ML12011A163

Columbia, Amendment 61 to Final Safety Analysis Report, Chapter 9, Auxiliary Systems

ML12011A163
Person / Time
Site:	Columbia
Issue date:	12/14/2011
From:	Energy Northwest
To:	Office of Nuclear Reactor Regulation
References
GO2-11-201
Download: ML12011A163 (374)

Text

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 Chapter 9 AUXILIARY SYSTEMS

TABLE OF CONTENTS

Section Page LDCN-11-000 9-i 9.1 FUEL STORAGE AND HANDLING ................................................. 9.1-1 9.1.1 NEW FUEL STORAGE

................................................................ 9.1-1 9.1.1.1 Desi gn Bases ........................................................................... 9.1-1 9.1.1.1.1 Safety Design Bases ................................................................ 9.

1-1 9.1.1.1.1.1 Safety Design Bases - Structural ............................................... 9.1-1 9.1.1.1.1.2 Safety Desi gn Bases - Nuclear ................................................. 9.1-1 9.1.1.1.2 Power Genera tion Design Ba ses .................................................

9.1-2 9.1.1.2 Facilities Description ................................................................. 9.1-2 9.1.1.3 Safety Evaluation ...................................................................... 9.1-3 9.1.1.3.1 Criticality Control .................................................................. 9.1-3 9.1.1.3.2 New Fuel Rack Structural Design ............................................... 9.1-4 9.1.1.3.3 New Fu el Handling ................................................................. 9.1-5 9.1.1.3.4 Other New Fuel Storage Design Fa ctors .......................................

9.1-5 9.1.2 SPENT FU EL STORAGE

............................................................. 9.1-6 9.1.2.1 Desi gn Bases ........................................................................... 9.1-6 9.1.2.1.1 Safety Design Bases ................................................................ 9.

1-6 9.1.2.1.1.1 Safety Design Bases - Struct ural. ..............................................

9.1-6 9.1.2.1.1.2 Safety Desi gn Bases - Nucl ear. ................................................

9.1-6 9.1.2.1.2 Power Genera tion Design Ba ses .................................................

9.1-7 9.1.2.2 Facilities Description ................................................................. 9.1-7 9.1.2.2.1 Spent Fuel Storage Ra cks .........................................................

9.1-7 9.1.2.2.2 Spent Fuel Storage P ool ...........................................................

9.1-8 9.1.2.3 Safety Evaluation ...................................................................... 9.1-9 9.1.2.3.1 Criticality Control .................................................................. 9.1-9 9.1.2.3.1.1 8 x 8 Fuel.

......................................................................... 9.1-9 9.1.2.3.1.2 9 x 9 and 10 x 10 Fuel .......................................................... 9.1-11 9.1.2.3.2 Spent Fuel Storage Rack Structural Design .................................... 9.1-15 9.1.2.3.3 Spent Fuel a nd Cask Handling ................................................... 9.1-18 9.1.2.3.4 Spent Fuel Rack Design Features ................................................ 9.1-19 9.1.2.3.5 Spent Fuel Storage Facilities De sign ............................................

9.1-21 9.1.2.3.6 Radiological Considerations ...................................................... 9.1-22 9.1.2.3.6.1 Normal Operation. ............................................................... 9.1-22 9.1.2.3.6.2 Radiological Consequences of Accidents

..................................... 9.1-22 9.1.2.3.7 Conc lusions ..........................................................................

9.1-23 9.1.3 SPENT FUEL POOL COOLIN G AND CLEANUP SYSTEM ................. 9.1-23 9.1.3.1 Design Bases ...........................................................................

9.1-23 C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Chapter 9 AUXILIARY SYSTEMS

TABLE OF CONTENTS (Continued)

Section Page LDCN-08-035 9-ii 9.1.3.2 System Description....................................................................9.

1-24 9.1.3.2.1 System Operation...................................................................9.

1-25 9.1.3.2.1.1 FPC System Cooling Function.................................................9.1-25 9.1.3.2.1.2 FPC System Clean-Up Functions..............................................

9.1-27 9.1.3.2.2 Residual Heat Removal/Fuel Pool Cooling Cleanup Assist Mode Operation..............................................................................9.1-29 9.1.3.2.3 Fuel Pool Cooling System Operation Following a Seismic Event or Major Plant Di sturbance.......................................................9.1-30 9.1.3.3 Safety Evaluation......................................................................9.

1-30 9.1.3.4 Testing and Inspection Requirements..............................................9.1-32 9.1.4 FUEL HAND LING SYSTEM........................................................9.1-32 9.1.4.1 Design Bases...........................................................................9.1-32 9.1.4.2 System Description....................................................................9.

1-34 9.1.4.2.1 Spent Fuel Cask.....................................................................9.

1-34 9.1.4.2.2 Reactor Building Crane............................................................9.1-34 9.1.4.2.2.1 Description........................................................................9.

1-34 9.1.4.2.2.2 Safety Features....................................................................9.

1-36 9.1.4.2.3 Fuel Servicing Equipment.........................................................9.1-37 9.1.4.2.3.1 Fuel Preparation Machine......................................................9.1-37 9.1.4.2.3.2 New Fuel Inspection Stand.....................................................9.1-38 9.1.4.2.3.3 Channel Bolt Wrench............................................................9.1-38 9.1.4.2.3.4 Channel Handling Tool..........................................................9.1-38 9.1.4.2.3.5 Fuel Pool Sipper..................................................................9.1-38 9.1.4.2.3.6 Fuel Inspection Fixture..........................................................9.1-39 9.1.4.2.3.7 Channel Gauging Fixture.......................................................9.1-39 9.1.4.2.3.8 Gene ral Purpose Grapple.......................................................9.1-39 9.1.4.2.4 Serv icing Aids.......................................................................9.

1-39 9.1.4.2.5 Reactor Vessel Servicing Equipment............................................

9.1-40 9.1.4.2.5.1 Reactor Vessel Service Tools..................................................9.1-40 9.1.4.2.5.2 Steam Line Plug..................................................................9.1-40 9.1.4.2.5.3 Shroud Head Bolt Wrench......................................................9.1-40 9.1.4.2.5.4 Head Holding Pedestal..........................................................9.1-40 9.1.4.2.5.5 Head Nut and Washer Racks...................................................9.1-41 9.1.4.2.5.6 Head Stud Racks.................................................................9.1-41 9.1.4.2.5.7 Dryer and Separator Slings.....................................................9.1-41 9.1.4.2.5.8 Head Strongback.................................................................9.1-41 C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Chapter 9 AUXILIARY SYSTEMS

TABLE OF CONTENTS (Continued)

Section Page LDCN-08-035,09-001 9-iii 9.1.4.2.5.9 Steam Li ne Plug Installation Tool.............................................9.1-42 9.1.4.2.5.10 Auxiliary Work Platform......................................................9.1-42 9.1.4.2.5.11 Cavity In-Vessel Service Platform (CISP)..................................9.1-42 9.1.4.2.5.12 Head Stud Tensioner Carousel/Strongback.................................9.1-43 9.1.4.2.6 In-Vessel Se rvicing Equipment...................................................

9.1-43 9.1.4.2.7 Refueli ng Equipment...............................................................9.1-44 9.1.4.2.7.1 Refueling Platform...............................................................9.1-44 9.1.4.2.8 Storag e Equipment..................................................................9.1-45 9.1.4.2.9 Under Reactor Vessel Servicing Equipment...................................

9.1-45 9.1.4.2.10 Description of Fuel Transfer....................................................9.1-46 9.1.4.2.10.1 Arrival of Fuel on Site.........................................................9.1-46 9.1.4.2.10.2 Re fueling Procedure............................................................9.1-46 9.1.4.2.10.2.1 New Fuel Preparation.......................................................9.1-48 9.1.4.2.10.2.1.1 Receipt and Inspection of New Fuel....................................9.1-48 9.1.4.2.10.2.1.2 Channeling New Fuel (Non Westinghouse Fuel).....................9.1-49 9.1.4.2.10.2.1.3 Channeling New Fuel (Westinghouse Fuel Only)....................9.1-49 9.1.4.2.10.2.1.4 Equipment Preparation....................................................9.1-49 9.1.4.2.10.2.2 Reactor Shutdown............................................................9.1-49 9.1.4.2.10.2.2.1 Drywell Head Removal...................................................9.1-50 9.1.4.2.10.2.2.2 Reactor Well Servicing...................................................9.1-50 9.1.4.2.10.2.3 Reactor Vessel Opening.....................................................9.1-50 9.1.4.2.10.2.3.1 Vessel Head Removal.....................................................9.1-50 9.1.4.2.10.2.3.2 Dryer Removal.............................................................9.1-51 9.1.4.2.10.2.3.3 Separator Removal........................................................9.1-51 9.1.4.2.10.2.3.4 Fuel Bundle Sampling.....................................................9.1-51 9.1.4.2.10.2.4 Refueling and Reactor Servicing..........................................9.1-51 9.1.4.2.10.2.4.1 Refueling....................................................................9.

1-51 9.1.4.2.10.2.5 Vessel Closure................................................................9.1-52 9.1.4.2.10.3 Departure of Spen t Fuel From the Reactor Building.....................9.1-53 9.1.4.3 Fuel Handling Safety Evaluation...................................................9.1-53 9.1.4.4 Testing and Inspection Requirements..............................................9.1-57 9.1.4.4.1 Testing and Inspection of Cranes................................................

9.1-57 9.1.4.4.2 Inspection a nd Testing Requireme nts of Refueling and Servicing Equipment............................................................................9.1-58 9.1.4.4.2.1 Inspection..........................................................................9.1-58 9.1.4.4.2.2 Testing.............................................................................9.1-58 C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Chapter 9 AUXILIARY SYSTEMS

TABLE OF CONTENTS (Continued)

Section Page LDCN-08-035,09-001, 09-027 9-iv 9.1.4.5 Instrumentation Requirements.......................................................9.1-59 9.1.4.5.1 Instrument Re quirements - Cranes...............................................

9.1-59 9.1.4.5.2 Instrumentation Requirements - Refueling and Servicing Equipment.....9.1-59 9.1.4.5.3 Fuel S upport Grapple..............................................................9.1.60 9.1.4.5.4 Other..................................................................................

9.1-60 9.1.4.5.5 Radiati on Monitoring...............................................................9.

1-60 9.1.5 CONTROL OF HEAVY LOADS....................................................9.1-60 9.1.5.1 Introduction/Licensing Background................................................9.1-60 9.1.5.2 Safety Basis.............................................................................9.1-61 9.1.5.3 Scope of Overhead Heavy Load Handling System..............................9.1-61 9.1.5.4 Control of Heavy Loads Program..................................................9.1-63 9.1.5.4.1 Columbia Generating St ation Commitments in Response to NUREG-0612 Section 5.1.1, Phase I Guidelines.............................9.1-63 9.1.5.4.1.1 Safe Load Path....................................................................9.

1-63 9.1.5.4.1.2 Procedures.........................................................................9.1-63 9.1.5.4.1.3 Crane and Hoist Operators.....................................................9.1-63 9.1.5.4.1.4 Speci al Lifting Devices..........................................................9.1-63 9.1.5.4.1.5 Lifting Devices that are not Specially Designed............................9.1-63 9.1.5.4.1.6 Crane Inspecti on, Testing and Maintenance.................................9.1-64 9.1.5.4.1.7 Hoist and Crane Design.........................................................9.1-64 9.1.5.4.1.7.1 Ho ist Design....................................................................9.

1-64 9.1.5.4.1.7.2 Cr ane Design...................................................................9.

1-64 9.1.5.4.2 Reactor Pressure Vessel Head Lifting Procedures............................9.1-64 9.1.5.4.3 Single-Failure-Proof Cr anes for Spent Fuel Casks............................9.1-64 9.1.5.4.4 Othe r Analyses......................................................................9.

1-64 9.1.5.5 Safety Evaluation......................................................................9.

1-64 9.

1.6 REFERENCES

...........................................................................

9.1-65 9.2 WATER SYSTEMS.......................................................................9.2-1

9.2.1 PLANT

SERVIC E WATER SYSTEM..............................................9.2-1 9.2.1.1 Design Bases...........................................................................9.2-1 9.2.1.2 System Description....................................................................9.2-1 9.2.1.3 Safety Evaluation......................................................................9.2-2 9.2.1.4 Testing and Inspection Requirements..............................................9.2-2 9.2.1.5 Instrumentation Requirements.......................................................9.2-3 C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 Chapter 9 AUXILIARY SYSTEMS

TABLE OF CONTENTS (Continued)

Section Page LDCN-08-032 9-v 9.2.2 REACTOR BUILDING CLOSED COOLING WATER SYSTEM ............ 9.2-3

9.2.2.1 Desi gn Bases ........................................................................... 9.2-3 9.2.2.2 System Description

.................................................................... 9.2-3 9.2.2.3 Safety Evaluation ...................................................................... 9.2-5 9.2.2.4 Testing and Inspection Requirements .............................................. 9.2-6 9.2.2.5 Instrumentation Requirements

....................................................... 9.2-6

9.2.3 PLANT

MAKEUP WATER TREATMENT AND DEMINERALIZED WATER SYSTEMS ..................................................................... 9.

2-6 9.2.3.1 Design Basis ............................................................................ 9.2-6 9.2.3.2 System Description

.................................................................... 9.2-7 9.2.3.2.1 Plant Makeup Water Treatment System ........................................ 9.2-7 9.2.3.2.2 Demineralized Water System ..................................................... 9.2-7 9.2.3.3 Safety Evaluation ...................................................................... 9.2-8 9.2.3.4 Testing and Inspection Requirements .............................................. 9.2-8 9.2.3.5 Instrumentation Requirements

....................................................... 9.2-8 9.2.4 POTABLE WATER AND SAN ITARY DRAIN SYSTEMS .................... 9.2-9 9.2.4.1 Desi gn Bases ........................................................................... 9.2-9 9.2.4.2 System Description

.................................................................... 9.2-9 9.2.4.2.1 Potable Water System .............................................................. 9.2-9 9.2.4.2.2 Sanitary Drain System ............................................................. 9.2-9 9.2.4.3 Safety Evaluation ...................................................................... 9.2-10 9.2.4.4 Testing and Inspection Requirements .............................................. 9.2-10 9.2.4.5 Instrumentation Requirements

....................................................... 9.2-10

9.2.5 ULTIMATE

HEAT SINK ............................................................. 9.2-10

9.2.5.1 Design Bases ...........................................................................

9.2-10 9.2.5.2 System Description

.................................................................... 9.2-11 9.2.5.3 Safety Evaluation ...................................................................... 9.2-12 9.2.5.4 Testing and Inspection Requirements .............................................. 9.2-17 9.2.5.5 Instrumentation Requirements

....................................................... 9.2-17

9.2.6 CONDENSATE

SUPPLY SYSTEM................................................. 9.2-17

9.2.6.1 Design Bases ...........................................................................

9.2-17 9.2.6.2 System Description

.................................................................... 9.2-18 9.2.6.3 Safety Evaluation ...................................................................... 9.2-19 9.2.6.4 Testing and Inspection Requirements .............................................. 9.2-19 9.2.6.5 Instrumentation Requirements

....................................................... 9.2-20 C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Chapter 9

AUXILIARY SYSTEMS

TABLE OF CONTENTS (Continued)

Section Page 9-vi 9.2.7 STANDBY SERVIC E WATER SYSTEM..........................................9.

2-20 9.2.7.1 Design Bases...........................................................................9.2-20 9.2.7.2 System Description....................................................................9.

2-21 9.2.7.3 Safety Evaluation......................................................................9.

2-23 9.2.7.4 Testing and Inspection Requirements..............................................9.2-25 9.2.7.5 Instrumentation Requirements.......................................................9.2-25

9.2.8 COMPRESSOR

JACK ET WATER SYSTEM.....................................9.2-25 9.2.8.1 Design Bases...........................................................................9.2-25 9.2.8.2 System Description....................................................................9.

2-26 9.2.8.3 Safety Evaluation......................................................................9.

2-27 9.2.8.4 Testing and Inspection Requirements..............................................9.2-27 9.2.8.5 Instrumentation Requirements.......................................................9.2-27 9.

2.9 REFERENCES

...........................................................................

9.2-27 9.3 PROCESS AUXILIARIES...............................................................9.3-1

9.3.1 COMPRESSED

AIR SYSTEMS......................................................9.3-1 9.3.1.1 Design Bases...........................................................................9.3-1 9.3.1.1.1 Control and Service Air Systems.................................................9.3-1 9.3.1.1.2 Containment Instrument Air System.............................................9.3-2 9.3.1.2 System Description....................................................................9.3-3 9.3.1.2.1 Control and Service Air System..................................................9.3-3 9.3.1.2.2 Containment Instrument Air System.............................................9.3-4 9.3.1.3 Safety Evaluation......................................................................9.3-6 9.3.1.3.1 Control and Service Air System..................................................9.3-6 9.3.1.3.2 Containment Instrument Air System.............................................9.3-6 9.3.1.4 Testing and Inspection Requirements..............................................9.3-6 9.3.1.5 Instrumentation Requirements.......................................................9.3-7 9.3.1.5.1 Control and Service Air Systems.................................................9.3-7 9.3.1.5.2 Containment Instrument Air System.............................................9.3-7 9.3.2 PROCESS SAMPLING SYSTEM....................................................9.3-8 9.3.2.1 Design Bases...........................................................................9.3-8 9.3.2.2 System Description....................................................................9.3-9 9.3.2.2.1 Sample Locations...................................................................9.

3-10 9.3.2.2.2 Liquid Sample Taps and Probes..................................................

9.3-10 9.3.2.2.3 Steam Samples.......................................................................9.

3-10 9.3.2.2.4 Sample Piping Design..............................................................9.

3-10 C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Chapter 9 AUXILIARY SYSTEMS

TABLE OF CONTENTS (Continued)

Section Page 9-vii 9.3.2.2.5 Fume Hood Desi gn and Grab Samples..........................................9.3-11 9.3.2.3 Safety Evaluation......................................................................9.

3-11 9.3.2.4 Tests and Inspections.................................................................9.3-12 9.3.2.5 Instru ment Requirements.............................................................9.3-12

9.3.3 EQUIPMENT

AND FLOO R DRAINAGE SYSTEMS..........................9.3-13 9.3.3.1 Design Bases...........................................................................9.3-13 9.3.3.2 System Description....................................................................9.

3-13 9.3.3.2.1 Radioactive Equi pment Drainage System.......................................9.3-13 9.3.3.2.1.1 Reactor Building Drains.........................................................9.3-13 9.3.3.2.1.2 Turbine Building Drains........................................................9.3-15 9.3.3.2.1.3 Radw aste Building Drains......................................................9.3-15 9.3.3.2.2 Radioactive Floor Drainage Subsystem.........................................

