results of this ATWS ev ent simulation are shown in Figure 15.8-4. The opening of a SRV allow steam to be discharged into the suppression pool. Th e sudden increase in the rate of steam flow leaving the reactor vessel causes a mild depressurization transient.
Discharge of steam into the s uppression pool increases the suppression pool temperature. The operator initiates SLC system 2 minutes after the suppression pool temperature reaches 110 F, trips the recirculation pumps, and initiates feedwa ter runback to lower the reactor water level to top of active fuel (TAF). Suppression pool cooling begins 11 minutes after the initiation of the event. The operator disables HPCS and RCIC level 2 initiation.
The MSIV Level 2 closure is also disabled. Turb ine steam flow is terminated u pon closure of the MSIVs due to low steam line pressure.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.8-11 15.8.8.3.4 Consideration of Uncertainties
Uncertainties in these an alyses involve protection system setpoints, system capacities, and system response times. For ATWS transient analyses, best estim ated values are used when possible. Examples of conser vative bounding values which were used to cover uncertainties are as follows:
- a. For conservatism, the analysis assumed the highest probable ATWS high-pressure trip setpoint, and
15.8.8.4 Barrier Performance
The IORV ATWS event is a mild depressurization which has no significant effect on RCPB.
During the event, the suppression pool is continually heated due to SRV discharge. The peak suppression pool temperature and pr essure are within the design criteria of the containment.
15.8.8.5 Radiological Consequences
While this event does not result in fuel failure, it does result in the discharge of normal coolant activity to the suppression pool by means of SRV operation. Since this ac tivity is contained in the primary containment, there will be no uncontrolled release to the environment.
15.8.9 PRESSURE REGULATOR FAILURE - OPEN (PREGO)
15.8.9.1 Identification of Causes and Frequency Classification
15.8.9.1.1 Identification of Causes
The causes for this event is detailed in Section 15.1.3. The PREGO event with failure to scram will initiate an ATWS event.
15.8.9.1.2 Frequenc y Classification This event is of extremely low probability and is categorized as a limiting fault.
15.8.9.2 Sequence of Events and System Operation
15.8.9.2.1 Sequence of Events
Table 15.8-8 lists the sequence of events for Figure 15.8-5.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.8-12 15.8.9.2.1.1 Identification of Operator Actions. For the si mulation purpose, the following operator actions have been assumed.
- b. Begin boron injection at 2 minutes following ATWS high-pressure trip or at BIIT, whichever is later, and
The emergency operating procedures provide operator actions for an ATWS event.
15.8.9.2.1.2 System Operation. For the PREGO ATWS event, a complete failure to scram is postulated to occur for all RPS scram signals. All other plant control systems maintain normal operation. For this event, a complete failure of ARI is also postulate
- d. The operator must manually initiate SLC system to inject boron into the reactor vessel for reactor shutdown.
15.8.9.2.3 The Effect of Single Failures and Operator Errors
For the conservative assumption of failure of ARI system, the ATWS event is terminated by boron injection through operator activation of the SLC system. This is a multiple equipment failure event. All of these aspects are designed to single failure criterion.
The instrumentation, which detects and audibly alarms the resulting suppression pool temperature rise, and the RHR containment heat removal system are designed to meet the single failure criteria.
15.8.9.3 Core and Sy stem Performance
15.8.9.3.1 Mathematical Model
Reference 15.8-7 describes the generic ev aluation methodology for the ATWS event evaluated at uprated power conditions. Additional plant sp ecific analyses were pe rformed for a bounding 10% power uprate using the same methodology.
15.8.9.3.2 Input Paramete rs and Initial Conditions
The initial operating conditions and equipment performance characteristics are given in Tables 15.8-1 and 15.8-2, respectively. The HPCS/RCIC fl ow rates are conservatively high and water level setpoints represent nominal values. The ATWS high pressure setpoint was set at the upper analytical limit. The SRV setpoint s were set using a stat istical spread of the C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.8-13 analytical setpoint limits for the first opening of each value and reset to a statistical spread of the nominal setpoints for all remaining SRV openings during the transient event.
15.8.9.3.3 Results
The results of this ATWS ev ent simulation are shown in Figure 15.8-5. The DEH control system failure with 130% steam flow demand signal is assumed to occur. Ensuing reactor depressurization results in form ation of voids in the reactor coolant and causes a decrease in reactor power almost immediately. The MSIV closure occurs due to trip signal from low steam line pressure. Reactor pressure rises to the relief setpoints and the recirculation pumps trip on the high pressure ATWS setpoint.