9.3-16 9.3.3.2.2.1 Reactor Building Floor Drains.................................................9.3-16 9.3.3.2.2.2 Turbine Building Floor Drains.................................................9.3-18 9.3.3.2.2.3 Radwaste Building Floor Drains...............................................9.3-19 9.3.3.2.3 Nonradioactive Water Drainage System........................................9.3-19 9.3.3.2.3.1 Turbine and Service Buildings.................................................9.3-19 9.3.3.2.3.2 Miscella neous Drainage System...............................................9.3-20 9.3.3.3 Safety Evaluation......................................................................9.

3-20 9.3.3.4 Testing and Inspection Requirements..............................................9.3-21 9.3.3.5 Instrumentation Requirements.......................................................9.3-21

9.3.4 CHEMICAL

AND VOL UME CONTROL SYSTEM............................9.3-21

9.3.5 STANDBY

LIQUID CONTROL SYSTEM........................................9.3-21 9.3.5.1 Design Bases...........................................................................9.3-21 9.3.5.2 System Description....................................................................9.

3-22 9.3.5.3 Safety Evaluation......................................................................9.

3-24 9.3.5.4 Testing and Inspection Requirements..............................................9.3-27 9.3.5.5 Instrumentation Requirements.......................................................9.3-27 9.

3.6 REFERENCES

...........................................................................

9.3-27 9.4 HEATING, VENTILATING, AND AIR CONDITIONING SYSTEMS........9.4-1 9.4.1 MAIN CONTROL ROOM/CABLE SPREADING ROOM/CRITICAL SWITCHGEAR AREA.................................................................9.4-2 9.4.1.1 Design Bases...........................................................................9.4-2 C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Chapter 9 AUXILIARY SYSTEMS

TABLE OF CONTENTS (Continued)

Section Page LDCN-09-028 9-viii 9.4.1.2 System Description....................................................................9.4-4 9.4.1.2.1 Main Control Room................................................................9.

4-4 9.4.1.2.2 Cable Sp reading Room.............................................................9.

4-7 9.4.1.2.3 Critical Switchgear Area..........................................................9.

4-7 9.4.1.3 Safety Evaluation......................................................................9.4-8 9.4.1.3.1 Main Control Room................................................................9.

4-8 9.4.1.3.2 Cable Sp reading Room.............................................................9.

4-8 9.4.1.3.3 Critical Switchgear Area..........................................................9.

4-9 9.4.1.4 Testing and Inspection Requirements..............................................9.4-9 9.4.1.5 Instrumentation Requirements.......................................................9.4-9 9.4.1.5.1 Main Control Room................................................................9.

4-9 9.4.1.5.2 Cable Sp reading Room.............................................................9.

4-10 9.4.1.5.3 Critical Switchgear Area..........................................................9.

4-11 9.4.2 REACTOR BUILDING................................................................9.4-11 9.4.2.1 Design Bases...........................................................................9.4-11 9.4.2.2 System Description....................................................................9.

4-13 9.4.2.2.1 Supply Air System..................................................................9.4-13 9.4.2.2.2 Exhaus t Air System.................................................................9.4-14 9.4.2.2.3 Sump Vent Exhaust Filter System...............................................9.4-14 9.4.2.2.4 Miscellaneous Ar ea Ventilation Systems.......................................9.4-15 9.4.2.3 Safety Evaluation......................................................................9.

4-16 9.4.2.4 Testing and Inspection Requirements..............................................9.4-17 9.4.2.5 Instrumentation Requirements.......................................................9.4-18

9.4.3 RADWASTE

BUILDING..............................................................9.4-21 9.4.3.1 Design Bases...........................................................................9.4-21 9.4.3.2 System Description....................................................................9.

4-21 9.4.3.3 Safety Evaluation......................................................................9.

4-23 9.4.3.4 Testing and Inspection Requirements..............................................9.4-24

9.4.4 RADWASTE

BUILDING CH ILLED WATER SYSTEM.......................9.4-24 9.4.4.1 Design Bases...........................................................................9.4-24 9.4.4.2 System Description....................................................................9.

4-24 9.4.4.3 Safety Evaluation......................................................................9.

4-25 9.4.4.4 Testing and Inspection Requirements..............................................9.4-25 9.4.4.5 Instrumentation Requirements.......................................................9.4-25 9.4.5 OFFGAS CHARCOAL ADSORBER VAULT REFRIGERATION SYSTEM..................................................................................

9.4-25 C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 Chapter 9 AUXILIARY SYSTEMS

TABLE OF CONTENTS (Continued)

Section Page LDCN-10-007 9-ix 9.4.6 TURBINE GENERA TOR BUILDING

.............................................. 9.4-26 9.4.6.1 Design Bases ...........................................................................

9.4-26 9.4.6.2 System Description

.................................................................... 9.4-27 9.4.6.2.1 Main Supply System

................................................................ 9.4-27 9.4.6.2.2 Main Exhaust System .............................................................. 9.4-27 9.4.6.2.3 Auxiliary Boiler Room Ventilation System .................................... 9.4-28 9.4.6.2.4 Transformer Vault Ventilation System .......................................... 9.4-28 9.4.6.2.5 Sample Room Ai r Conditioning System ........................................ 9.4-29 9.4.6.3 Safety Evaluation ...................................................................... 9.4-29 9.4.6.4 Testing and Inspection Requirements .............................................. 9.4-29 9.4.6.5 Instrumentation Requirements

....................................................... 9.4-30 9.4.7 EMERGENCY DIESEL GENERATOR BUILDING

............................. 9.4-30 9.4.7.1 Design Bases ...........................................................................

9.4-30 9.4.7.2 System Description

.................................................................... 9.4-31 9.4.7.3 Safety Evaluation ...................................................................... 9.4-32 9.4.7.4 Testing and Inspection Requirements .............................................. 9.4-32 9.4.7.5 Instrumentation Requirements

....................................................... 9.4-32

9.4.8 DIESEL

GENERATOR AREA CABLE COOLING SYSTEM ................. 9.4-33 9.4.8.1 Design Bases ...........................................................................

9.4-33 9.4.8.2 System Description

.................................................................... 9.4-34 9.4.8.3 Safety Evaluation ...................................................................... 9.4-34 9.4.8.4 Inspection and Te sting Requirements .............................................. 9.4-35 9.4.8.5 Instrumentation Requirements

....................................................... 9.4-35

9.4.9 REACTOR

BUILDING EMERGENCY COOLING SYSTEMS ............... 9.4-36

9.4.9.1 Design Bases ...........................................................................

9.4-36 9.4.9.2 System Description

.................................................................... 9.4-37 9.4.9.3 Safety Evaluation ...................................................................... 9.4-37 9.4.9.4 Testing and Inspection Requirements .............................................. 9.4-37 9.4.9.5 Instrumentation Requirements

....................................................... 9.4-37 9.4.10 STANDBY SERVICE WATER PUMP HOUSE ................................. 9.4-38 9.4.10.1 Design Bases ..........................................................................

9.4-38 9.4.10.2 System De scription

.................................................................. 9.4-39 9.4.10.3 Safety Evaluation

.................................................................... 9.4-40 9.4.10.4 Testing and Insp ection Requirements ............................................ 9.

4-40 9.4.10.5 Instrumentation Requirements ..................................................... 9.4-40 C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 Chapter 9 AUXILIARY SYSTEMS

TABLE OF CONTENTS (Continued)

Section Page LDCN-10-007 9-x 9.4.11 PRIMARY CO NTAINMENT

....................................................... 9.4-41 9.4.11.1 Design Bases ..........................................................................

9.4-41 9.4.11.2 System De scription

.................................................................. 9.4-42 9.4.11.3 Safety Evaluation

.................................................................... 9.4-42 9.4.11.4 Testing and Insp ection Requirements ............................................ 9.

4-43 9.4.11.5 Instrumentation Requirements ..................................................... 9.4-43 9.4.12 MAKEUP WATER PUMP HOUSE

................................................ 9.4-43 9.4.12.1 Design Bases ..........................................................................

9.4-43 9.4.12.2 System De scription

.................................................................. 9.4-44 9.4.12.3 Safety Evaluation

.................................................................... 9.4-45 9.4.12.4 Testing and Insp ection Requirements ............................................ 9.

4-45 9.4.12.5 Instrumentation Requirements ..................................................... 9.4-46 9.4.13 SERVICE BU ILDING ................................................................

9.4-46 9.4.13.1 Design Bases ..........................................................................

9.4-46 9.4.13.2 System De scription

.................................................................. 9.4-46 9.4.13.3 Safety Evaluation

.................................................................... 9.4-47 9.4.14 WATER TREATMENT AREA AND MACHINE SHOP

...................... 9.4-47 9.4.14.1 Design Bases ..........................................................................

9.4-47 9.4.14.2 System De scription

.................................................................. 9.4-47 9.4.14.3 Safety Evaluation

.................................................................... 9.4-47 9.4.15 CIRCULATING WATER PUMP HOUSE ........................................ 9.4-48 9.4.15.1 Design Bases ..........................................................................

9.4-48 9.4.15.2 System De scription

.................................................................. 9.4-48 9.4.15.3 Safety Evaluation

.................................................................... 9.4-49 9.4.15.4 Testing and Insp ection Requirements ............................................ 9.

4-49 9.4.15.5 Instrumentation Requirements ..................................................... 9.4-49 9.4.16 PLANT HEATING STEAM SYSTEM ............................................ 9.4-50 9.4.16.1 Design Basis ..........................................................................

9.4-50 9.4.16.2 System De scription

.................................................................. 9.4-50 9.4.16.3 Safety Evaluation

.................................................................... 9.4-51 9.4.16.4 Testing and Insp ection Requirements ............................................ 9.

4-51 9.4.16.5 Instrumentation Requirements ..................................................... 9.4-51

9.5 OTHER

AUXILIA RY SYSTEMS ...................................................... 9.5-1 9.5.1 FIRE PROTECTION SYSTEM

....................................................... 9.5-1 C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Chapter 9 AUXILIARY SYSTEMS

TABLE OF CONTENTS (Continued)

Section Page 9-xi 9.5.2 COMMUNICATIONS SYSTEMS...................................................9.5-1 9.5.2.1 Design Basis............................................................................9.5-1 9.5.2.2 System Description....................................................................9.5-2 9.5.2.2.1 Public Te lephone Access..........................................................9.5-2 9.5.2.2.2 Private Br anch Exchange..........................................................9.

5-3 9.5.2.2.3 Sound-Powere d Telephone System..............................................9.5-4 9.5.2.2.4 Public Address and Building-Wide Alarm Systems...........................9.5-4 9.5.2.2.4.1 Public Address System..........................................................9.5-4 9.5.2.2.4.2 Building/

Zone-Audio Alarm System..........................................9.5-4 9.5.2.2.5 Radio Comm unications System...................................................9.5-5 9.5.2.2.6 Telephone Link for Connection to the BPA Dittmer Control Center......9.5-6 9.5.2.3 Inspection a nd Testing Requirements..............................................9.5-6 9.5.2.4 Capability During Postulated Accident and Anticipated Transients..........9.5-6 9.5.2.4.1 Protective Measures................................................................9.5-6 9.5.2.4.2 Severing of Lines or Trunks......................................................

9.5-7 9.5.2.4.3 Hi gh Noise...........................................................................9.5-7 9.5.2.4.4 Post-Fire Safe Shutdown..........................................................9.5-7

9.5.3 PLANT

LIGHT ING SYSTEM........................................................9.5-8 9.5.3.1 Design Bases...........................................................................9.5-8 9.5.3.2 System Description....................................................................9.5-8 9.5.3.2.1 Normal Alternating Cu rrent (ac) Lighti ng Systems...........................9.5-8 9.5.3.2.2 Normal-Emergenc y ac Lighting Systems.......................................9.5-8 9.5.3.2.3 Emergency Direct Curre nt (dc) Lighting Systems............................9.5-9 9.5.3.2.4 Battery-Pack Emer gency Lighting Systems....................................9.5-9 9.5.3.3 Safety Evaluation......................................................................9.

5-10 9.5.4 DIESEL GENERATOR FUEL OIL STORAGE AND TRANSFER SYSTEM..................................................................................

9.5-10 9.5.4.1 Design Bases...........................................................................9.5-10 9.5.4.2 System Description....................................................................9.

5-11 9.5.4.3 Safety Evaluation......................................................................9.

5-14 9.5.4.4 Testing and Inspection Requirements..............................................9.5-15 9.5.4.5 Instrumentation Requirements.......................................................9.5-15 9.5.5 DIESEL GENERATOR COOLING WATER SYSTEM.........................9.5-16

9.5.5.1 Design Bases...........................................................................9.5-16 9.5.5.2 System Description....................................................................9.

5-17 9.5.5.3 Safety Evaluation......................................................................9.

5-19 C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 Chapter 9 AUXILIARY SYSTEMS

TABLE OF CONTENTS (Continued)

Section Page LDCN-10-017 9-xii 9.5.5.4 Testing and Inspection Requirements .............................................. 9.5-19

9.5.6 STARTING

AI R SYSTEM ............................................................ 9.5-20 9.5.6.1 Design Bases ...........................................................................

9.5-20 9.5.6.2 System Description

.................................................................... 9.5-20 9.5.6.3 Safety Evaluation ...................................................................... 9.5-22 9.5.6.4 Testing and Inspection Requirements .............................................. 9.5-22

9.5.7 DIESEL

GENERATOR LU BRICATION SYSTEM .............................. 9.5-22 9.5.7.1 Design Bases ...........................................................................

9.5-22 9.5.7.2 System Description

.................................................................... 9.5-23 9.5.7.3 Safety Evaluation ...................................................................... 9.5-25 9.5.7.4 Testing and Inspection Requirements .............................................. 9.5-25

9.5.8 DIESEL

GENERATOR COMBUS TION AIR INTAKE AND EXHAUST SYSTEM .................................................................................. 9.5-26 9.5.8.1 Design Bases ...........................................................................

9.5-26 9.5.8.2 System Description

.................................................................... 9.5-26 9.5.8.3 Safety Evaluation ...................................................................... 9.5-27 9.5.8.4 Inspection and Te sting Requirements .............................................. 9.5-28

9.5.9 PLANT

DECONTAMINATION FACILITY ...................................... 9.5-28 9.5.9.1 Design Bases ...........................................................................

9.5-28 9.5.9.2 System Description

.................................................................... 9.5-29 9.5.9.3 Safety Evaluation ...................................................................... 9.5-29 9.5.9.4 Testing and Inspection Requirements .............................................. 9.5-29 9.5.10 REFERENCES ......................................................................... 9.5-30

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 Chapter 9 AUXILIARY SYSTEMS

LIST OF TABLES

Number Title Page LDCN-11-000 9-xiii 9.1-1 9 x 9-9X, SVEA-96, and AT RIUM-10 Assembly Parameters

.......... 9.1-67 9.1-2 Calculated Dose Rate During Refueling .................................... 9.1-68 9.1-3 Fuel Pool Cooling and Cleanup System

..................................... 9.1-69

9.1-4 Spent Fuel Pool Projected Heat Loads (DELETED) ..................... 9.1-70

9.1-5 Fuel Pool Cooling and Cleanup System Performance Data ............. 9.1-72

9.1-6 Bounding Fuel Pool Cooling Events ......................................... 9.1-73

9.1-7 Tools and Servici ng Equipment

.............................................. 9.1-75

9.2-1 Ultimate Heat Sink Spray Cooling Pond Design .......................... 9.2-29

9.2-2 Spray System Design ...........................................................

9.2-30 9.2-3 Diurnal Variation in Me teorological Data .................................. 9.2-31

9.2-4 Flow Rates and Associated Heat Loads Used in the Ultimate Heat Sink Analysis

.............................................................. 9.2-35

9.2-5 Heat Load Rates Used in Ultimate Heat Sink Analysis .................. 9.2-38

9.2-6 Spray Pond Water Losses and Content (30 days after design basis LOCA event) .....................................................

9.2-40 9.2-7 Source of Spray P ond Makeup Water ....................................... 9.2-41

9.2-8 Standby Service Water System Failure Analysis .......................... 9.2-42 9.2-9 Integrated Heat Data - Ultimate Heat Sink Reanalysis ................... 9.2-44 9.3-1 Equipment Characteristics ..................................................... 9.3-29

9.3-2 Available Sample Locations

................................................... 9.3-31

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Chapter 9 AUXILIARY SYSTEMS

LIST OF TABLES (Continued)

Number Title Page 9-xiv 9.4-1 Main Control Room/Cable Spr eading Room/Critical Switchgear Area Major Components of HVAC Systems...............................9.4-53 9.4-2 Reactor Building and Primary Containment Areas Major Components of HVAC Systems..............................................9.4-55

9.4-3 Radwaste Building Area Major Components of HVAC Systems.......9.4-61

9.4-4 Turbine Generator Building Major Components of HVAC Systems..9.4-65

9.4-5 Diesel Generator Buildi ng Areas Major Components of HVAC Systems..................................................................9.4-67

9.4-6 Standby Service Water Pump House Areas Major Components of HVAC Systems..............................................................9.4-69

9.4-7 Makeup Water Pump House Major Components of HVAC Systems..9.4-70

9.4-8 Circulating Water Pump House Major Components of HVAC Systems..................................................................9.4-71

9.4-9 Plant Heating Steam System..................................................9.4-72

9.5-1 Strategic Work Ar ea Communication.......................................9.5-31

9.5-2 Locations of Fixed Emergency Lighting....................................9.5-32

9.5-3 Diesel Generator Fuel Oil St orage and Transfer System.................9.5-35

9.5-4 Diesel Generator Heat Exchange r Design and Performance Data......9.5-36

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Chapter 9 AUXILIARY SYSTEMS

LIST OF FIGURES

Number Title 9-xv 9.1-1 New Fuel Storage Vault Fuel Loading Pattern 9.1-2 New Fuel Storage

9.1-3 Typical Spent Fuel Rack 12 x 16 Array

9.1-4 Spent Fuel Pool Arrangement

9.1-5 General Arrangement Plan - El. 572 ft 0 in. and El. 606 ft 10-1/2 in. - Reactor Building 9.1-6 Fuel Pool Cooling and Cleanup System (Sheets 1 and 2)

9.1-7 Fuel Preparation Machine Shown Installed in Facsimile Fuel Pool

9.1-8 New Fuel Inspection Stand

9.1-9 Channel Bolt Wrench

9.1-10 Channel Handling Tool

9.1-11 Fuel Inspection Fixture

9.1-12 General Purpose Grapple

9.1-13 Plant View of Refueling Facilities

9.1-14 Simplified Section of New Fu el Handling Facilities - Section X-X

9.1-15 Simplified Section of Refueling Facilities - Section Y-Y

9.1-16 Simplified Section of Fuel Shipping Facilities - Section Z-Z

9.1-17 Cavity In-Vessel Service Platform (CISP) Plan View

9.1-18 Cavity In-Vessel Service Platform (CISP) Sectional View C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Chapter 9 AUXILIARY SYSTEMS

LIST OF FIGURES (Continued)

Number Title 9-xvi 9.2-1 Plant Service Water System (Sheets 1 and 2)

9.2-2 Closed Cooling Water System - Reactor and Radwaste Buildings (Sheets 1 and 2)

9.2-3 Makeup Water Treatment System

9.2-4 Demineralized Water System

9.2-5 Potable Hot and Cold Water (Sheets 1 through 3)

9.2-6 Plant Sanitary Drain System

9.2-7 Composite Piping Plan, Sec tions and Details , Spray Ponds

9.2-8 Drift Loss Test Results

9.2-9 Temperature Response Following Design Basis LOCA (Standby Service Water)

9.2-10 Water Inventory in UHS Following Design Basis LOCA

9.2-11 Condensate Supply Sy stem (Sheets 1 and 2)

9.2-12 Standby Service Water Sy stem (Sheets 1 through 3)

9.2-13 Composite Piping - St andby Service Water Pump House No. 1A Plan, Sections, and Details

9.2-14 Composite Piping - St andby Service Water Pump Ho use No. 1B Plan, Sections, and Details

9.2-15 Flow Diagram - Compressor Jacket Water System - Turbin e Generator Building

9.3-1 Control and Service Air System (Sheets 1 through 5)

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Chapter 9 AUXILIARY SYSTEMS

LIST OF FIGURES (Continued)

Number Title 9-xvii 9.3-2 Containment Instrument Air System (Sheets 1 and 2) 9.3-3 Steam and Liquid Sampling -

Turbine and Service Buildings 9.3-4 Steam and Liquid Sampling - Reactor Building

9.3-5 Steam and Liquid Sampling - Radwaste Building

9.3-6 General Sample Probe

9.3-7 Sample Probe RFW-SP-3

9.3-8 Feedwater Conde nsate Sample Probe

9.3-9 Equipment Drain System - Reactor Building

9.3-10 Floor Drains - Turbine Building Equipment

9.3-11 Radwaste Building Equipment and Floor Drains

9.3-12 Floor Drains - Reactor Building

9.3-13 Nonradioactive Floor Drains

9.3-14 Standby Liquid Contro l System - Reactor Building

9.3-15 Sodium Pentaborate Decahydrat e Volume Concentration Requirements 9.3-16 Saturation Temperature of Sodium Pentaborate Solution

9.3-17 Flow Diagram, Steam and Liquid Sampling - Reactor Building

9.4-1 HVAC for Control and Switchgear Room s - Radwaste Building (Sheets 1 and 2)

9.4-2 HVAC Systems - Reactor Building (Sheets 1 through 3)

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Chapter 9 AUXILIARY SYSTEMS

LIST OF FIGURES (Continued)

Number Title 9-xviii 9.4-3 HVAC Systems - Radwaste Building (Sheets 1 and 2) 9.4-4 HVAC Chilled Water System - Radwaste Building (Sheets 1 and 2) 9.4-5 HVAC Offgas Charcoal Adsorber Vault - Radwaste Building

9.4-6 HVAC Systems - Turbine Gene rator Building (Sheets 1 and 2)

9.4-7 HVAC - CW, SW, and MUW Pump Houses and Diesel Generator Building

9.4-8 Flow Diagram - Reactor Building Primary Containment Cooling and Purging System (Sheets 1 through 3)

9.4-9 Heating Steam System - A ll Buildings (Sheets 1 and 2)

9.5-1 Diesel Oil and Miscellane ous Systems (Sheets 1 through 4)

9.5-2 General Arrangement - Diesel Generator Building

9.5-3 Supporting Systems - L ubricating Oil System and E ngine Cooling Water System with Immersion Heater Syst em - Turbocharged Units

9.5-4 Normal Engine Operation Lube Oil System - Turbocharged Engines

9.5-5 Decontamination Area General Arrangement Plan

9.5-6 Schematic Diagram - Standby Lube Oil Circulating and Keep Full System

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-1 Chapter 9 AUXILIARY SYSTEMS 9.1 FUEL STORAGE AND HANDLING

9.1.1 NEW FUEL STORAGE

9.1.1.1 Design Bases

9.1.1.1.1 Safety Design Bases

9.1.1.1.1.1 Safety Design Bases - Structural. a. The new fuel storage racks containing a full complement of fuel assemblies are designed to withstand all credible static and dynamic loadings, to prevent damage to the structure of the racks and the contained fuel, and to minimize distortion of the rack arrangement [see Table 3.9-2(s)

], b. The racks are designed to protect the fuel assemblies from excessive physical damage so as to prevent the release of radioactive materials in excess of 10 CFR 20 limits under normal or abnormal conditions, c. The racks are constr ucted in accordance with the Quality Assurance Requirements of 10 CFR Part 50, Appendix B, d. The new fuel storage racks are categorized as Safety Class 2 and Seismic Category I, and

e. The new fuel storage vault is locate d within secondary containment in the reactor building and therefore complies with the objectives set forth in Regulatory Guide 1.13, and General Design Criteria (G DC) 2, 3, 4, 5, 61, 62 and 63 of 10 CFR Part 50, Appendix A.

9.1.1.1.1.2 Safety Design Bases - Nuclear. a. The new fuel storage racks are design ed and maintained with sufficient spacing between the new fuel assemb lies to ensure that the a rray shall be subcritical by at least 5% k, including allowance for calcula tion biases and uncertainties.

Calculations were perfor med to ensure that k eff 0.95 for all water mi st densities up to 1.0 gm/cm 3 in and around the fuel. The new fuel storage vault was modeled using the KENO

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 9.1-2 computer code (Reference 9.1-1). GE14 fuel was modeled usi ng the MCNP4A computer code (Reference 9.1-10).

The biases between the calculated results and experimental results, as well as the uncertainty involved in the calculations, are taken into account as part of the calculation procedure to ensure that the specified K eff limits are met.

9.1.1.1.2 Power Gene ration Design Bases

a. New fuel storage racks provide for approximately 8% of the full core fuel load, and

b. New fuel storage racks are designed and arranged so that the fuel assemblies can be handled efficiently dur ing refueling operations.

9.1.1.2 Facilities Description

The new fuel storage vault containing the new fuel storage racks is a concrete structure adjacent to the spent fuel pool at the refu eling floor level of the reactor building (see Figure 1.2-10

). The reactor building is built to Seismi c Category I requireme nts and is further discussed in Section 3.2. The new fuel rack design features are as follows:

a. The new fuel storage vault contains 24 sets of castings, with each casting containing 10 possible storage locati ons. A maximum of 60 fuel storage positions will be used. The remaining stor age locations will be blocked with a positive physical barrier, such that no fuel can be placed in the blocked positions.

Figure 9.1-1 shows the locations where fuel will be stored;

b. There are three tiers of castings positioned by fixed box beams. The castings hold the fuel assemblies in a vertical position; the fuel assemblies are supported at the lower tie plate, with additional lateral support;

c. The lower casting is designed to suppor t the weight of the fuel assembly and restrict lateral movement when used in the as-built configuration. To facilitate the assembly of ABB type fuel, a stub t ube rests in the bottom casting to raise the fuel approximately 30 in. The center and top casti ngs restrict only lateral movement of the fuel assembly;

d. The new fuel storage racks are made from aluminum. Materials used for construction are specified in accordance with the latest issue of applicable ASTM specifications. The material c hoice is based on a consideration of the susceptibility of various me tal combinations to electrochemical reaction. When considering the susceptibility of metals to galvanic corrosion, aluminum and

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-3 stainless steel are si milar insofar as their coupled potential is concerned. The use of stainless steel fasteners in aluminum to avoid detrimental galvanic corrosion is a recommended practice and has been used successfully for many years by the aluminum industry. The stub tubes are used when assembling Westinghouse type fuel, are ma de of stainless steel and are not directly fastened to lower casting;

e. Lead-in and lead-out of the casting provide guidance of the fuel assembly during insertion or withdrawal; and

f. The nominal center-to-center spacing with rows is 7 in. and between rows is 12 in. However, as shown in Figure 9.1-1, face adjacent locations are not used.

The fuel assembly cente r-to-center spacing is a multiple of the above dimensions.

The fuel assemblies are lo aded into the rack through the top.

Each hole for a fuel assembly has adequate clearance for inserting or withdrawing the assembly channeled or unchanneled.

Sufficient guidance is provided to preclude damage to the fuel assemblies. The design of the racks and the blocking devices pr event accidental insertion of th e fuel assembly into a position not intended for the fuel. The only spaces in the racks are thos e into which it is intended to insert fuel. The weight of the fuel assembly is supp orted by the lower tie plate which is seated in a chamfered hole or stub tube in the base casting.

The floor of the new fuel storage vault is sloped towards a drain located at the low point. This removes any water that may be accidentally and unknowingly intr oduced into the vault. The drain is part of the floor drain subs ystem of the liquid radwaste system.

The radiation monitoring equipment for the new fuel storage area is described in Section 12.3.4.

9.1.1.3 Safety Evaluation

9.1.1.3.1 Criticality Control

The calculations of k eff are based on the geometri cal arrangements of the fuel array, and that subcriticality does not depend on th e presence of neutron absorbi ng materials. To meet the requirements of GDC 62, geometrica lly safe configurations of fuel stored in the new fuel array are used to en sure that k eff will not exceed 0.95 if fuel is stor ed in the dry condition or in an abnormal mist condition with water densities up to 1.0 g/cm

3.

The new fuel storage vault has conc rete covers with rubber seals.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 9.1-4 The fuel storage rack is designed using noncombustible materials.

New fuel storage vault covers prevent optimum moderation in the ne w fuel vault (water density between 0 and 1 g/cc).

The GE14 fuel assembly is a 10 x 10 fuel rod arra y with 2 internal water channels offset in the center of the assembly (displacing 8 fuel rod lo cations). The GE14 nomina l design parameters are listed in Table 9.1-1. The GE14 analysis new fuel st orage vault model assumes several key assumptions:

a. New fuel assemblies are assumed to contain the highest enriched unexposed lattice for the entire length of the assembly and are associated with a beginning-of-life reactivity of 1.31 with benchmark uncertainties included.

b. The GE14 analysis assumes infinite fuel in the axial direction (no axial reflector) for normal dry storage conditions and fully flooded abnormal conditions. Neutron absorption in fuel struct ural components is neglected.

c. The GE14 analysis assumes a finite geometry for a single new fuel vault loaded in accordance with the loading pattern shown in Figure 9.1-1. Neutron leakage is assumed on all sides, top, and bottom for optim ally moderated abnormal conditions. Neutron absorption in fuel st ructural components is neglected with the exception of the concrete walls of the new fuel storage vault.

d. Up to 3 GE14 bundles can be safely handled in the spent fuel pool, outside of the spent fuel rack, in any configuration.

e. The results of the analysis include a statistical rollup of all significant manufacturing and calculational uncertainties and abnormal storage condition biases. f. An infinitely long fuel bundle is modeled to lie horiz ontally at 1 cm above the active fuel region of the rack.

g. An abnormal condition that evaluates a misplaced fuel bundle that is loaded in addition to the fully loaded fu el loading pattern, shown in Figure 9.1-1 , in a position that yields the optimum reactivity for the scenario.

9.1.1.3.2 New Fuel R ack Structural Design

The new fuel storage racks are designed to meet Seismic Category I requirements.

The maximum stress in the fully load rack in a faulted cond ition is 16.5 kips. This is significantly lower than the allowable stress.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-5 The storage rack is designed to withstand horizontal combined loads up to 222,000 lb, well in excess of expected loads.

The storage rack is designed to withstand the pull-up force of 400 0 lb and a horizontal force of 1000 lb. There are no readily av ailable forces in excess of 1000 lb. The racks are designed with lead outs to prevent sticking. However, in the event of a stuck fu el bundle, the upward force on the racks is limited to the 1000 lb load limit of the jib crane (M T-CRA-9A or 9B) used to handle new fuel.

The storage rack structure (Figure 9.1-2) is designed to withstand the impact resulting from a falling weight. Tests using a simulated fuel bundle have been conducted to verify that the rack casting can withstand the impact from a bundle dropped from above the array. During testing the lowest drop to cause the rack casting to exceed ultimate stress was a drop of 6.17 ft (4314 ft/lb) made at mid span.

Therefore, procedural requireme nts dictate that no more than one bundle at a time can be handled over the storage array.

These requirements ensure that the racks cannot be displaced in a manner causing critical spacing as a result of impact from a dropped fuel assembly. Since the 125-ton reactor building cr ane can traverse th e full length of the refueling floor, administrative controls will prohib it carrying loads over the new fuel.

9.1.1.3.3 New Fuel Handling

New fuel is carried to the new fuel vault a nd placed in the storag e rack with jib crane MT-CRA-9A or 9B or the reactor building cran e auxiliary hoist. To handle the new fuel, rigging controlled by plant procedures is used, which interfaces with the lifting device at the upper end and the fuel bundl e bail at the lower end.

During the positioning of a new fu el assembly into the new fuel rack, the rigging is always above the upper fuel rack casting. The rigging in terfaces only with the fuel bundle bail, thus precluding engagement of the fuel rack. The transfer devices used for new fuel handling to the new fuel vault, therefore, cannot impose uplift loads on the rack castings. See Section 9.1.4.2.10.1 for further discussion of new fuel handling.

9.1.1.3.4 Other New Fuel Storage Design Factors The new fuel racks are designed to be restrained by holddown bolts to en sure that rack spacing does not vary during the safe shutdown earthquake (SSE).

The storage rack structur e is so designed that the height of the rack is less than the length of the fuel bundle. Therefore, the upper tie plate of the bundle ca nnot pass below the top cross member of the rack. Also, the fuel bundle insertion spaces in th e top casting of the rack have

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-6 a lead in chamfer on the upper and lower surf aces. These design fe atures prevent any tendency of the fuel bundle to jam during insertion into or rem oval from the rack.

The new fuel storage rack castings are made from aluminum and are s ecured by stainless steel fasteners. Materials used for construction are specified in accordance with ASTM Specifications B108, B179, B209, B 211, and B221 date d January 1971.

The new fuel storage racks do not require any pe riodic special testing or inspection for nuclear safety purposes.

9.1.2 SPENT

FUEL STORAGE

9.1.2.1 Design Bases

9.1.2.1.1 Safety Design Bases

9.1.2.1.1.1 Safety Design Bases - Structural. a. The spent fuel storage racks are designed to withstand the affects of the SSE and remain functional and mainta in subcriticality. The r acks are also designed to withstand the impact of a dropped fuel as sembly or the upward force of a stuck assembly without loss of function. Th e racks are designed and fabricated to meet seismic and quality class requirements per 10 CFR 50, Appendix B;

b. The spent fuel stor age facility is located so that no missile s can enter the fuel pool with the necessary energy to cau se any damage to the fuel; and

c. The reactor building containing the spent fuel storage facility provides the capability for limiting the potential offsite exposures in accordance with 10 CFR Part 100 in the event of signifi cant release of radioactivity from the stored fuel.

9.1.2.1.1.2 Safety Design Bases - Nuclear. a. The center-to-center spacing between stored fuel assemblies in a fully loaded rack is sufficient to maintain a k eff less than 0.95 at a conservative water temperature.

b. The spent fuel storage rack design precludes storage of a fuel assembly other than where intended at a nominal 6.5 in. center to center distance between fuel assembly placed in the storage rack.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-7 c. The spent fuel storage racks are designed to allow ade quate cooling of the stored spent fuel assemblies; and

d. Shielding for the spent fu el storage arrangement is su fficient to protect plant personnel from exposure to radiation in excess of 10 CFR Part 20 limits.

9.1.2.1.2 Power Gene ration Design Bases

a. Spent fuel storag e space in the fuel storage pool is for 2658 fuel assemblies, and

b. Spent fuel storage racks are designed and arranged so that fuel assemblies can be handled efficiently dur ing refueling operations.

9.1.2.2 Facilities Description

9.1.2.2.1 Spent Fuel Storage Racks

Spent fuel storage racks provide a place in the fuel pool for storing the spent fuel discharged from the reactor vessel. They are top entry racks, designed to maintain the spent fuel in a space geometry that precludes the possibility of criticality under both normal and abnormal conditions. This is accomplished with the aid of neutron absorbing plates. The location of the spent fuel pool within the plant is shown in Figures 1.2-10 and 1.2-11.

The spent fuel storage rack design, shown in Figure 9.1-3, consists of fuel storage cells which are square stainless steel tubes with neutron absorbing B 4 C plates between them. A stainless steel plate grid at the top and the bottom of the tubes, to which the tubes are welded, form the tubes into racks and maintain center-to-center spacing between the tubes at 6.5 in. The racks are welded together into modules which are held firmly in place by seismi c restraints attached between the rack modules and the pool wall. The storage racks are made of stainless steel.

The square tube storage cells are 1/8 in. thick.

The neutron absorber plates have nominal di mensions of 19 in. long, 5.88 in. wide, and 0.2 in. thick. They are composed of B 4 C granular material bonded together to form a plate of uniform properties. They have a nominal B-10 loading of 0.0959 g/cm 2 of plate and a plate density of 0.05 lb/in

3. The plate has been shown by tests to have negligible corrosion in water and thermally stable over the range of pool water temperatures that can occur. The plates are seal welded in a stainless steel cavity to prevent water intrusion and are vented at the pool curb through sampling valves.

There are no load bearing requirements for the plates. Based on the results of the Modulus of Rupture tests, the plates will w ithstand approximately two times the calculat ed stresses caused by a postulated seismic event. Pl ate integrity and mechanical pr operties have been verified by comprehensive tests. These tests included Modulus of Rupture and Modulus of Elasticity tests.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-8 The Modulus of Rupture testi ng was performed using a three point support method and was done on specimens at temperatur es varying from ambient to 300°F, specimens soaked in water, and irradiated specimens. The modulus of elastic ity was performed using a resonance procedure and was done at varying temperatures and after the plate had been immersed in water. The tests showed no swelling, cracking, or dimens ional changes and provided verification of the plate mechanical pr operties required for the rack design.

To prevent distortion of the spent fuel rack cavi ties and binding of spen t fuel assemblies due to pressure buildup, the enclosures are vented through an arrangement of tubing and sampling valves to above the storage pool water surfa ce at the pool curb. Monitoring for offgas pressure, sampling of offgas , and venting to relieve pr essure are provided for.

Different rack sizes are used (12 x 16, 12 x 13, 8 x 13, 7 x 18, and 11 x 16 arrays) to take full advantage of the fuel st orage space in the pool (see Figure 9.1-4

). The upper rack structures are welded to an elevated base plate which, in turn, is support ed by a system of welded beams and stiffeners. The base serves to support the weight of the fu el assemblies and to distribute the load to the pool floor. Th e base plate contains an opening at each fuel assembly storage location which accommodates the fuel assembly lower nozzle. Natura l circulation of pool water flows upward through the lo wer nozzle and the fuel assemb ly to remove decay heat.

The storage cells are designed to provide lateral support for the storage assemblies

The minimum edge-to-edge distance of the assembly array from adjacent concrete walls is 16.75 in. between the edge of the C-14 storage rack and the shipping cask storage area wall.