Discharge of steam into the s uppression pool increases the suppression pool temperature. The operator initiates feedwater runb ack to lower the reactor wa ter level to TAF after the suppression pool temperature reaches 110F. The HPCS and RCIC systems are initiated at low reactor water level. The SLC system is manually initiated 2 minutes after the ATWS high pressure setpoint was reached.
The PREGO ATWS event was also analyzed with four SRVs inoperable for maximum reactor vessel pressure determination. The PRE GO and MSIVC event described in Section 15.8.7 were selected based upon previous ATWS analyses that indicated these two events are most limiting with respect to vessel pressure. The analysis performed in Reference 15.8-6 determined the peak vessel pr essure for the PREGO ATWS even t is bounded by the results for the MSIVC event.
15.8.9.3.4 Consideration of Uncertainties
Uncertainties in these an alyses involve protection system setpoints, system capacities, and system response times. For ATWS transient analyses, best estim ated values are used when possible. Examples of conser vative bounding values which were used to cover uncertainties are as follows:
- a. For conservatism, the analysis assumed the highest probable ATWS high-pressure trip setpoint, and
15.8.9.4 Barrier Performance The calculated peak vessel bo ttom head pressure is 1306.5 psig , which is below ASME Code limit of 1375 psig for the RCPB and well below the ASME Se rvice Level C of 1500 psig. The consequences of this event do not result in any temperature or pressure transient in excess C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 15.8-14 of the criteria for which the fuel pressure vessel or containment are designed. Therefore, barrier integrity and function is maintained.
The calculated peak suppression pool temperature for this even t is less than 173 F, which is below the containment design limit. The calculated peak containmen t pressure for this event is 8.10 psig, which is also belo w the containment design limit.
15.8.9.5 Radiological Consequences
While this event does not result in fuel failure, it does result in the discharge of normal coolant activity to the suppression pool by means of SRV operation. Since this ac tivity is contained in the primary containment, there will be no uncontrolled release to the environment.
15.8.10 SINGLE REACTOR RECIRCULATION SYSTEM PUMP OPERATION
The following discussion is based on pre-uprate power level of 3323 MWt. Thus, the 100%
rod line corresponds to 3323 MWt power at rated core flow.
For pre-uprate condition, it wa s shown that operation of th e plant with only a single RRC pump and the resulting transient conditions whic h could occur while in this mode are bounded by the other transients analyzed in this section, and the pa rametric studies performed in Reference 15.8-4. This conclusion was based on eval uating the effects of power, void reactivity worth and doppler wo rth at both 100% conditions, and at conditions present under single RRC pum p operation.
Sensitivity studies presented in Reference 15.8-4 compare the turbine trip at 100% power condition with the turbine trip at lower power conditions, such as one would have under single RRC pump.
The rod line for single RRC pump operation for pre-uprate condition was normally maintained between 100% and 104.25% power le vel. At less than 100% power the average void in the core was slightly higher for operation on the 104.25% rod line than for operation on the 100%
rod line. In addition, the doppler worth at lower power conditions is higher than at 100%
power. The presence of higher voids and the in creased doppler worth when operating at the 104.15% rod line is bounded by the para metric analyses in Reference 15.8-4. These parametric analyses determined the sensitivity of plant response between the MSIV closure at 100% power and the MSIV closure with higher reactivity coefficients at 100% power. The void worth assumed in the higher reactivity coefficient case gives a much higher effect than the increased average void present in the single RRC pump ope ration mode at the 104.25% rod line, which bounds this case.
The doppler reactivity worth used in the MSIV closure with higher reactivity coefficients is representative of the doppler reactivity worth found at lower power conditions such as those present in the single RRC pump operation mode.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-09-007,10-029 15.8-15 15.8.11 EXTENDED LOAD LINE LIMIT ANALYSIS OPERATION
Power uprate ATWS analysis were performed with Extended Load Line Limit Analysis (ELLLA) operating conditions. These analyses show that performance at the power uprate condition is within vessel maximum pressure, fuel temperature, and containment pressure limit for the most severe ATWS transients (Reference 15.8-6).
15.8.12 REFERENCES
15.8-1 Hatch Unit 1 FSAR. Amendment 10, Appendix L, "Failure-to-Scram Analysis," Oct ober 27, 1971.
15.8-2 Michelotti, L. A., "A nalysis of Anticipated Tr ansients Without Scram,"
15.8-3 NEDE-24222, "Assessment of BWR Mitigation of ATWS, Volume II (NUREG-0460 Alternate No. 3)."