The seismic restraints are stai nless steel turnbuckle s located between the pool walls and the racks around the periphery of the pool (Figure 9.1-4). They are located at both the top and bottom of the rack and will transmit the seismic forces of the operati ng basis earthquake (OBE) and the SSE between the racks and the walls and remain functional. The turnbuckles are connected at the wall to stainless steel bands which are embedded in the concrete wall and seal welded to the pool liner.

9.1.2.2.2 Spent Fuel Storage Pool

The spent fuel storage pool is designed to withstand earthquake loadings as a Seismic Category I structure. It is a reinforced-concrete structure completely lined with stainless steel, which provides a leakproof membrane that is resistant to abrasion and damage during normal and refueling operations. The stainless steel liner plates are seamwelded together and are anchored to the surrounding conc rete by concrete anchors in the walls and structural members in the floor welded to the liner plates. Each liner weld seam is backed up by a drainage monitoring channel. These channels form a series of inte rconnecting systems designed to provide the following:

a. Detection, measurement, and location of any liner leakage, C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-9 b. Prevention of pressure buildup behind the liner plat es due to leakage, and

c. Prevention of uncontrolled loss of cont aminated pool water to other relatively clean locations within th e secondary containment.

This network of drainage monitoring channels is embedded in the concrete behind the liners and is designed to permit free gravity drainage to the radioactive drai n system, the flow of which is monitored.

The refueling canal connecting th e spent fuel storage pool to the reactor well is provided with two gates in series, with a monitor drain betwee n them. This arrangeme nt permits monitoring for any leaks and facilitates repair of a gate or seal, if required.

A gamma scan collimator port is located 15 ft 11 in. from the bottom of the fuel pool and extends through the side of the pool. Through this port, gamma scanning of radioactive reactor components and spent fuel assemblies can be performed. This nondestructive method of analysis by the measurement of the gamma radiation being emitted by a material can indicate fuel enrichment, reactor power distribution, or fission product content of the component being analyzed.

The water supply to the spent fuel pool is provided and level maintain ed as described in Section 9.1.3.

9.1.2.3 Safety Evaluation

9.1.2.3.1 Criticality Control

9.1.2.3.1.1 8 x 8 Fuel. The design of the spent fuel stor age racks provides for a subcritical multiplication factor (k eff) of <0.95 for both normal a nd abnormal storage conditions. Normal conditions exist when the fuel storage racks are covered with a normal depth of water (about 23 ft above the active fu el) for radiation shielding, and with the maximum number of fuel assemblies or bundles in their design storage position. An abnormal c ondition may result from damage caused by accidental drop of a fuel assembly or stuck fuel assembly during attempted withdrawal.

The criticality analyses of the normal condition included severa l conservative assumptions as well as the effect of uncertainti es in calculation met hod and geometric and ma terial variations of the fuel storage rack. The following conservative assumptions were used in the calculation:

a. Fresh fuel of 3.25 wt %

235 U enrichment.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-10 Initially the maximum enrich ment will be much lower than this, but could approach this value if an 18-month fuel cycle is used. The enrichment selected is higher than the average enrichment of any fuel expected to be stored in the spent fuel pool. It was chosen because the fully loaded rack of fuel with this enrichment gives a more reactive condition than any presently foreseen;

b. Uniform planar array of 3.25 wt % enrichment fuel.

Calculations have shown that this is conservative compared to the realistic, planar distributed enrichments within an assembly;

c. Spent fuel pool bulk wate r temperature 68°F. This is considerably lower than expected. Nevertheless, a calculation wa s done to determine the increase in reactivity due to a decrease in pool temperature to 32°F. The results showed the effect to be negligible;

d. Fuel racks are infinite in three dimensions; and

e. Fixed neutron poisons in the fuel assembly are neglected.

The majority of the calculations were performed with methods commonly used in light water reactor design; i.e., four-group diffusi on theory cell calcul ations using PDQ-7 (Reference 9.1-2). Cross sections for these calcul ations are generated with NUMICE-2 (Reference 9.1-3), the NUS Corporation version of th e Westinghouse LEOPARD code. This code uses the same cross section library tape and calculationa l techniques as LEOPARD.

Selected cases were checked and the final design multiplication factors were verified with Monte Carlo calculations using KENO-IV (Reference 9.1-4), with a 123-group cross section library generated from a basic GAM-THERMOS library using two subroutines, NITAWL and XSDRNPM, in the AMPX1 (Reference 9.1-5) code package. Bo th the PDQ-7 and the KENO-IV calculation methods, as described above , have been benchmarked. These calculation methods, as describe d, were used for the Columb ia Generating Station (CGS) calculation and do not contain any significant modifications.

Under normal conditions, for a ce nter-to-center spacing of 6.5 in. between fuel assemblies with B 4C plates surrounding each stored fuel assembly, the k eff , as determined using KENO, is 0.851. With the void space between the B 4 C plates and the stainles s steel box flooded with water, the KENO calculation yielded a lower keff. Calculation uncertainties were determined from comparison between calculation and experime nts using KENO and a statistical evaluation of Monte Carlo runs. The results indicated a calculational uncertainty for the former of 0.013 k and for the latter 0.010 k at a 95% confidence level; this represents a total calculation uncertainty of 0.023 k. Mechanical spacing and to lerances acting in a direction close to the water gaps between adjacent racks result in a slight reac tivity increase of 0.002 k. Production tolerances of B 4 C plates result in a r eactivity increase of 0.003 k.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 9.1-11 To determine the effect of reduced or missing neutron absorbing material, it was assumed that one out of every 25 neutron absorber plates was missing. This case is extremely unlikely but shows poison variation sensitivity. The results of the calculation was an increase in reactivity of 0.015 k. A temperature decrease in pool water temperature of 32°F was included, i.e., 0.004 k. Adding the total calcul ational uncertainties of 0.023 k and the total geometric and material uncerta inties of 0.024 k to the nominal keff results in a k eff of 0.897 with a confidence level of 95%. This is well below the design basis of <

0.95% for the normal wet condition.

Two abnormal conditions were also considered. They are (a) a dropped assembly assumed to lay across the top of a fuel rack and (b) a fuel assembly in tr ansport in a vertical position, accidentally dropped into the water channels between racks.

Of these two, the second condition is more severe. For the first cond ition, the end fittings on the top of the BWR assembly prohibits a spacing between the dropped assembly and the active fuel in the storage racks of less than 11.6 in. Using the same techniques, assumpti ons, and uncertainties previously discussed, the second case resulted in a k eff of 0.903. This is only slightly different from the conservative normal condition and within the design basis k eff of <0.95.

9.1.2.3.1.2 9 x 9 and 10 x 10 Fuel. The criticality safety of the spent fuel storage rack with 9 x 9-9X, SVEA-96 (10 x 10), ATRIUM-10 (10 x 10) and GE14 (10 x 10) fu el is assessed in accordance with NUREG 0800 and ANSI/ANS 57.2 (References 9.1-7 , 9.1-8 , 9.1-9 and 9.1-11). The spent fuel storage rack meets the app licable criticality safe ty criteria subject to the conditions given below:

a. Fuel design - As specified below w ith a maximum enrichment of 4.0 wt %

235 U for 9 x 9-9X fuel.

b. For SVEA-96 fuel - Enrichment above 3.77 wt %

235 U is allowed in the spent fuel pool by determining the reactivity e quivalency. Reactiv ity equivalency is predicated on the reduction in reactivity associated with the co mbination of fuel depletion and gadolinia burnable absorbers (Reference 9.1-8). The reactivity equivalency is checked for each SVEA assembly design for each reload. Any

significant deviation from the fuel asse mbly geometry specified by this design will require additional analysis as specified in Westinghouse methodology (CE NPSD-786-P, Revision 1).

c. For ATRIUM-10 fuel, the maximum enriched lattice zone is 4.6 wt % 235 U with 10 gadolinia rods at 2.0 wt % Gd 2 O 3. Any significant deviation from the fuel assembly geometry specified by this design will require additional analysis as specified in AR EVA NP methodology (Reference 9.1-9).

d. The CGS spent fuel fuel storage racks were analyzed for the storage of GE14 fuel bundles. The analyses were perf ormed with the MCNP4A Monte Carlo neutron transport program (Reference 9.1-10). The GE14 spent fuel design

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 9.1-12 basis bundle was analyzed in rack with an infinite lattice geometry and uniform 4.9w% U-235 enrichment, using both full and part length fuel rods with initial distributed fissile inventor ies, burnable absorber, and evaluated at its respective cold, uncontrolled, exposure-dependent reactivity state points. Representative placement and numbers of Gadolinium rods were used to model a peak in-core eigenvalue (k) of 1.33 using the TGBLA06A production lattice physics code with benchmark uncertainties included. Any fuel meeting a k of 1.33 and the established in-core calcula ted benchmark uncertainties of the GE14 analysis are also bounded by the results of the GE14 an alysis. No further evaluation is required.

e. When stored, the fuel a ssemblies must be inserted into the fuel storage boxes such that no fully enriched fuel protrudes above the B 4C absorber plates to a maximum fuel assembly protrusion of less than six in. (i.e., no more than 11.4 in. from the top of the bail handle to the top of the storage rack).

f. B 4C content in the absorber plat es is constant over time; and

g. The B 4 C plates contain no significant gaps.

The key assembly design parameters used in these calculations are listed in Table 9.1-1. The 9 x 9 assembly has 72 enriched uranium fuel rods. An internal water channel is located in the central portion of the assembly.

The SVEA-96 assembly has 96 fuel rods in a 10 x 10 array with four centr al rods missing. An internal cruciform water channel separates the 96 rods into four subassemblies of 24 rods each.

a. The neutron multiplication factors were calculated using KENO V.a,

b. Specular reflection was used with all m odels which means that no credit is taken for neutron leakage. In other words, k and keff are identical for the models used in this analysis, and

c. All codes and cross sections have been benchmarked against critical experiment data. The major assumptions made in this analysis are as follows:

a. Fuel enrichment is the bundle average in all rod locations, b. Fuel contains no burnable poisons,

c. No soluble poisons are present in the water, C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 9.1-13

d. Spent fuel pool bulk water temperature 68°F including a temperature decrease to 32°F,

e. B 4C content in the absorber plat es is constant over time, and

f. The B 4 C plates contain no significant gaps.

The first four assumptions make the analysis conservative. The last two are limiting conditions.

The ATRIUM-10 fuel assembly is a 10 x 10 fuel rod array with an internal water channel offset in the center of the assembly (displacing ni ne fuel rod locations). The assembly contains part length fuel rods constituting a "top" la ttice (above the part length fuel rods) and a "bottom" lattice (below the part length fuel rods). The ATRIUM-10 nominal design parameters are listed in Table 9.1-1.

The following conservative assumptions are made:

a. The results are based on a moderato r temperature of 4°C (39°F), which gives the highest reactivity fo r the fuel pool racks;

b. Fuel assemblies are assumed to contain the highest enriched lattice (highest reactive lattice) for the entire length of the assembly;

c. Each fuel assembly in the array is assumed to be at the peak reactivity of its lifetime;

d. Analyses assumes infinite fuel in the axial direction (no axial reflector);

e. Neutron absorption in fuel struct ural components is neglected; and

f. The maximum reactivity values incl ude all significant manufacturing and calculational uncertainties.

The GE14 fuel assembly is a 10 x 10 fuel rod arra y with 2 internal water channels offset in the center of the assembly (displacing 8 fuel rod lo cations). The GE14 nomina l design parameters are: GE14 Features (mm) (in.) Channel Corner Thickness 3.05 0.12 Side Thickness 1.91 0.075

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 9.1-14 Groove Thickness 1.91 0.075 Inside Width 134.06 5.278 Inside Radius 11.43 0.45 Fuel Rod Dimensions Total Cladding Thickness 0.66 0.026 Outside Diameter 10.26 0.404 Inside Diameter 8.94 0.352 Pellet Diameter 8.76 0.345 Water Rod Dimensions Outside Diameter 24.89 0.98 Inside Diameter 23.37 0.92 Bundle Lattice Dimensions Rod Pitch 12.95 0.51 Rod to Rod Gap 2.69 0.106 Rod to Channel Gap 3.61 0.142 The GE14 analysis 12 x 16 spent fuel storage rack model assumes several key assumptions:

a. Fuel assemblies are assumed to contain the highest enriched lattice for the entire length of the assembly and are associated with a peak in-cor e reactivity of 1.33 with benchmark uncertainties included.

b. The GE14 analysis assume s infinite fuel in the axial direction (no axial reflector).

c. Neutron absorption in fuel stru ctural components is neglected.

d. The outer sides of the 12 x 16 storage rack array (Figures 9.1-3 and 9.1-4) contain no boron carbide absorber plates. Boron carbide absorber plates are only located along the inner walls of the spent fuel rack lattic e structure. The 12 x 16 infinite array m odel is bounding for the act ual spacing and rack configurations in the spent fuel pool w ith mixed adjacent 12 x 16 and 12 x 11 fuel arrays.

e. Up to three GE14 bundles can be safely handled in the spent fuel pool, outside of the spent fuel rack, in any configuration.

f. The results of the analysis include a statistical rollup of all significant manufacturing and calculation uncertain ties and abnormal storage condition biases.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 9.1-15 g. The following abnormal conditions have b een evaluated as part of this analysis and have been accounted for within the bi as of the maximum k-effective result:

i. Four damaged fuel bundles surround the non-borated intersection of the four 12 x 16 array spent fuel storage racks. This model approximates the damage to the bundles by optimizing the fuel rod pitch of the design basis bundle within the fuel box with out consideration of the bundle channel. The fuel rod arrays in each of the four fuel boxes are placed as close to the intersection of the four 12 x 16 arrays as possible. This model does not consider damage to the storage rack.

ii. Misplaced/dropped fuel bundle in between storage rack modules.

iii. An infinitely long fuel bundle is modeled to lie horizontally at 1 cm above the active fuel region of the rack.

iv. A single bundle protruding from the top of the borated region of the spent fuel rack by a maximum distance of 6 in.

All normal and credible abnormal conditions are found to have an acceptable reactivity, (k eff) of <0.95 after adding the calculational uncertainty. The assumption th at the entire fuel storage rack contains 9 x 9-9X fuel with up to 4.0 wt %

235 U, SVEA-96 fuel with up to 3.77 wt % 235U, ATRIUM-10 fuel with each lattice zone 4.6 wt % 235 U with 10 gadolinia rods at 2 wt % Gd 2 O 3 and GE14 fuel with 4.9 wt% 235 U and a Gd 2 O 3 loading that corresponds to an in-core reactivity of 1.33 conservatively accounts for the presence and potential intermixing of older and lower enriched 8 x 8 fuel bundles.

9.1.2.3.2 Spent Fuel Storag e Rack Structural Design

In accordance with Regulatory Guide 1.13, Revision 1, the spent fuel storage racks are designed to Seismic Category I requ irements. Structural integrity of the racks when subject to normal and abnormal loads, as well as seismic loads, is demonstrated in accordance with the load requirements and acceptance criteria of Standard Review Plan Section 3.8.4. The loads considered are

a. Dead loads which are the dead weight of the rack and fuel assemblies and hydrostatic loads,

b. Live loads, i.e., the effect of lifting empty racks during installation,

c. Thermal loads, which include the uniform thermal expansion of the racks due to increases in average pool te mperature, and a thermal gr adient between adjacent storage locations, C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-16 d. The seismic forces of the OBE and SSE,

e. Accidental drop of a fuel assembly fr om the maximum possible height consistent with fuel handling operations, which is 4 ft above the top of fuel rack, and

f. Postulated stuck fuel assembly causing an upward force of 1200 lb, equal to the fuel grapple load limit to be exerted on the assembly upon attempted withdrawal.

The spent fuel storage racks were analyzed for six combinations of these loads using elastic working stress design methods. The combinations are

a. Dead loads plus live loads,

b. Dead loads plus OBE,

c. Dead loads plus thermal loads plus OBE, d. Dead loads plus thermal loads plus SSE,

e. Dead loads plus thermal load s plus fuel assembly drop, and f. Dead loads plus thermal lo ads plus stuck fuel assembly.

Live loads are not included in lo ad combinations b through f. Th e only live load on the rack is that due to lifting of the racks wh ich is performed with the racks empty. In all cases the loads were below the strength limits, which were determ ined from Part I of the AISC "Specification for the Design, Fabrication and Reaction of St ructural Steel for Build ings," February 12, 1969, and Supplements 1, 2, and 3. (Supplement 3 was effective June 12, 1974.)

Individual fuel racks are welded into either 1 x 2 or 2 x 2 rack arrays or "super modules."

These super modules are then atta ched to the fuel pool walls at two elevations by adjustable seismic restraints which are essentially large turnbuckles. These restraints are designed and positioned to eliminate any significant thermal grow th loads on the walls.

The fundamental frequency of lateral vibration of the welded rack array yielding the lowest frequency was determined using the STARDYNE3 (Reference 9.1-6) computer program. The model, consisting of beam and plate elements and lumped masses, represents a three-dimensional 2 x 2 rack a rray. The model has an array of beams, representing fuel storage cans, connecting the upper and lower grids, and resting on a base grid which is held off the pool floor by support feet. The rack is restrained laterally by two levels springs, representing the seismic restraints. The stiffnesses of the fuel assemblies and B 4 C neutron absorber plates conservatively were neglected. However, the mass of the fuel assemblies, as well as the mass of the B 4C plates and an effective mass of water was considered to be uniformly distributed over the height of th e fuel boxes. The resu lts of the STARDYNE analysis showed a lateral natural frequency of 14.7 Hz. A dynamic analysis was performed using the horizontal fl oor response spectra (dam ping 0.5% of critical).

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 9.1-17 The fundamental frequency of ve rtical vibration of the rack was also determined using the STARDYNE computer program. The same model replacing lateral mass with vertical mass was utilized. In this case, sin ce the fuel rests on the base fram e, the entire ma ss of the fuel was lumped at the base grid. Since the calculated frequenc y was 50.9 Hz and the vertical floor response spectra (damping 0.5% of critical) showed constant ac celeration at frequencies in excess of 18 Hz, the effects of the vertical acce lerations were consider ed using the zero-period acceleration in a static analysis. The lateral and vertical loads were considered to be acting simultaneously.

In the general seismic/structural analysis of the fuel racks, the mass of a fuel assembly is assumed to be uniformly distribut ed along the length of each of th e fuel storage cans. Since a maximum gap on the order of 3/8 in. exists between the side of a fuel assembly and the can (when the fuel is not encased in a channel), the fuel will actually move within the can during a seismic event and cause impact loads to be transmitted to the fuel rack restraints. The effects of this fuel can interaction are determined using a simplified finite element model of the rack and fuel. A nonlinear dynamic analysis is performed using the ANSYS computer program. Details of this analysis are given in NUS Corporation Technical Report 2060, entitled "Fuel-Can Interaction Analysis," October 1977.

Using the given loads, load comb inations, and analytical methods, stresses were calculated at critical sections of the r ack and compared to the st ructural acceptance criteria. In all cases, the calculated stress did not exceed the allowable stress.

To ensure the integrity of the sp ent fuel storage racks in the event that wa ter has leaked into the racks, specially designed control samples, consisting of B 4 C plates in vented (to pool water) canisters, are placed in a readily accessible position in th e spent fuel pool.

These samples are subjected to periodic examinations to check for possible deterioration and they are also analyzed to ensure that the boron has not leached from the plates.

Section 15.7.4 presents an analysis (for radiological considerations) of fuel handling accidents.

The high density fuel rack designer, NUS Corpor ation, analyzed the fu el racks from both a structural and criticality standpoint concerning a 1510-lb object dropped from the surface of the fuel pool. The results indicated that none of th e fuel rack damage that might occur in this situation would lead to a criticality problem.

Details of these analys es are given in NUS Corporation Technical Reports 5326-FA-01 and G-RA-17 entitled, "Structural Analysis of the CGS Rack and Fuel Assemblies for an Accidental Object Drop Loading Condition," and "Criticality Analysis of Dro pped Object Accident for WNP-2 Spent Fuel Storage Racks," respectively.

Similarly, independent an alysis of GE14 fuel indi cates that none of the fuel rack damage that might occur would result in incr eased reactivity that could exceed the 0.95 safety limit. This C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 9.1-18 evaluation is based the guidance of NUREG-0612 a nd includes damage to the four fuel bundles that surround the non-borated inte rsection of the four 12 x 16 array spent fuel storage racks (Figures 9.1-3 and 9.1-4). The 12 x 16 infinite array mode l is bounding for the actual spacing and rack configurations in the spent fuel pool with mixed adjacent 12 x 16 and 12 x 11 fuel arrays. This model approximates the damage to the bundles by op timizing the fuel rod pitch of the design basis bundle within the fuel box without consideration of the bundle channel. The fuel rod arrays in each of the four fuel boxes are placed as close to the intersection of the four 12 x 16 arrays as possible.

The model does not consider damage to the storage rack.

During shutdown, crane operations over the spen t fuel storage pool (whe n fuel assemblies are stored within) are suspended wh en operability of less than th e required number of onsite and offsite power sources occurs, as defi ned in the Technical Specifications.

9.1.2.3.3 Spent Fuel and Cask Handling

The 125-ton reactor building crane traverses the full length of the refueling floor level of the reactor building. The design of the refueling floor provides aisles on both sides of the fuel pool for moving components past (and not over) the fuel storage pool. Interlocks on the reactor building crane prevent tr avel over the spent fuel racks. The interlock-controlled restricted area for crane travel is shown in Figure 9.1-5. The interlocks are bypassed only when it is necessary to operate the crane in the fu el pool area in conjunction with activities associated with fuel handling and storage. Duri ng these rare occasions wh en the interlocks are bypassed, administrative controls ar e used to prevent th e crane from carrying loads that are not necessary for fuel handling or storage and whic h are in excess of the rack design drop load (one fuel assembly at 4 ft above the top of the fuel rack). See Section 9.1.2.3.2.

Transfer of fuel assemblies betw een the reactor well a nd the spent fuel pool is performed with the refueling platform (see Section 9.1.4.2.10.2). The fuel grapple or the auxiliary fuel hoist may be used, depending on the transfer operation.

The grapple and hoists are provided with load sensing and limiting de vices designed to the following limits:

NF-500 Fuel Grapple Auxiliary Hoists (lb) (lb)

Load limiting switch NF500 1700 1000 Load sensing switch Auxiliary Hoist 535 Load sensing switch NF500 750 Stall torque for hoist system 3000 3000

The load limiting features of the refueling platfo rm grapple and auxiliary fuel hoist will prevent damage to the fuel racks if a fuel assembly accidentally enga ges a rack while being lifted.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-19 These load limits provide a re dundant safety feature since the fuel handling grapple is not lowered below the upper fuel rack and is designed to interface only with the fuel bail. Thus, the possibility of inadvertent direct lifting of the racks with the grapple is precluded.

Guard rails around the spent fuel pool prevent the falling of fuel handling area machinery into the pool. Other objects that could conceivably fall into the pool will not transfer energy amounts exceeding the specified limits of the fuel racks.

The preclusion of accidental dropping of the sp ent fuel cask on the spent fuel racks is accomplished by incorporating a sepa rate cask storage area in the spent fuel pool, by interlocks on the reactor building crane, and by designing the path of the sp ent fuel cask to avoid passing over the spent fuel racks. The pool cask storage area is separated from the spent fuel racks by a wall over which the spent fuel is transferred.

For removal of spent fuel from the plant, a spent fu el cask is lowered into the cask area. Transfer of fuel to the cask is made over the wall between the spent fu el racks and the cask storage area. When the main hook of the reactor building crane removes the cask from the cask area, it follows the travel path shown in Figure 9.1-5. In addition, sufficient redundancy is provided in the reactor buildi ng crane such that no credible postulated failure of any crane component will result in the dropping of the fuel cask. See Section 9.1.4.2.2.

Failure of the gates between the reactor well and the spent fuel storage pool is improbable.

However, in the event of this failure, the loss of water from th e storage pool into the reactor well would not uncover the stored spent fuel due to the elevation of the weir wall under the gates. This elevation ensures that sufficient water is retained in the pool to cover the spent fuel.

To avoid unintentional draining of the spent fuel storage pool to levels below that required for adequate shielding of the spent fuel, no inlets, outlets, or drains that would normally permit the pool to be drained are provided. Discharge lines extending below the pool water level are designed to prevent any siphon back flow. Two skimmer surge tanks are provided and are sized to accommodate water displacement due to large items being placed into or removed from the spent fuel storage pool.

See Section 9.1.3 for additional evaluation of continuous cooling capabilities of the spent fuel pool cooling and cleanup (FPC) system.

9.1.2.3.4 Spent Fuel Rack Design Features

The rack, rack modules, and restra ints are all stainless steel, as is the spen t fuel pool liner, to minimize the potential for galvanic corrosion.

Stainless steel has also been shown to be compatible with spent fuel pool water and the stored assemblies.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-20 The fuel rack base is elevated above the floor to ensure adequate flow under the rack in each fuel assembly. Analyses have been performed and show that sufficient flow is induced by natural convection to preclude local boiling in the hottest storage location.

The analyses were based on the following assumptions:

a. The fuel element inlet temperature is the mixed hot design temperature of the pool, b. A hot assembly peaking fa ctor of 1.74 is applied to the core average assembly energy release rate of 5.3 x 10 4 Btu/hr, c. The maximum local peaking factor is 2.49, giving a maximum local heat flux of 1334 Btu/hr-ft 2 ,

d. A film coefficient of 31 Btu/hr-ft 2-°F is based on pure conduction through a stagnant boundary layer at the fuel rod surface,

e. The downcomer region feeds 12 assemb lies in a row, each assumed to be generating the maximum heat rate defined in assumption b, and

f. One dimensional fluid flow analysis applies.

During full core offload with the bulk pool te mperature at a design va lue of 150°F, the mixed temperature of the water exiting from the hottest storage location is less than 181°F. This is 58°F below the local saturation temperature of 239

°F, indicating that adequate margin to bulk boiling exists. Under normal operating conditions, the fuel rod surface temperature calculated on the basis of the heat flux and film coefficient defined above is more than 14°F below the local saturation temperature.

Local boiling is thus precluded.

The fuel racks are designed, constructed, and fabricated with a high degree of reliability and integrity. A list of the industr y design codes and sta ndards used for the spent fuel storage racks is given below.

Design Codes

a. AISC Manual of Steel C onstruction, 7th Edition, 1970,

b. ASME Boiler and Pressure Vessel Code Section III-1971, Nuclear Power Plant Components. (Tables I-2.2, I-5.0, and I-6.0 are used for yield strength values, coefficients of thermal expansi on, and moduli of elasticity),

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-21 c. AISC Specification for the Design, Fabri cation, and Erection of Structural Steel for Buildings, February 12, 1969 and Supplements 1, 2, and 3 (Supplement 3 effective 6/12/74), d. ASTM Specification A 240-75a, Specifi cation for Heat-Resisting Chromium and Chromium-Nickel Stainless Steel Plat e Sheet and Strip for Fusion-Welded Unfired Pressure Vessels,

e. ASTM Specification A 320-74, Specifica tion for Alloy Steel Bolting Materials for Low-Temperature Service,

f. AWS A 5.9, Corrosion Resisting Ch romium and Chromium-Nickel Steel Welding Rods and Ba re Electrodes, and

g. ASME Boiler and Pressure Vessel C ode Section IX-1974, Welding and Brazing Qualification.

9.1.2.3.5 Spent Fuel Storage Facilities Design

The spent fuel storage pool is designed to Seismic Category I requirements to prevent earthquake damage to the stored fuel. The exte rior walls and roof of the reactor building are designed as low-leakage barriers to confine potential airborne ra diation (contamination) within the reactor building and the e xhaust air treatment systems.

Release of radioactive products from damaged or failed fuel in the spent fuel pool will be detected by a high level gamma radiation monitor located in the fuel pool vicinity. This instrument has a range of 100 to 10 6 mrem/hr with remote readout and alarm in the main control room. Backup detection provided by high radioactivity monitors in the reactor building ventilation system exhaus t plenum (see Section 9.4.2) will initiate reactor building ventilation system isolation and operation of the st andby gas treatment system (see Section 6.5.1) to block potential leakage of contaminated air to the environment. See Chapter 15 for radiological considerations.

The spent fuel storage pool contains a minimu m water depth of 22 ft above the top of the irradiated fuel assemblies seated in the spent fuel storage rack

s. A low water level alarm is provided in the control room in the event of loss of pool water.

As a backup, flow alarms are provided in the drain lines to detect leakage in the reactor vessel to drywell seal, drywell to concrete seal, and fuel pool gate. An adequate fuel pool water level can be maintained, even in the unlikely event of a pipe break between the skimmer surge tanks and the FPC system

pumps, since fuel pool discharge to the skimme r surge tanks is by ove rflow only. A pipe break would drain the skimmer surge tank but would not reduce the spent fuel storage level.

A check valve in each supply pipe from the FP C system prevents sipho n back flow to the cleanup system. Provision is also made so that standby service water (S W) can be used as

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-22 backup for fuel pool makeup upon failure of th e normal makeup system.

This connection is capable of supplying enoug h water to prevent the uncovering of the spent fuel. By use of the standby SW as makeup, the fuel pool will be cooled by evaporation of the pool water. The residual heat removal (RHR) system can be ope rated in parallel with the fuel pool cooling system, to remove abnormal heat lo ads in the fuel pool. See Section 9.1.3 for details.

The reactor building below the refueling floor is designed to be tornado proof. Tornado missiles below this elevation cannot impair the structural integr ity of the pool. Missiles which could reach the pool from above are small enough that they could not impair the structural integrity of the pool. Metal siding above the refueling floor blows out during a tornado. All large objects on the refueling floor are secured so that they cannot be carried into the fuel pool.

See Section 3.3 and GE Topical Report, APED-5696, Tornado Protection for the Spent Fuel Storage Pool.

Leak detection channels are provi ded on the concrete sides of the spent fuel storage pool liner.

Surveillance of flow indications from these leak channels pe rmits early determination and localization of any leakage.

9.1.2.3.6 Radiological Considerations

9.1.2.3.6.1 Normal Operation. Three sources of exposure to personnel in the area of the spent fuel storage pool are considered:

a. Direct dose from the stored fuel,

b. Dose from the radionuclides in the spent fuel pool water, and c. Dose from airborne tritium.

Direct dose from the stored fuel is negligible due to the heig ht of water above the storage positions. Calculations indicat e a direct dose from stored fuel of less than 1 x 10-5 mrem/hr.

Most of the personnel exposure in the area of the spent fuel storag e pool comes from the radionuclide inventory in the water.

Estimates of the dose from this source were calculated to be 3.5 mrem/hr during the first refueling, in creasing to 6.9 mrem/h r during the eleventh refueling. These calculat ed values are given in Table 9.1-2. The dose from airborne tritium is limited by spent fuel storage pool water temperature limitations as described in Section 9.1.3 and by the fuel pool exha ust ventilation described in Section 9.4.2.

9.1.2.3.6.2 Radiological C onsequences of Accidents. See Chapter 15.

The storage racks, storage pool and associated equipment, and the reactor building in which the pool is located, are designed to rema in functional through and after an SSE.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-23 The impact of heavy objects on the racks has been considered in the design. The reactor building below the refueling floor is tornado proof and tornado missiles that could reach the pool from above this level are small enough that the structural integrity of the pool (and racks therein) could not be impaired. The drop of a spent fuel shipping cask is precluded by redundancy features of the cr ane and other design featur es discussed in Section 9.1.2.3.3.

9.1.2.3.7 Conclusions

From the foregoing analyses, it is concluded that the spent fuel storage arrangement and design comply with the objectives in Re gulatory Guide 1.13, Revision 1.

9.1.3 SPENT

FUEL POOL COO LING AND CLEANUP SYSTEM

9.1.3.1 Design Bases

The FPC system has been designe d to comply with the objectives in Regulatory Guides 1.13, Revision 1, and 1.26, Revision 3, to the extent specified in the following subsections. The system and equipment are designed to the classifications given in Tables 3.2-1 and 9.1-3.

During normal reactor operation the system is designed to remove decay heat released from the stored spent fuel elemen ts and maintain a specified fuel pool water temperature, water clarity, and water level by accomplishing the following:

a. Minimizing corrosion product buildup and controlling fuel pool water clarity so that fuel assemblies can be ef ficiently handled underwater,

b. Minimizing fission prod uct concentration in the fuel pool water thereby minimizing the release of fission products from the pool to the reactor building environment,

c. Monitoring surge tank water level to thereby maintain a pool water level above the fuel sufficient to pr ovide shielding for normal building occupancy and to control makeup flow rate from th e condensate transf er system, and

d. Maintaining the fuel pool water te mperature below 12 5°F under normal (nonrefueling) operating conditions. The maximum heat load in the fuel pool occurs during refueling. The magnitude of this heat load will vary for each refueling based on the number of irradiated fuel assemblies moved from the reactor core to the fuel pool, the burnup of each fuel assembly, and the decay time of each fuel assembly.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-24 Additionally, the system also provides cleanup for purity and clarity control to the reactor well, dryer-separator pool , and the suppression pool.

During shutdown conditions, the RHR system may be operated to remove the decay heat load.

However, the RHR assist mode may only be credited for spent fuel pool cooling during a full core offload.

In Section 1.3.2 it was indicated that a significant cha nge was made in th e fuel pool cooling system between the time of the PSAR and FSAR. The design basis originally relied on fuel pool boiling following a severe seismic event and provided for safety grade makeup from the standby SW system. In the final design, the design basis of the fuel pool cooling portion of the system was changed to Seismic Category I.

Following a seismic ev ent or major plant disturbance, the system is de signed to prevent fuel pool boili ng and maintain adequate water level in the spent fuel pool by means of the following:

a. Automatic isolation on low fuel pool water level of the Seismic Category I cooling portion of the system from th e Seismic Category II cleanup portion of the system,

b. Remote-manual startup from the cont rol room of redundant , active components of the fuel pool cooling portion of the system, and initiation of safety grade cooling water from the standby SW syst em, to the fuel pool cooling heat exchangers,

c. Remote valve operation, from the control room, to initiate SW cooling to the FPC heat exchangers, and

d. Redundant SW system makeup to the fuel pool (see Section 3.1.2.6.2) and fuel pool level monitoring from the control room.

9.1.3.2 System Description

The FPC system must operate in a variety of conditions. During normal reactor operations the FPC provides for cooling and cleaning of the spent fuel pool containing discharged fuel

assemblies. During refueling outages, the FPC system may be required to provide additional decay heat removal. The spent fuel pool coo ling function can be prov ided by operating both trains of FPC. The RHR B loop may only be credited for FPC assist during a full core offload. The RHR B loop may be operated, but not credited, to provide supplemental cooling as required. The RHR B loop can be cross connect ed to the FPC system.

In addition, with the reactor cavity flooded and fuel pool gates removed, RHR A or B may be lined up to take suction from the surge tanks and discharge to the RPV.

The FPC system flow diagram is shown in Figure 9.1-6.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-25 System performance data are summarized in Table 9.1-5. Major components of the system are summarized in Table 9.1-3. The system is designed to dissipate the fuel pool heat load during normal operation and re fueling conditions.

The FPC system consists of two separate trains, each containi ng a circulating pump and a heat exchanger. The system also contains two filt er demineralize rs and two skimmer surge tanks, as well as the required piping, valves, and instrumentation. Except for the pumps, heat exchangers and filter demineralizers, common piping header s are used. The FPC pumps normally circulate the pool water in a closed loop, taking suction from the surge tanks, circulating the water through a heat exchanger and a filter demineralizer, and discharging it through the diffusers at the botto m of the fuel pool. During refu eling, pool water is discharged through the fuel pool diffusers or through diffusers in the reacto r well. The water flows from the pool surface through skimmer weirs to the surge tanks. Makeup water for the system is normally transferred from the condensate storage tank to a sk immer surge tank to make up pool water losses. The fuel pool pumps and heat exchangers are located in an enclosed room on the 548-ft level of the reacto r building beneath the fuel pool.

Instrumentation is provided for both automatic and remote manual opera tion. Indication is provided in the control room. Surge tank high and low water level switches are provided.

Control of flow to or from the reactor well can be accomplished during re fueling. A fuel pool high/low water level switch sounds an alarm in the control room whenever the level is either too high or too low. The alarm points are fixed at the bottom of the we irs (low) and 1-1/8 in.

above the normal water level (high).

The pumps are controlled from instrument racks on the 522 ft el. and the control room. Pump low suction pressure automatically shuts down the operating pump(s). A pump low discharge pressure alarm annunciate s in the control room a nd starts the standby pump.

A high rate of leakage through the reactor dr ywell refueling bellows assembly, drywell to reactor well seal, or the fuel pool gates is indicated lo cally and is alarmed in the control room.

Failure of the reactor closed cooling water sy stem (RCC), RHR, or a loss of inventory, are addressed in procedures. Th e FPC system operation and RHR/

FPC assist modes of operation are each discussed in the follo wing sections in more detail.

9.1.3.2.1 System Operation

9.1.3.2.1.1 FPC Syst em Cooling Function

The system normally cools the fuel pool by transferring the spent fuel decay heat from the tube side of the two fuel pool cooling heat excha ngers to the RCC. St andby SW can be manually aligned as the alternate heat sink for the heat exchangers (see Section 9.1.3.2.3 for additional

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-26 details). The fuel pool is main tained at or below 125°F duri ng normal plant operations. The fuel pool temperature may rise above this value during refueling activities or during an anticipated operational transient of the loss of one train of the FPC system. The RHR system can also be manually aligned into several configurations to provide supplemental cooling of the fuel pool. One of these RHR configurations is the RHR/FPC assist mode of RHR (see Section 9.1.3.2.2 for additional details).

The maximum heat load is pres ent in the spent fuel pool du ring refueling activities when recently irradiated fuel bundles are discharged from the reactor core to the fuel pool. The magnitude of this heat load is contingent upon the cycle-specific refueli ng activities, i.e., the number of the bundles discharged, the burnup of the discharged bundles, and the decay time of each bundle when it is placed in the fuel pool.