15.8-4 EI International, Inc., "Final Re port, Anticipated Transients Without Scram Analysis for the WNP-2 Nuclear Power Plant," SA-JAD-087-90, December 1989.
15.8-5 Supply System Letter G02-90-116, G. C.
Sorensen (Supply System) to NRC, "Nuclear Plant No. 2. Operating License (NPF-21 Resolution of Anticipated Transient Without Scram (ATWS) fo r WNP- 2," dated June 29, 1990.
15.8-6 GE Nuclear Energy, "WNP-2 Powe r Uprate Project NSSS Engineering Report," GE-NE-208-17-0993, Re vision 1, December 1994.
15.8-7 GE Nuclear Energy, "Generic Eval uations of General Electric Boiling Water Reactor Power Uprate," Licensing Topical Report NEDC-31984P, July 1991 and Supplements 1 & 2.
15.8-8 GEH-0000-0075-4920, "G E14 Fuel Design Cycle-I ndependent Analyses for Energy Northwest Columbia Generati ng Station" (most recent version referenced in the COLR).
15.8-9 GE Hitachi Nuclear Energy, "Li cense Amendment Request for Proposed Changes to Columbia Technical Spec ifications: Changing Group 1 Isolation Valves' Low Reactor Water Level Isolation Signal from the Current Level 2 to Level 1," 0000-0081-6730-R1, July 2008.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-07-011 15.8-17 Table 15.8-1
Anticipated Transients Without Scram Analysis Initial Conditions
Parameters Value Reactor dome pressure (psig) 1020 Vessel core flow (Mlb/hr) 108.5 Vessel steam flow (Mlb/hr) 15.728 Reactor thermal power (MWt) 3629 Initial vessel and recirculation piping inventory (lbm) 609,600 Narrow range sensed initial water level (ft above separator skirt) 4.13 Initial core average void fraction (%)
41.8 Void reactivity coefficient (¢/%)
-12.937 Doppler coefficient (%/F)
-0.31087 Feedwater enthalpy (Btu/lb) 403.1 Sodium penetaborate solution concentration (% by weight) 13.6 Suppression pool liquid volume (ft
- 3) 112,197 Suppression pool temperature (°F) 90 Service water temperature (°F) 90 C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 15.8-18 Table 15.8-2
Anticipated Transients Without Scram Analysis Equipment Performance Characteristics
Parameter Value Main steam line isolation valve nominal closure time (sec) 4 Relief valve system capacity (% of current NBR steam flow at 1144 psia) 100.18 Number of SRVs 18 Relief valve and sensor time delay (sec)
0.4 Relief
valve opening time (sec) 0.15 Relief valve closure time delay (sec)
0.3 Standby
liquid control system injection rate (gpm) 86.0 High-pressure core spray/RCIC low water level initiation nominal setpoint (ft above separator skirt)
-3.04 (L2)
High-pressure core spray/RCIC high wa ter level shutoff setpoint (ft above separator skirt) 5.667 (L8)
High-pressure core spray flow rate (gpm at 1035 psia) 3875 Reactor core isolation c ooling flow rate (gpm) 600 Anticipated transients without scram high pressure UAL setpoint (psia) 1186 Anticipated transients without scram dome pressure sensor and logic time delay (sec) 0.53 Total bypass capacity (Mlb/hr) 35.5 Total bypass capacity (% of uprate steam flow) 22.7 Pump inertia cons tant (sec) 5.4729 Residual heat removal pool cooling capacity (Btu/sec-
°F) 578 C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 15.8-19 Table 15.8-3 Summary of Anticipated Transients Without Scram Results ATWSEvent Parameter PREGO MSIVC LOFWIORV Maximum neutron flux (%)
Time (sec) 421.22 22.59 683.2 4.02 282.07 22.36 116.6 7.96 Maximum average fuel heat flux (%)
Time (sec) 170.99 24.41 151.83 5.11 102.9 0.49 103.14 0.69 Maximum bottom pressure (psig)
Time (sec) 1306.5 27.62 1310.1 8.28 1202.2 23.62 1061.4 0.19 Peak suppression pool temperature (°F) 172.79 173.40 161.06 165.29 Peak containment pressure (psig)
Time (sec) 8.10 4600 8.20 4500 5.97 8400 6.87 6600 Peak cladding temperature (°F) Time (sec) 1473.67 83.5 N/A N/A N/A Min. water level (ft above sep. skirt)
Time (sec)
-10.47 211.18 -10.27 197.29 -11.24 91.38 -10.66 975.4 Time of hot shutdown a (sec) 984.6 962.6 977.0 1524.6 Time of reaching ATWS setpoint (sec) 24.09 4.73 17.5 N/A Time of BIIT (sec) 67.9 54.6170.0554.0 a Hot shutdown is defined as generated power remaining below 1% NBR.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-09-007 15.8-20 Table 15.8-4
Sequence of Events for Loss of Feedwater
Time Event 0 sec Feedwater pump trip. 17.5 sec High-pressure core spray and RCIC initiated on Level 2. 17.5 sec Main steam line isolatio n valve closure on Level 2 (see Section 15.8.2.3.2) - scram fails. 17.5 sec Recirculation pump tripped on Level 2 (ATWS setpoint reached, ARI fails). 22.9 sec Relief valves lift.