The heat load associ ated with a planned discharge can be calculated with the ORIGEN-ARP computer code assuming a 2% thermal power uncertainty or other accep table means to account for c ode bias and uncertainties.

During refueling activ ities, the fuel pool temperature is managed by controlling the number and schedule of fuel assemblies discharged, controlling the number of heat removal systems in service, and controlling the temperatures of the systems (RCC or SW) used to remove the heat from the FPC heat exchangers.

The fuel pool cooling system was originally desi gned to maintain the pool at a temperature of less than or equal to 125°F during refueling activities with both trains of fuel pool cooling in operation and RCC cooling water at 95°F. The decay heat load assume d for the original design basis (normal offload) was based on the original licensed power of 3323 MW-thermal, a one-year fuel cycle, and a quarter core offload with a 20-day decay period.

Since the original design, the licensed power was increased to 3486 MW-thermal and the operating cycle was revised from a one-year cycle to a two-year cycle.

These changes resulted in an increased heat load in the fuel pool, particularly dur ing refueling outages. For the current design (3486 MW-thermal and a two-year cycle), a 150°F fuel pool temperature limit (see Table 9.1-6) applies to normal refueli ng activities for the scenario of both trains of fuel pool cooling in operation.

The FPC system is also designed to provide sufficient cooli ng for an anticipated operational transient of the loss of one train of the FPC system. For this transient, the maximum bulk fuel pool water temperature is limited to 155°F.

This limit applies to both normal operation and normal refueling activities (i.e., excluding a full core offload).

The system's ability to satisfy these temperat ure limits could be ch allenged based on outage-specific plans or activities. Outage-specific calc ulations are performed, as needed, to ensure acceptance criteria limits are maintained and adequate decay heat removal capability exists. Management of the rate and magnitude of the heat load added to the fuel pool during refueling activities and the temperature of the credited heat sink (i.e., RCC or SW) are considered.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-27 Energy Northwest does not routinely perform full core offloads at Columbia and as such, a full core offload is considered a non-routine evolution. If pe rforming a full core offload, supplemental RHR cooling (RHR/FPC assist mode) is required to maintain fuel pool temperatures below 145°F. The limit for this scenario is 175°F. A single failure scenario involving the loss of one train of the FPC is not postulated for a full core offload. This exemption from a single failure is reasonable based on the expected infrequent performance of a full core offload.

During normal plant operations, the heat load in the fuel pool will be less than the heat load experienced during refueling act ivities. There are three funda mental operational scenarios considered in the design and licensing basis of the FPC. These scenarios are: 1) normal operations with both FPC trains av ailable; 2) normal operations with an anticipated operational transient of the loss of one train of the FPC system; and 3) a design basis LOCA condition. The fuel pool temperature limits for each of these scenarios are presented in Table 9.1-6.

The FPC system is not credited fo r mitigating the consequences of a design basis event. For a postulated LOOP/LOCA event, RCC will be lost (a utomatically load shed s in response to an F or A signal) and one train of the FPC system is lost as the result of an assumed single failure. Standby SW can be manually aligned from the cont rol room to replace the lost RCC. Prior to the postulated LOCA, the SW is ma intained at a temperature of le ss than or equal to 77°F in accordance with the Technical Specifications. During the mitigation of the design basis LOCA, the SW will rise to a pproximately 90°F. The heat up of the fuel pool following a design basis LOCA has been eval uated using a heat load of a pproximately 10.1 MBTUs/hr and the peak temperature is within the acceptance criteria of 175°F. This te mperature is consistent with the design limits of the FPC system.

9.1.3.2.1.2 FPC System Clean-Up Functions

Water purity and clarity in the storage pool, reactor well, and dryer-separator pool are maintained by filtering and demineralizing the pool water through the FPC system filter demineralizers. In addition to fuel pool water demineralization, the system may be used to mix and demineralize the suppression pool water. The suppressi on pool cleanup portion of the FPC may also be used to periodically let down suppression pool water inventory.

To establish a circulating pattern of flow in the reactor well a nd fuel storage pool, the diffusers and skimmer drains are placed to sweep partic les dislodged during refue ling operations away from the work area and out of the pool.

Particulate material is removed from the water by the pressure precoat filter demineralizer units. The finely divided disposable filter me dium is replaced when the pressure drop is excessive or the ion exchange resin is depleted. The spent filter medium is backwashed to the waste sludge phase separa tor tank for processing in the solid radwaste handling system. New filter medium is mixed in a precoat tank and is transferred as a slurry by a precoat pump where

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-28 the solids deposit on the filter elements. The holding pump connected to each filter demineralizer maintains circulation through the filter in the interval between the precoating operation and the return to norma l system operation. A strainer is provided in the effluent stream of the filter demineralizers to limit the migration of the filter material.

The two filter demineralizer units are located separately in sh ielded cells in the radwaste building. Sufficient clearance is provided in th e cells to permit removal of the filter elements from the vessels. Each cell contai ns only the filter demineralizer and its associated piping. All valves are located on the outside of one shielding wall of the cell, together with necessary piping and headers, instrument elements, a nd controls. The FPC cleaning portion of the system is controlled from the radwaste control room.

The flow rate through the filter demineralizers is given by a flow indicator in the radwaste control room located on the demineralizer control panel.

Differential pressure and conductivity instrumentation is provided for each unit to indicate when backwash is required. Suitable alarms, differential pressure indicators, and flow indicators are provided to monitor the c ondition of the filter demineralizers.

There are two sampling points: SP-25A&B, at the effluent from the fuel pool filter demineralizers, FPC-DM-1A&B. Ther e are also sample points: SP

-69, at the fuel pool return line, and SP-24, at common inlet header to th e demineralizers. All four sample points are piped to the sample room. All four samp le points continuously transmit conductivity measurements to a 4-point recorder and also provide grab samples at the nearby associated fume hood.

Weekly fuel pool analyses is performed to ensu re that the water qualit y specifications for the fuel pool are maintained. The water quality parameters are as follows:

Conductivity 3

µS/cm at 25°C Chloride 0.5 ppm pH 5.3-7.5 at 25°C Total insolubles 1 ppm Heavy metals 0.1 ppm

Weekly gamma analyses are performed following fuel load or when irradiated fuel is stored in the pool. Special tests on iodine or other significant radionuclide removal by the fuel pool filter demineralizers will be performed when gross gamma activity levels in the fuel pool exceed 10-3 mCi/cm 3 during normal power operation.

Continuous influent and effluent conductivity for the fuel pool demineralizers are monitored and recorded. A high conductivity effluent alarm setpoint of 1.5

µS/cm is chosen to reflect marginal performance of the demineralizer since they will eventually satura te with air saturated

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-29 water at an equilibrium level of about 1.1

µS/cm. Differential pressure drop is continuously monitored across the filter demineralizers and at a pressure al arm setpoint of 10.5 psid, the units are removed from service and re-precoated with a comb ined filteration/ion exchange media.

9.1.3.2.2 Residual Heat Removal/Fuel Pool Cooling Cleanup Assist Mode Operation

As discussed in Section 9.1.3.2.1 , the RHR loop B may be operated to assist the FPC system, however, the RHR system may only be credited for the decay heat removal function during a full core offload.

The cross connect between RH R and FPC is provided by tw o flanged spool pieces. The installation and removal of the spool pieces is administratively controlled. However, design evaluation of the system found it acceptable to leave them perman ently installed. Procedures direct initiation and operation of the RHR/FPC assist mode a nd provide guidance to mitigate the consequences in the even t of the loss of this mode.

The supplementary cooling provided by RHR/FPC assist is not required the majority of the time. The separation of the two systems is maintained by system operating procedures and administrative controls on the valves needed to isolate the spool pieces. The two valves on the discharge side of RHR-P-2B (RHR-V-104 and FPC-V-119) are ma intained as locked closed valves. The suction side isolation is provided by the locked closed FPC-V-141 and the check valve, RHR-V-105. The initiation of the RHR/FP C assist mode is not a routine evolution. The installation, testing, and operation of this mode will re quire an evaluation of plant

conditions and requirements.

While operating in RHR/FPC assist mode, the RHR B loop will not be able to provide the low-pressure cool ant injection (LPCI), s hutdown cooling (SDC), or containment spray functions w ithout local manual operations.

While in RHR/FPC assist, RHR and FPC take a suction on the skimmer surge tanks. Water directed to RHR B loop is cool ed by SW and the FPC system prov ides for cleaning of the fuel pool water. As in normal FPC operation, the water is directed to the surge tank by skimmers near the top of the spent fuel pool. The cool ed and cleaned water is returned to the pool through the 8-in. diffusers normally used by the FPC system.

To limit RHR flow rates while operating in FPC assist, two restricting orifices are installed in the FPC system. While in the FPC assist mode, an optional, additional return path for the RHR B discharge is into the reactor cavity through the line normally providing RPV head spray (Section 5.4.7.2.1). During refueling outage conditions the head spray line is disconnected at a flange in the cavity and removed with the reactor head insulation package. Normally this flange is then blanked off. However, for the additional cooling, the blank can be left off providing the additional flow path of RHR to the cavity, wi th a rate of up to 1000 gpm.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-30 9.1.3.2.3 Fuel Pool Cooling System Opera tion Following a Seismic Event or Major Plant Disturbance The portion of the FPC system that is required for cooling the fu el pool is located within the reactor building and is designed to Seismic Category I criteria. The portion of the FPC system which is used for fuel pool cleanup is located within the radwaste building and is isolable from the reactor building by means of two Seismic Category I isola tion valves per line located within the reactor building. The isolation valves are either ch eck valves or motor-operated gate valves. The motor-operated valves close automatically on a fuel pool low water level condition.

The redundant, active components required for fuel pool cooling are powered from Division 1 and 2 power sources. Following a loss of offsite power, these compone nts can be energized from onsite emergency power. On a loss of RC C to the fuel pool heat exchangers, the RCC supply lines to the heat exchangers can be isolated by redundant Se ismic Category I motor-

operated gate and check valves and the return lines by re dundant Seismic Category I motor-operated gate valves. Standby SW can be supplied to the h eat exchangers through motor-operated valves which are normally key locked closed. Radiation detectors are located on the SW return lines. The SW system can also be used to mitigate inventory reduction.

To preclude leakage of servi ce water into the spent fuel pool during operation of the SW system or leakage of fuel pool cooling water in to the SW system when the SW system is not operating, the manual valves (SW-V-75AA and SW-V-75BB) and the motor-operated valves (SW-V-75A and SW-V-75B) are kept normally clos ed when spent fuel pool temperatures are below 138°F. If during normal plant operations the spent fuel pool te mperature rises above 138°F, the manual valves will be maintained open. The manual valves are located on the west side of the 522-ft el. of the reactor building and are accessible and can be opened if necessary following a LOCA prior to spent fuel pool temperatures exceeding 175°F. The access route to

these manual valves is shown in Appendix J. Once the manual valve(s) are opened, spent fuel pool level can be maintained us ing the remote-manual valves from the main control room.

Operation and monitoring of fuel pool cooling portion of the FP C system can be done entirely from the control room.

All components required for fuel pool cooling are qualified to the r eactor building accident environment. The fuel pool equipment room located on the 548 ft el. is provided with redundant Seismic Category I room coolers with one each on Di vision 1 and 2 power sources.

9.1.3.3 Safety Evaluation

The maximum heat load is the decay heat of one full core load of the fu el plus the remaining decay heat of previously discharged fuel assemblies. The RHR B loop is operated in parallel

with the FPC system during this condition or other events which require supplemental cooling.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-31 To permit this use of the RHR system, normally locked closed valves interconnecting the RHR system to the FPC system are opened after verifi cation of the installatio n of the flanged spool pieces. A full core offload is a non-routine evolution. A single failure is not postulated to occur during this evolution and credit for RHR/FPC assist is allowed. Du ring this evolution, sufficient cooling should be provide d to maintain the fuel pool temperature at less than 145°F.

The licensing basis limit is 175°F.

The fuel pool heat exchangers are normally cooled by the RCC system to contain released radioactivity in the event of a fuel pool heat exchanger tube fa ilure. During normal operations, the system maintains th e fuel pool water temperature below 125°F when removing the nominal heat load from the fuel pool. During normal refueling outage conditions, the system is evaluated to ensure it has the capability to maintain the fu el pool water temperature below 150°F. This limit applies to core offloads up to, but not includi ng a full core offload, and is based on the assumption of 2 FPC trains in serv ice. The system maintains fuel pool water temperature below 155°F in the event that only one pump and one heat exchanger are available. Depending on the heat load in the pool during refueling ac tivities, RCC or SW temperature is controlled to ensure the FPC system can perform its design functions within the acceptance criteria shown in Table 9.1-6.

Following a seismic event or major plant disturbance the SW system is available to cool the fuel pool (by means of FPC or RHR-B heat exchangers) to preclude boiling of the fuel pool water. The SW pressure is hi gher than the fuel pool pressure; thus, any leakage will be into the fuel pool system. In addition, radiation monitors of the SW return line detect any gross failure in the heat exchangers.

The fuel pool design precludes a ny condition which could allow th e fuel pool to be drained below the pool gate between the reactor well a nd the fuel pool. Two diffusers are placed in both the reactor well and the fuel pool to distribut e cooled return water efficiently. Diffusers are placed to minimize stratification of either temperature or contamination. Valving is provided to prevent water from being siphoned out of the pool. All piping connected to the fuel pool and reactor well, except for drains and liner drains, are Se ismic Category I, including any normally closed manual or normally open automatic valves that provi de isolation from the Seismic Category II portion of the system. Drain and liner drain piping connected to the fuel pool, reactor well, and dryer se parator pool are Seismic Cate gory II supported to Seismic Category I requirements. Since the fuel pool system is at low temperature and pressure (moderate energy system) postulated breaks in the Seismic Category I portion are limited to cracks.

Fuel pool cooling can be established and monitored from the c ontrol room following a design basis LOCA. One of the two FPC trains is ad equate to prevent fuel pool boiling by a large margin. However, during normal plant operatio n, one or both trains operate to maintain 125°F pool water. Should one of the trains be unavailable, the se cond train operates to maintain pool water temperat ure below 155°F. The 155°F va lue is applicable for an

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-32 anticipated operational transient i nvolving the loss of one FPC loop. If the fuel pool heat load is such that the resulting temperature transient exceeds 125°F, pr ocedural guidance is in place to restore the fuel pool temperature to < 125°F.

Due to the large thermal capacity of the fuel pool sufficient operator time is available for the operator to take necess ary corrective action to supplement cooling.

The results of bounding fuel pool cooling analyses are provided in Table 9.1-6. These analyses demonstrate that the FPC system w ill maintain the pool temp erature below system design limits and regulatory acceptance criteria.

A makeup water valve cont rolled by skimmer surge tank level switches supplies water from the condensate transfer system to the fuel pool to replace losses. The backup source of makeup water is from the Seismic Cate gory I, safety class 3 SW system. This source supplies makeup for long-term pool water losses.

Each filter deminerali zer is capable of continuous operation at a normal fuel pool water flow rate of 575 gpm or a maximum fuel pool water fl ow rate of 1000 gpm a nd will maintain water conditions as specified in Section 9.1.3.2.

A radiological evaluation of the cleanup system is presented in Chapter 12.

From the foregoing analysis, it is concluded that the FPC system meets its design basis and satisfies the requirements of Regulatory Guide 1.13, Revision 1.

9.1.3.4 Testing and In spection Requirements

Special testing is not required for the system except as noted below b ecause, when fuel is stored in the pool, at least one pump, heat exch anger, and filter demineralizer are routinely in operation.

Routine, periodic visual inspec tion of system components, inst rumentation, and alarms are adequate to verify system operability. Likewise, the interconn ecting valves between the FPC system and the RCC, SW, and RHR systems are periodically inspected to verify their operability.

9.1.4 FUEL HANDLING SYSTEM

9.1.4.1 Design Bases

The fuel handling system is designed to provide a safe and effective me ans for transporting and handling fuel from the time it reaches the plant until it leaves the plant after postirradiation cooling. Safe handling of fuel includes design considerati ons for maintaining occupational radiation exposures as low as reasonably achievable during transportation and handling.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-33 Design classification criteria for major fuel handling system equipment is provided in Table 3.2-1 which lists the safety cla ss, quality class, quality gr oup, and seismic category for the equipment.

The cask storage area in the spent fuel pool and the cask ha ndling facilities are designed to accommodate a spent fuel cask of 125 tons. The reactor building crane is designed to transport the spent fuel cask between the cask receiving area, the cask storage area in the spent fuel pool, and the cask washdown area without indu cing a permanent deformation of any crane element. Adequate safety features are desi gned into the reactor building crane system and controls to ensure that no credible postulated failure of any crane component will result in the dropping of the fuel cask.

The transfer of new fuel as semblies between the uncrating area and the new fuel inspection stand and/or the new fuel storage vault is accomplished using the jib crane or the reactor building crane auxiliary hoist.

The refueling floor jib cranes are used to transfer new fuel from the new fuel vault to the fuel storage pool. From this point on, the fuel wi ll be handled by the telescoping grapple on the refueling platform.

The refueling platform is General Class G and Seismic Category 1M from a structural standpoint in accordance with 10 CFR Part 50, Appendixes A and B. Allowable stress due to SSE loading is 120% of yield or 70% of ultimate, whichever is less.

A dynamic analysis is performed on the structures using the response spectrum method with load contributions resulting from each of three components earthquake motion being combined by the RMS

procedure.

Working loads of the platform structures are in accordance with the AISC Manual of Steel Construction. All parts of the hoist systems are designed to have a safety factor of five, based on the ultimate strength of the material. A redunda nt load path is inco rporated in the fuel hoists so that no single component failure could result in a fuel bundle drop. Maximum deflection limitations are imposed on the main structures to main tain relative s tiffness of the platform. Welding of the platform is in accordance with AW S D14-1. Gears and bearings meet AGMA Gear Classification Manual and ANSI B3.5. Materials used in construction of load bearing members are to ASTM specifications. For personnel safety, OSHA Part 1910-179 is applied. El ectrical equipment and controls meet ANSI-ANS C1, National Electric Code, and NEMA Publica tions No. IC1 and MG1.

The general purpose grapples and the main te lescoping fuel grapples have redundant hooks.

The main telescoping fuel grapple has an indicator which conf irms positive grapple engagement.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-34 The fuel grapple is used for lifting and transporting fuel bundles. It is designed as a telescoping grapple that can extend to the proper work level and in its fully retracted state still maintains adequate wate r shielding over fuel.

In addition to redundant electric al interlocks to preclude the po ssibility of raising radioactive material out of the water, the cables on the aux iliary hoists incorporate an adjustable, removal stop that will jam the hoist cable against the platform structure to prevent hoisting when the free end of the cable is at a preset distance below water level.

Sufficient redundancy is provided in the reactor building crane such that no credible postulated failure of any crane component will result in dr opping of the fuel cask and rupturing the fuel storage pool. Furthermore, limitation of the travel of the crane handling the cask will preclude transporting the cask over any fuel storage rack. See Chapter 15 for accident considerations.

9.1.4.2 System Description

Table 9.1-7 is a listing of typical tools and servicing equipmen t supplied with the nuclear system. The following paragraphs describe the use of some of the major tools and servicing equipment and address safety aspects of the design, where applicable.

9.1.4.2.1 Spent Fuel Cask

The designs of cask storage a nd dry cask handling facilities are based on a design cask weighing approximately 125 tons and being appr oximately 17 ft long by 8 ft in diameter.

The following description of the spent fuel cask is based on the licensed Holtec HI-STORM 100 System. That system include s a canister that contains the fuel and a transfer cask that contains the canister. The canister confines the fuel. The transfer cask provides shielding and structural protection of the canister during canister loading or movement.

Overland offsite transportation of the cask will conform to tran sportation rules and regulations, 49 CFR Part 173.

9.1.4.2.2 Reactor Building Crane

9.1.4.2.2.1 Description. The main purpose of the reactor building crane is to handle the spent fuel cask between the cask receiving area, the cask load ing area in the spent fuel pool, and the cask decontamination wa shdown area. Secondary pur poses of the crane include servicing and refueling the reactor pressure vesse l, and handling of equipment or parts thereof received or shipped through the loading facility in the reactor building.

The reactor building crane is a single-trolley top-running electric overhead traveling crane with a 125-ton capacity main hoist, a nd a span of approximately 126 ft. The general arrangement

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-35 of the crane in the reactor building is shown in Figures 1.2-11 and 1.2-12. The crane is Class A1 as defined for nucl ear fuel handling by the Cran e Manufacturers Association of America Specification No. 70 for Electric Over head Traveling Cranes , (CMAA No. 70). The reactor building crane is designed, fabricated, installed, and tested in accordance with ANSI Standard B30.2, Safety Code for Cranes , Derricks and Hoists, and CMAA Specification No. 70. The crane is Seismic Category I.

Operation of the reactor building crane is from the cab or from the floor by radio control. The crane radio control system uses crystals highly tuned to a narrow frequency band, thereby precluding interferences from othe r signaling systems. Control at any one time is from one point only.

The structure of the crane bridge consists of welded box type girders with truck saddles and truck frames of welded-steel c onstruction. The trolley side frames, sheave frames, and truck frames are of structural steel-welded construction.

The rated full-load capac ities, lifts, and full-load speeds are as follows:

Main Hook Rated full-load capacity, tons 125 Hook travel, ft 190 ft 0 in.

Hoisting speed, fpm at full load 5.5

Auxiliary Hook

Rated full-load capacity, tons 15 Hook travel, ft 199 ft 8 in.

Hoisting speed, fpm at full load 20

Travel Speeds

Trolley, fpm at full load 40 Bridge, fpm at full load 50 The crane is capable of raising, lowering, holding, and transpor ting a test work load at 125% of rated load without damage to any parts and without induci ng a permanent deformation of any crane element.

Structural members not covered by the CMAA Specification No. 70 are designed in accordance with the AISC Specifica tion for the Design, Fabrication, and Erection of Structural Steel for Buildings (September 1972).

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-36 All parts of the crane are designed to resist the maximum stress es caused by loading combinations called for in the CMAA Specification No. 70 and for the following loading combinations:

a. Dead load + trolley weight and rated load + SSE, not to exceed 90% of the yield strength,

b. Dead load + weight of unloaded tro lley + tornado loading, not to exceed 90%

of the yield strength, and

c. Rated breakdown torque of the motors, not to exceed allowable stresses.

The lateral loads on the crane runway, in the normal operating condition, are taken as 10% of the sum of the weights of the hook load and the crane trolley, plus th e seismic load of the crane dead load, acting simultaneous ly at the top of each rail in either direction normal to the runway rails.

Safety factors for the main hoist cables are discussed in Section 9.1.4.2.2.2. The main hook and auxiliary hook are capable of sustaining the full-rated load with a design safety factor of five.

9.1.4.2.2.2 Safety Features. The single hoist motor drives two separate shafts. The motor has two centrifugally tripped limit switches, one outboard of each hoist input pinion at each end of the motor shaft assembly. These provide an automatic safety shutdown and protection from any control or motor malfunction which might result in a runaway condition of the load.

Each motor driven shaft passes through a 150% capacity solenoid-act uated brake. A failure of either the motor shaft, the connecting shafts, or the shaft couplings woul d not result in a load drop since the redundant dual solenoid actuated brak es would be effective in holding the load.

On loss of power to the motor, both brakes engage. They can also be engaged by the operator. Additionally, there is a 90% capacity eddy current brake to limit the rate of load lowering.

After the brake, each motor shaf t enters its own gear reducer. If a component of one gear case (gear teeth, shafts, bearings, or structural component) should fa il, the other gear reducer will hold the load with its brake. The brake is designed with a safety factor of five.

Each gear case is fitted on its output end with a pinion meshing with the drum gear. A failure of a pinion, drum gear, pinion shaft, or pinion bearing will result in the load being carried by the other similar set of parts on the other end of the drum. E ach functioning pa rt is designed with a safety factor of five. In each of the main hoist gear cases, there is a mechanical load brake with cooling of the gear case oil to offer additional safety in loading handling.

In the event of failure of the drum shaft, drum bearing, or drum bearing bracket, the drum flange will drop a fraction of an inch onto stru ctural seats so located that the drum is

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-37 supported. The remaining pinion and gear will stay in mesh to restrain the load. Again, a safety factor of five is incorporated into th e design of these parts.

Two separate wire ropes are led from the main hoist drum, each being reeved through a set of sheaves, upper and lower (block sheaves), and back to an equalizer bar arranged for equal division of the load between the two ropes.

The individual wire rope safety factors for maximum static load and static plus inertia load (dynamic) are 9.7 and 9.

0 respectively. The equalizer bar is fitted with dou ble acting springs to minimize th e shock when the entire load is transferred to one rope. Theref ore, load drop is precluded.

To protect against overloading of the cables and to provide indicat ion of load balancing, a load sensing system consisting of tension type load cells is instal led in each of the hoist cables at its connec tion to the equalizer bar assembly under the trolley.

All sheaves, both upper and bloc k sheaves, are contained in h eavy structural casings which usually carry a negligible load. In the event of a sheave pin fa ilure, the sheaves would rise to the top of the block or drop to the base of the upper sheave housing and stop at those points. Thus a load drop is precluded. The block asse mbly contains two 100% capacity load carrying devices consisting of a sister hook and an eye hook. This redundancy, in attachment to lifting assembly and in load carrying capability, is such that a single fa ilure will not cause load drop.

Additional non-destructive examination (NDE) (ultrasonic and magnaflux testing for the load block swivel and the sheave shafts of the upper assembly) provide further assurance that this

crane is of a quality suitable fo r nuclear services. Sufficient el ectrical circuitry is provided such that no single credible electrical component failure wi ll cause the lo ad to drop.

To preclude any dislodgement of the crane bridge girder and trolley system during seismic or tornado excitation, the following is provided:

a. From the bridge trucks, lugs or brackets are attached to the truck frame to limit total crane drop to 1 in. or less should a wheel or axle break, and

b. The trolley is provided with latches to engage rack s attached to the bridge girder; the bridge trucks are provided with latches to engage racks attached to the crane runway girders. The latches and racks ensure that the trolley is rigidly attached to the crane runway girders. These provisions are used wh en the crane is not operational.

9.1.4.2.3 Fuel Servicing Equipment

The major fuel servicing equipment described below has been designed in accordance with the classification cr iteria listed in Table 3.2-1.

9.1.4.2.3.1 Fuel Preparation Machine. The fuel preparation machine, Figure 9.1-7 , is mounted on the wall of the fuel storage pool and is used for stripping re usable channels from

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-38 the spent fuel and for rechanneling of the new fuel. The machine is also used with the fuel inspection fixture to provide an underwater inspection capability, and with the defective fuel storage container to contain a defective fuel assembly for stripping of the channel.

The fuel preparation machine consists of a work platform, a frame, a nd a movable carriage.

The frame and movable carriage ar e located below the normal water level in the fuel storage pool, thus providing a water shield for the fuel assemblies being handled. The fuel preparation machine carriage has an up-travel-stop to prevent raising fuel a bove the safe water shield level.

The movable carriage is operated by a foot pedal controlled air hoist.

9.1.4.2.3.2 New Fuel Inspection Stand. The new fuel inspection stand, Figure 9.1-8 , serves as a support for the new fuel bundles undergoing receiving inspection and provides a working platform for technicians engage d in performing the inspection.

The new fuel inspection stand consists of a vertical guide column, a lift unit to position the work platform at any desired le vel, bearing seats and upper clamps to hold the fuel bundles in position.

9.1.4.2.3.3 Channel Bolt Wrench. The channel bolt wrench, Figure 9.1-9 , is a manually operated device approximately 12 ft in overall length. The wren ch is used for removing and installing the channel fastener asse mbly while the fuel as sembly is held in the fuel preparation machine.

The channel bolt wrench has a socket which mates and captures the channel fastener capscrew.

9.1.4.2.3.4 Cha nnel Handling Tool. The channel handling tool, Figure 9.1-10 , is used in conjunction with the fuel preparation machine to remove, install, and transport box fuel channels in the fuel storage pool.

The tool is composed of a handling bail, a lock/release knob, extensi on shaft, angle guides, and clamp arms which engage the fuel channel. The clamps are actuated (extended or

retracted) by manually rota ting lock/release knob.

The channel handling tool is suspended by its bail from a spring balancer on the channel handling boom located on the fuel pool periphery.

Similar tooling is provided as appropr iate by the current reload fuel vendor.

9.1.4.2.3.5 Fuel Pool Sipper. The originally supplied equi pment is not used. Any fuel sipping is performed by the refueling vendor using their own equipment and procedures. Fuel sipping heads, panels, and containers are separate pieces of equipment used for out-of-core wet sipping at any time. They are us ed to isolate a fuel bundle while circulati ng water through the C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-39 fuel bundle in a closed system. The containers cannot be used for transporting a fuel bundle.

The bail on the container head is designed not to fit into any of the grapples.

9.1.4.2.3.6 Fuel Inspection Fixture. The fuel inspection fixture, Figure 9.1-11 , is used in conjunction with the fuel preparation machine to permit remote inspection of fuel elements.

The fixture consists of two parts: (1) a lower bearing asse mbly, and (2) a guide assembly at the upper end of the carriage.

The fuel inspection fixture perm its the rotation of the fuel assembly in the carriage, and, in conjunction with the vertic al movement of the carriage, provides complete access for inspection.

9.1.4.2.3.7 Channel Gauging Fixture. The originally supplied equipment is not used. Any channel gauging is performed by the refuel ing vendor using their own equipment and procedures.

9.1.4.2.3.8 General Purpose Grapple. The general purpose grapple, Figure 9.1-12, is a handling tool used generally with the fuel. The grapple can be attached to the jib crane and the auxiliary hoists on the refueling platforms.

The general purpose gr apple or ot her rigging controlled by plant procedures ma y be used to remove new fuel from the fuel container, place it in the inspection stand or vault, and transfer it to the fuel pool. It can be used to handle new fuel during channeling.

9.1.4.2.4 Servicing Aids

General area underwater lights are provided with a suitable reflector for illumination. Suitable light support brackets are furnishe d to support the lights in the reactor vessel and to allow the light to be positioned over the ar ea being serviced independent of the platform. Local area underwater lights are small diameter lights for a dditional illumination. Drop lights are used for illumination where needed.

A radiation hardened designe d portable underwater closed circuit television camera is provided. The camera may be lowe red into the reactor vessel and/

or fuel storage pool to assist in the inspection and/or maintenance of these areas. The camera is also equipped with a right angle lens to allow viewing at 90 degrees.

A general purpose, plastic viewing aid is provided to float on the water surface to permit better visibility. The sides of the view ing aid are brightly colored to allow the operator to observe it in the event of filling with water and sinking. A portable, submersible type, underwater vacuum cleaner is provided to assist in rem oving crud and miscellane ous particulate matter from the pool floors, or the reactor vessel. The pump and the filter unit are completely

submersible for extended periods. The filter "package" is capable of being remotely changed, and the filters will fit into a standard shipping container for offsite burial. Fuel pool tool accessories are also provided to meet servicing requirements.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-40 9.1.4.2.5 Reactor Vessel Servicing Equipment

The following is a description of the equipment designs.

9.1.4.2.5.1 Reactor Vessel Service Tools. These tools are used when the reactor is shut down and the reactor vessel head is being removed or reinstalled. Tools in this group are

a. Stud handling tool, b. Stud wrench, c. Nut runner, d. Stud thread protectors, e. Thread protector mandrel, f. Bushing wrench, g. Seal surface protector, h. Stud elongation measuring rods, and

i. Head guide cap.

These tools are designed for a 40-year life in the specified environment. Lifting tools are designed for a safety factor of five or better with respect to the ultimate strength of the material used. When carbon steel is used, it is either hard chrome plated, parkerized, or coated with an acceptable paint.

9.1.4.2.5.2 Steam Line Plug. The steam line plugs are used during reactor refueling or servicing; they are inserted in the steam outlet nozzles from inside of the reactor vessel to prevent a flow of water from the reactor well into the main steam lin es during servicing of safety/relief valves, main steam isolation valves, or other components of the main steam lines. (The reactor water level is raised to the refueling level during servicing.)

The steam line plug design provides two seals of di fferent types. Each one is independently capable of holding full head pressure. The equipment is constructed of noncorrosive materials.

All calculated safety fact ors are five or greater.

9.1.4.2.5.3 Shroud Head Bolt Wrench. This is a hand held tool for operation of shroud head bolts. It is designed for a 40-year life, and it is made of aluminum to be easy to handle and to resist corrosion. Testing has been performed to confirm the design.

9.1.4.2.5.4 Head Holding Pedestal. Three pedestals are provided for mounting the reactor vessel head on the refueling floor. The pedestals have studs which enga ge three ev enly spaced stud holes in the head flange. The flange su rface rests on replaceable wear pads made of aluminum. When resting on the pedestals, the head flange is approximately 3 ft above the floor to allow access to the seal surface for inspection and O-ring replacement.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-41 The pedestal structure is a carbon steel weldment, coated with an approved paint. It has a base with bolt holes for mounting it to the concrete floor. The stru cture is designed in accordance with "The Manual of Steel Construction" by AISC.

9.1.4.2.5.5 Head Nut and Washer Racks. The RPV head nut and washer racks are used for transporting and storing up to six nuts and wa shers. The rack is a box-shaped aluminum structure with dividers to provide individual compartments for each nut and washer. Each corner has a lug and shackle for attaching a four-leg lifting sling.

The racks are designed in accordance with the "Aluminum Construction Manual" by the Aluminum Associations and fo r a safety factor of five.

9.1.4.2.5.6 Head Stud Racks. The head stud racks are used for transporting and storage of up to six reactor pressure vessel studs. They are su spended from the aux iliary building crane hook when lifting studs from the reactor well to the operating floor.

The racks are made of aluminum to resist corrosion.

9.1.4.2.5.7 Dryer and Separator Slings. The dryer and separator slings are lifting device assemblies used for transporting the steam dryer or the steam separator between the reactor vessel and the equipment pool. E ach sling assembly consists of a cruciform shaped structure which is suspended from a hook box with four slings. The hook box engages the reactor service crane sister hook, with two hook pins, and the hook lif ting eye with one pin. On the end of each arm of the cruciform is a socket with a pneumatically operated pin for engaging the four lift eyes on the st eam dryer or shroud head.

Each sling assembly has been designed such that one hook pin and one main beam of the cruciform is capable of carryi ng the total load and so that no single component failure will cause the load to drop or swing uncontrollably out of an essentially level attitude.

The safety factor of the lifting members is five or better in reference to the ultimate breaking strength of the material. The structure is designed in accordance with "The Manual of Steel Construction" by AISC. The completed assembly is proof tested at 125% or greater of rated load and all structural welds are magnetic particle inspected after load test.

9.1.4.2.5.8 Head Strongback. The RPV head strongback is used for lifting both the pressure vessel head and the dryw ell head. It is a cr uciform shape with four equally spaced lifting points on the ends of the arms. In the center it has a hook box which enga ges with two pins to the reactor service cran e sister hook and one pin to the hook lifting eye.

The strongback is designed such that one leg of the cruciform will support the rated load and such that no single component failure will cause the load to drop or swing uncontrollably out of an essentially level attitude. The structure is designed in accordance with "The Manual of

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-42 Steel Construction" by AISC. All welding is in accordance with the ASME Boiler and Pressure Vessel (B&PV) Code Section IX. A safety factor of five or greater in reference to the ultimate material strength is us ed for the design. The complete d assembly is proof tested at 125% rated load. After the load test, all structural welds are magnetic pa rticle inspected.

9.1.4.2.5.9 Steam Line Plug Installation Tool. The steam line plug installation tool is suspended from the building crane auxiliary hook or refueling platform monorail or auxiliary hoist for transporting and installing the steam line plugs in the steam line nozzles of the reactor vessel. This tool is designed for a safety factor of five.

9.1.4.2.5.10 Auxiliary Work Platform. The auxiliary work plat form provides an alternate work space over the reactor cavity and equipmen t pool for maintenance work. The platform spans the reactor cavity and travel s on the same rails as the refueling platform. The platform is equipped with two jib cranes, two electrica lly operated jib hoists, three jib crane support pedestals, and is driven by electric motors. To increase accessibility for maintenance work above the cavity, the cente r floor section of the platform is removable.

Positioning of the platform over the reactor cavity is controlled by administrative procedures.

The auxiliary work platform is designed structur ally to meet Seismic Ca tegory I requirements. The structural design is in accordance with "The Manual of Steel Construction" by AISC.

Materials are in accordance with ASTM Standards. The structure is coated with an approved paint.

The cranes are exempted from the NUREG 0612 requirement because the definition of heavy loads at CGS is 1200 lb. Each 500 lb hoist is equipped with a load limiting device to prevent lifting fuel bundles or unauthori zed heavy loads. Interlocks on the platform or hoists are not provided since the hoists and plat form are controlled administrativ ely and will not be used for reactivity manipulation.

9.1.4.2.5.11 Cavity In-Vessel Service Platform (CISP). The CISP, Figures 9.1-17 and 9.1-18, serves as a circular H-shaped platform with an inside diameter equal to that of the reactor vessel. The CISP has an opening which may be aligned with the transfer canal (cattle chute). The CISP is used to allow personnel access over the r eactor vessel annulus, between the vessel wall and the core shroud, while core alterations are in process using the refueling platform. The CISP can be orie nted to allow access to either the spent fuel pool or the dryer-separator pit. The CISP is attached to three support frame le gs which rest on the upper cavity shield block ledge. The platform is partially submerged with th e walkway floor at a depth of approximately 16 in. when the cavity is flooded up for refueling operations. The CISP also has a 750 lb capacity electric hoist which may be positioned around the inside diameter of the CISP.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-001 9.1-43 The CISP was designed and constructed to Seismic I criteria and all lifting components meet NUREG 0612 requirements for a singl e failure proof lift. The plat form is designed to meet the requirements of the AISC Manual of Steel Construction and the NUREG 0800, Standard

Review Plan (SRP) Section 3.8.4. It may be installed in the reactor cavity after the steam separator is removed from the ve ssel, the cavity is flooded up to normal refueling level and the fuel pool gates are removed. The CISP is removed from the cavity prior to the commencement of vessel restoration.