23.6 sec Vessel pressure peaks.
2 minutes 18 sec Operator initiates SLCS (2 minutes after ATWS setpoint reached). 3 minutes 3 sec Liquid contro l flow enters the core. 16 minutes Hot shutdown achieved.
140 minutes Suppression pool temperatur e and containment pressure peak.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-07-011 15.8-21 Table 15.8-5
Sequence of Events for Main Steam Line Isolation Valve Closure (Long Term Transient)
Time Event 0 sec Nominal 4-sec MSIV closure - scram fails. 4.37 sec Relief valves lift.
4.73 sec Recirculation pump trip on high pressure (ATWS setpoint reached, ARI fails). 8.28 sec Vessel pressure peaks.
54.60 sec Operator initiates feedwater runback (suppression pool at 110
°F). 1 minute 15 sec High-pressure core spray and RCIC initiated on Level 2. 2 minutes 4 sec Operator initiates SLC system 2 minutes after ATWS setpoint reached. 2 minutes 51 sec Liquid control flow enters the core.
11 minutes Suppression pool cooling begins. 16 minutes Hot shutdown achieved.
75 minutes Suppression pool temperatur e and containment pressure peak.
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 15.8-22 Table 15.8-6
Sequence of Events for Main Steam Line Isolation Valve Closure with Four Safety/Relief Valves Out-Of-Service
Time (sec) Event 0 Nominal 4-sec MSIV closure - scram fails. 4.47 Relief valves lift. 4.73 Recirculation pump trip on high pressure (ATWS setpoint reached). 12.30 Vessel pressure peaks.
C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-09-007 15.8-23 Table 15.8-7
Sequence of Events for Inadvertent Open Relief Valve
Time Event 0 sec Relief valve with the lowest opening se tpoint opens. 9 minutes 14 sec Operator initiates SL CS 2 minutes afte r suppression pool temperature = 110F (scram and ARI fail). 9 minutes 14 sec Operator trips recirculation pumps. 9 minutes 14 sec Operator initiates feedwa ter runback to bring level to TAF. 11 minutes Suppression pool cooling begins. 13 minutes Operator disables HPCS, RCIC Level 2 initiation and MSIV Level 2 closure a. 14 minutes Liquid control flow enters the core. 25 minutes Hot shutdown achieved. 29 minutes Main steam line isolatio n valve closure on low pressure. 110 minutes Suppression pool temperatur e and containment pressure peak.
a The analysis has not been updated for the change in MSIV isolation setpoint from Level 2 to Level 1 because the analysis is bounding and conclusions of th e analysis are not affected (Reference 15.8-9).
C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 15.8-24 Table 15.8-8
Sequence of Events for Pressure Regulator Failure Open (Long Term Transient)
Time Event 0 sec Pressure regulator fails to maximum demand.
15.3 sec Main steam line isolation valve closure on low steam line pressure - scram fails.
23.7 sec Relief valves lift.
24.1 sec Recirculation pump trip on high pressure (ATWS setpoint reached, ARI fails).
27.6 sec Vessel pressure peaks.
1 minute 8 sec Operator initiates feedwater r unback (Suppression pool at 110
°F). 1 minute 30 sec High-pressure core spray and RCIC initiated on Level 2. 2 minutes 24 sec Operator initiates SLC system 2 minutes after ATWS setpoint is reached. 3 minutes 9 sec Liquid control flow enters the core. 11 minutes Suppression pool cooling begins. 16 minutes Hot shutdown achieved.
77 minutes Suppression pool temperatur e and containment pressure peak.
Amendment 53 November 1998 Columbia Generating Station Final Safety Analysis Report Loss of Feedwater Event 020002.29 15.8-1.1 Amendment 53 November 1998 Columbia Generating Station Final Safety Analysis Report Loss of Feedwater Event 020002.30 15.8-1.2 Amendment 53 November 1998 Columbia Generating Station Final Safety Analysis Report Loss of Feedwater Event 020002.31 15.8-1.3 Amendment 53 November 1998 Loss of Feedwater Event Columbia Generating Station Final Safety Analysis Report 020002.32 15.8-1.4 Amendment 53 November 1998 Columbia Generating Station Final Safety Analysis Report Loss of Feedwater Event 020002.33 15.8-1.5 Amendment 53 November 1998 Columbia Generating Station Final Safety Analysis Report Main Steam Isolation Valve Closure Event 020002.34 15.8-2.1 Amendment 53 November 1998 Columbia Generating Station Final Safety Analysis Report Main Steam Isolation Valve Closure Event 020002.35 15.8-2.2 Amendment 53 November 1998 Columbia Generating Station Final Safety Analysis Report Main Steam Isolation Valve Closure Event 020002.36 15.8-2.3 Amendment 53 November 1998 Columbia Generating Station Final Safety Analysis Report Main Steam Isolation Valve Closure Event 020002.37 15.8-2.4 Amendment 53 November 1998 Columbia Generating Station Final Safety Analysis Report Main Steam Isolation Valve Closure Event 020002.38 15.8-2.5 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Main Steam Isolation Valve Closure Event with4 SRVs Out-of-Service 960222.40 15.8-3.1 0 10 20 30 40 0 50 100 150Time, (Seconds)
Flow (% Rated) 200 HPCS Flow % of Rated FWRV Flow in %
Columbia Generating StationFinal Safety Analysis Report Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Main Steam Isolation Valve Closure Event with4 SRVs Out-of-Service 960222.41 15.8-3.2 0 10 20 30 40-20-10 0 10Time, (Seconds)
Level , (Above Sep Skirt) 20 Actual Level in Feet WR Sensed Level in Feet Columbia Generating StationFinal Safety Analysis Report Main Steam Isolation Valve Closure Event with 4SRVs Out-of-Service 960222.42 15.8-3.3 0 10 20 30 40 0.8 1.0 1.2 1.4 x10 3Time, (Seconds)
Dome Pressure (psia) 1.6 Dome Pressure in psia REDYV 50 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Columbia Generating StationFinal Safety Analysis Report Figure Form No. 960690 LDCN-07-011 Draw. No.Rev.Main Steam Isolation Valve Closure Event with 4 SRVs Out-of-Service 960222.43 15.8-3.4 Columbia Generating Station Final Safety Analysis Report Amendment 59 December 2007 0 10 20 30 40-3-1 1 3Time, (Seconds)
Reactivity ($)
5Void in $Doppler in $
Net Reactivity in $
REDYV 50 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Main Steam Isolation Valve Closure Event with 4SRVs Out-of-Service 960222.44 15.8-3.5 0 10 20 30 40 0 50 100 150Time, (Seconds)
Flux (% Rated) 200 Neutron Flux in %Avg Surf Heat Flux in %
REDYV 50 Columbia Generating StationFinal Safety Analysis Report Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Inadvertent Opening of Relief Valve Event 960222.45 15.8-4.1 0 1 2 3 4x10 3-20-10 0 10Time, (Seconds)
Level (Above Sep Skirt) 20 Actual Level in Feet WR Sensed Level in Feet REDYV 5 Columbia Generating StationFinal Safety Analysis Report Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Inadvertent Opening of Relief Valve Event 960222.48 15.8-4.2 0 1 2 3 4x10 3-3-1 1 3Time, (Seconds)
Reactivity ($)
5Void in $Doppler in $
Net Reactivity in $
REDYV 5 4 2 0-2 Columbia Generating StationFinal Safety Analysis Report Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Inadvertent Opening of Relief Valve Event 960222.46 15.8-4.3 0 1 2 3 4x10 3 0 50 100 150Time, (Seconds)
Reactor Vessel Flow (% Rated) 200 REDYV 5 Columbia Generating StationFinal Safety Analysis Report Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Inadvertent Opening of Relief Valve Event 960222.47 15.8-4.4 0 1 2 3 4x10 3 0.8 1.0 1.2 1.4Time, (Seconds)
Dome Pressure (psia)Dome Pressure in psia REDYV 5 1.6x10 3 Columbia Generating StationFinal Safety Analysis Report Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Inadvertent Opening of Relief Valve Event 960222.49 15.8-4.5 0 1 2 3 4x10 3 0 40 80 120Time, (Seconds)
Flux (% Rated) 160 Neutron Flux in %Avg. Surf Heat Flux in %
REDYV 5 Columbia Generating StationFinal Safety Analysis Report