During movement of the CISP to and from th e temporary lay-down area (north of the spent fuel pool), a small portion of the CISP must pass over the spent fuel pool in order to clear the stack monitoring equipment along the center of the refueling floor north wall. The lifting of the CISP is accomplished by the single failure proof reactor building crane, and with NUREG 0612 compliant rigging and handling equipmen

t. Therefore drop of the CISP into the spent fuel pool is not a credible accident. In addition the distance that the CISP can extend over the spent fuel pool is administratively c ontrolled such that only a small portion of the CISP passes over the spent fuel pool and the center of gravity of the CISP is at least 7.5 ft north of the spent fu el pool north curb.

9.1.4.2.5.12 Head Stud Te nsioner Carousel/Strongback. The RPV head stud tensioner carousel/strongback is a lifting device that provides an alternate means to remove and install the RPV head. It combines th e functions of the original st ud tensioner handling frame, RPV head lifting strongback, and head nut and washer racks. The design allows the carousel/strongback to be releas ed from the reactor building crane once it has been mounted onto the four lifting lugs of the RPV head.

The head stud tensioners , which ride on the carousel/strongback's integral monorail, are po sitioned to effect head stud nut removal and reinstallation. The nuts and washers for all 76 of the RPV head studs can be stowed on the carousel/strongback's head stud nut storage ring. The RPV head, head st ud nuts and washers, and carousel/strongback (with the RPV head attached) are together lifted by the reactor building crane.

The head stud tensioner carousel/strongback was designed, fabricated, and te sted to meet the requirements of NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants" and the guidelines of ANSI N14.6, "A merican National Sta ndard for Special Li fting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More for Nuclear Materials."

9.1.4.2.6 In-Vessel Servicing Equipment

The instrument strongback attached to the reactor building cran e auxiliary hoist is used for servicing neutron monito r dry tubes should they require replacement. The strongback initially supports the dry tubes into the vessel. The in-core dry tube is then decoupled from the strongback and is guided into place while being supported by th e instrument handling tool. Final in-core insertion is accomplished from below the reactor vessel. The instrument handling tool is attached to the refueling platform monorail hoist and is used for removing and C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-44 installing fixed in-core dry tube s as well as handling neutrons source holders and the source range monitor/intermediate range monitor dry tubes.

Each in-core instrumentation gui de tube seals by metal-to-metal contact with the guide tube flange seal. In the event that the seal needs replacing, an in-core guide tube seal tool is provided. The tool is inserted into an empty guide tube and sits on the beveled guide tube entry in the vessel. When the drain on the water seal cap is ope ned, hydrostatic pressure seats the tool. The flange can then be removed for seal replacement.

The auxiliary hoists on the refueling platform are used with appropriate grapples to handle control rods, fuel suppo rt pieces, double blade guides, flux monitor dry tubes, sources, and other internals of the reactor. Interlocks on th e fuel grapple hoist and both auxiliary hoists are provided for safety purposes; the refueling interlocks are described and evaluated in Section 7.7.1.13.

The Westinghouse Multi-Lift Tool (MLT) is used exclusively with the Monorail Auxiliary Hoist. The MLT is used to remove and replace a control rod blade with its associated fuel support piece and double blade guide. The MLT ma y also be used for uncoupling the control rod drive mechanism from the c ontrol rod blade to be remove

d. The monorail hoist backup emergency stop block position is procedurally controlled to accommodate the length of the MLT and grappled components.

9.1.4.2.7 Refueling Equipment

Fuel movement and reactor serv icing operations are performed from a platform which spans the refueling, servicing, and storage cavities.

9.1.4.2.7.1 Refueling Platform. The refueling platform is a gantry crane that is used to transport fuel and reactor components to and from pool storage and the reactor vessel. The platform spans the fuel stor age and vessel pools on rails be dded in the refueling floor.

A telescoping mast and grapple susp ended from a trolley system is used to transport and orient fuel bundles for core, storage rack , or spent fuel cask placement.

Control of the platform is from an operator station on the main trolley with a position indicating system provided to position the grapple over core locations. The plat form control system in cludes interlocks to verify hook engage ment and grapple load, prevent uns afe operation over the vessel during control rod movements, and lim it vertical travel of the grapple. Two 1000-lb capacity auxiliary hoists, one main trolley mounted and one auxiliary trolley mounted, are provided for servicing such as local powe r range monitor (LPRM) replacement, fuel support replacement, jet pump servicing, and control rod replacement. The grapple in its normal up position provides 7 ft 6 in. minimum water shieldi ng over the active fuel during transit.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-45 9.1.4.2.8 Storage Equipment

Specific storage equipment is used for new, used , or spent fuel, for cont rol rods, and for other core components.

Specially designed fuel storage racks are provided in the spent fuel pool and in the new fuel vault. Additional storage equipment is listed in Table 9.1-7. For fuel storage racks description and fuel a rrangement, see Sections 9.1.1 and 9.1.2. 9.1.4.2.9 Under Reactor Vessel Servicing Equipment The primary functions of the under reactor vessel servicing equipment are to (a) remove and install control rod drives , (b) service thermal sleeve and control rod guide tube, and (c) install and remove the neut ron detectors.

Table 9.1-7 lists the equipment and tools required for servicing. Of the equipment listed, the equipment handling platform and the control rod drive handling equipment are powered electrically.

The control rod drive handling equipment is used for the removal and installation of the control rod drives from their housings. This e quipment is designed in accordance with the requirements of National Electrical Manufacturers Association (NEMA, MG1: Motor and Generator Standards), American National Sta ndards Institute Standa rds (ANSI C1, National Electric Code), Occupational Safety and Health Administ ration (OSHA, 1910.179), and American Institute of Steel Construction (AICS, Manual of Steel Constr uction). All lifting components are equipped with adequate brakes or gearing to prevent uncontrolled movement on loss of power or component failure.

The equipment handling platform provides a working surface for equipment and personnel performing work in the under vessel area. It is a polar platform capable of 360 degree rotation. This equipment is designed in accordance with the applicable requirements of OSHA (Vol. 37, No. 202, Part 191 ON), AISC, and ANSI C1 (National Electric Code).

The spring reel is designed to be used to pull th e in-core guide tube seal or in-core detector into the in-core tube during in-core servicing.

The thermal sleeve installation tool locks, unlocks, and lowers the thermal sleeve from the control rod drive guide tube.

In-core flange seal test plug is used to determine the pressure integrity of the in-core flange seal. It is constructed of noncorrosive material. The key bender is designed to install and remove the antirotation key that is used on the thermal sleeve.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-46 9.1.4.2.10 Descripti on of Fuel Transfer The fuel handling system provides a safe and effective means for transporting and handling fuel from the time it reaches th e plant until it leaves the plant after postirradiation cooling.

The previous subsection has desc ribed the equipment and methods used in fuel handling. The following paragraphs describe the integrated fuel handling system that ensures the design bases of the fuel handling system is satisfied.

9.1.4.2.10.1 Arrival of Fuel on Site. The new fuel arrives at the site in either one of two types of special shipping containers.

The one type of cont ainer, designated the "RA Series," has space for two fuel bundles and the other type of container, designated "927 Series," has space for four fuel bundles.

The RA Series container consis ts of a metal inner container positioned by means of cushioning materials within an outer container. These containers with their contents meet all NRC requirements. Both inner and outer containers are reusable.

The 927 Series container is a steel unit with a removable cover. The fuel bundles are secured in the interior by a series of steel brackets. These containers also meet all NRC requirements.

After arrival of the fuel shipment the containers are visually inspected for evidence of damage during shipping. If there are indications of rough handling or damages, further investigation, for possible damage to fuel bundles, will be perf ormed. With the RA Series the inner metal container is then rem oved from the outer container with appropriate lifting equipment. The containers are then taken to the refuel floor using lifting rigging designe d for the task and the container is placed in a horizontal position. The covers are then removed to expose the fuel bundles. The fuel is then visually inspected fo r any obvious dama ge and a radiation survey for contamination is made.

Bundle restraints will be installed on the metal container in preparation for raising it to the vertical position. The container is then raised to the vertic al position and secured using the reactor building auxiliary hoist with rigging designed for the task.

New fuel may also be shippe d with channels already installed, in which case channel installation will not occur in the inspection stand or the fuel pool.

9.1.4.2.10.2 Refu eling Procedure. Fuel handling procedures are described below and shown visually in Figure 9.1-13 through Figure 9.1-16.

The refueling floor layout is shown in Figure 9.1-5 and component drawings of the principal fuel handling equipment are shown in Figures 9.1-7 through 9.1-12.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-47 During core alterations (excep t movement of control rods w ith their normal drive system), direct communication is maintained between the control room and refueling platform personnel. Direct communication is demonstrated within 1 hr prior to the start of and no less than once per 15 hr during core alterations (with the exceptions noted above). When direct communication cannot be maintained, core alte rations (with the exceptions noted above) are immediately suspended.

The fuel handling process takes place primarily on the refueling floor above the reactor. The principal locations and equipment are shown on Figure 9.1-13. The reactor and fuel pool are connected to each other by slot (A) which is open during reactor refueling. At other times the slot is closed by means of blocks and gates, which make watertight barriers.

Since provisions for portable shielding are not pr ovided in the drywell, administrative control is used during refueling opera tions to avoid overexposure of personnel as the result of a postulated fuel drop accident such as a drop occurring on the reactor seal plate.

New fuel in shipping containers is brought up to the refueling floor through the hatches.

The handling of new fuel on the re fueling floor is illustrated in Figure 9.1-14. The transfer of the bundles between the fuel cont ainer (C), and the new fuel in spection stand (D), the new fuel storage vault (E) and the fuel pool storage racks (F) is accomplished using the reactor building crane auxiliary hoist or jib crane.

The jib crane is also used to transfer new fuel from the new fu el vault or inspection stand to the fuel prep machine in the spent fuel pool. From this point on the fuel is handled by the telescoping grapple on the refueling platform.

The storage racks in both the vault and the fuel pool hold the fuel bundles or assemblies vertical, in an array which is subcritical unde r all possible conditions.

The new fuel inspection stand holds one or two bundles in vertical position. The inspector(s) ride up and down on a platform, an d the bundles are manually rotate d on their axes. Thus the inspectors can see all visibl e surfaces on the bundles.

The refueling platform uses a grapple on the te lescoping mast for lifting and transporting fuel bundles or assemblies. The tele scoping mast can extend to the pr oper work level, and, in its normal up position, maintains adequate water shielding over the fuel being handled.

The reactor refueling procedur e is shown schematically in Figure 9.1-15. The refueling platform moves over the fuel pool, lowers th e grapple on the telescoping mast (H), and engages the bail on a new fuel asse mbly which is in the fuel storage rack. The assembly is lifted clear of the rack and moved through slot (A) and ove r the appropriate empty fuel C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-48 location in the core (J). The ma st then lowers the assembly in to the location, and the grapple releases the bail.

The operator then moves the plat form until the grapple is over a spent fuel assembly which is to be discharged from the core. The assembly is grappled, lif ted, and moved through slot (A) to the fuel pool.

To preclude the possibility of ra ising radioactive material out of the water, redundant electrical limit switches are incorporated in the hoist and interlocked to prevent hoisting above the preset limit. In addition, the cables on the auxiliary hoists incorporate adjustable stops that will jam the hoist cable against the hoist structure, which prevents hoisting if th e limit switch interlock system should fail. Prior to moving irradiated fuel in the reactor pressure vessel, the reactor is verified to have been subcritic al for at least 24 hr. This is consistent with fuel handling accident assumptions described in Section 15.7.4.5.

When spent fuel is to be transferred, it is placed in a cask, as shown in Figure 9.1-16. The refueling platform grapples a fuel bundle from the storage rack in the fuel pools, li fts it, carries it to the transfer cask area of the pool, and lowers it into the canister within cask (M). When the canister is loaded, the building crane sets the cover on the canister, and then lifts the cask out of the pool. The cask is then decontaminated and lowered through the open hatchways (P) to the cask transport system at near grade level.

9.1.4.2.10.2.1 New Fuel Preparation.

9.1.4.2.10.2.1.1 Receipt and Inspection of New Fuel. The incoming new fuel shipping container will be visually inspect ed for shipping damage as the containers are offloaded from the transport vehicle. The fuel is also visually inspected upon opening the container on the refuel floor.

Incoming fuel is inspected as gove rned by plant procedures prior to being placed in the reactor vessel. Preferably, the inspection will be done before the fu el is stored.

However, depending on specific plans for the in itial fueling and/or subs equent refueling, fuel may be stored until a more desirable time for inspection.

Inspections may be performed on site or at the vendor facilitie

s. Onsite inspections are performed with the fuel bundles secured in th e fuel inspection stand. Inspections are performed, by qualified inspectors, in accordance with plant approved procedures.

If a fuel bundle fails inspection, discrepancies are noted and acti on will be taken to repair the fuel bundle or it will be set aside fo r disposition by the fuel manufacturer.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-49 After satisfactory inspection, the fuel bundle is then transferred to a st orage rack in the new fuel vault or to a storage rack in the spent fuel storage pool. The fuel bundle may or may not be channeled prior to storage.

9.1.4.2.10.2.1.2 Ch anneling New Fuel (N on Westinghouse Fuel). New fuel is channeled with new channels after the fuel is inspected in the inspection sta nd. The channels are visually inspected for defects prior to installing on the fuel. The inspected channel is then attached to lifting rigging and installed on the fuel while in the inspection stand.

The fuel can also be ch anneled in the spent fuel pool. For a channel located in the channel rack in the pool, the channel handli ng tool (fastened to the channel jib crane) is attached to the channel and raised manually with the assistance of the channel jib cr ane then positioned over the bundle to be channeled. The bundle is then raised into the ch annel and the channel fastener is installed with th e channel bolt wrench.

9.1.4.2.10.2.1.3 Cha nneling New Fuel (Westinghouse Fuel Only). Westinghouse Fuel (SVEA-96 type) requires the fuel bundle to be inserted into the combination fuel channel/lower tie plate. The new channels will be received up on the refue ling floor and inspected then installed into the new fuel stor age vault or the fuel prep mach ine. The Westinghouse fuel is inspected in the inspection stand then moved to the storage vau lt or fuel prep machine by use of the jib crane. The bundle is positioned over the channel and the temporary lower tie plate, installed for shipping purposes, if not removed earlier, is then removed and the four mini bundle assemblies are inserted into the channel. With th e bundle seated, the rigging is removed, the temporary upper bale handle is removed and the permanent bail handle is installed and torqued. The bundle, if channeled in the new fuel st orage vault, is then ready to be transferred to the spen t fuel pool prep machine.

9.1.4.2.10.2.1.4 Equipment Preparation. Prior to use for refueling, refueling equipment will be placed in readiness. All tools, grapples, slings, strongbacks, stud te nsioners, etc. will be given a thorough check and any defective (or well worn) parts will be replaced. Air hoses on the main grapples will be routin ely leak tested. Crane cables will be routinely inspected. All necessary maintenance and interlock checks will be performed to ensure no extended outage due to equipment failure.

The channeled new fuel will be ready in the storage pool.

9.1.4.2.10.2.2 Reactor Shutdown. The reactor is shut down according to a prescribed procedure. During cooldown the reactor pressure vessel is vented.

The reactor well shield plugs are removed using the reactor bu ilding crane and the supplied slings.

This operation can be immediatel y followed by removal of the canal plugs. These activities are covered in Sections 9.1.4.2.10.2.3.3 and 9.1.4.2.10.2.4.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-001 9.1-50 9.1.4.2.10.2.2.1 Drywell Head Removal. Immediately after removal of the reactor well shield plugs, the work to unbolt the drywell head can begin. The drywell head is attached by removable bolts. The bolts are unscrewed a nd together with their nuts are removed and stored.

The head strongback is attached to the unbolted drywell head and lifted by the overhead reactor building crane to its appointed storage space on the refueling floor.

9.1.4.2.10.2.2.2 Reactor Well Servicing. When the drywell head has been removed, an array of piping is exposed that must be serviced.

Various vent piping pe netrations through the reactor well must be removed and the penetrati ons made watertight. Vessel head piping and head insulation must be removed and tran sported to storage on the refueling floor.

Water level in the vessel is now adjusted in preparation for head removal.

9.1.4.2.10.2.3 Reactor Vessel Opening.

9.1.4.2.10.2.3.1 Vessel Head Removal. The stud tensioners are transported by the reactor building crane and positioned on the reactor vessel head. Each stud is tensioned and its nut loosened in a prescribed manner to uniformly detension the head. When the nuts are loose, they are backed off using a nut runner until only a few threads remain engaged. The vessel nut

handling tool is engaged in the upper part of the nut and the nut is rotated free from the stud.

The nuts and washers are placed in the racks provided for them a nd transported to the refueling floor for storage. With the nuts and washer s removed, the vessel stud protectors and vessel head guide caps are installed.

The head strongback, transported by the reactor building crane, is attached to the vessel head and the head transported to the head holding pedestals on the refueling floor. The head holding pedestals keep the vessel head elevated to facilitate inspection and O-ring replacement.

Alternatively, the reactor ve ssel head can be removed usin g the RPV head stud tensioner carousel/strongback, a tool that combines the functions of the stud te nsioner handling frame and the head lifting strongback. The carousel/str ongback, with the stud tensioners suspended from its integral monorail, is transported by th e reactor building crane a nd attached to the RPV head. Using the stud tensioners, the nuts are removed from the RPV head studs in the manner previously described, the nuts placed in a circular nut storage tran sporting rack, also integral to the carousel/strongback. When all the head stud nuts are re moved, the RPV head stud tensioner carousel/strongback, with the RPV head attached and all head stud nuts stowed in the integral nut rack, is transporte d by the reactor building crane to the head holding pedestals on the refueling floor.

The studs in line with the fuel transfer canal are removed from the vessel flange and placed in the rack provided. The loaded rack is transported to the refueling floor for storage.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-51 9.1.4.2.10.2.3.2 Dryer Removal. The dryer sling assembly is lowered by the reactor building crane and attached to the dryer lifting lugs. The dryer is lifted from the reactor vessel and transported to its storage location in the dryer-separator storage pool adjacent to the reactor well. The dryer is sprayed with water to control the spread of contamination during transfer and until water level in reactor well and dryer-separator pool cover the dryer.

9.1.4.2.10.2.3.3 Separator Removal. The four main steam lines are plugged from inside the vessel using the furnished plugs for this duty. Servicing of the sa fety and relief valves can thus be accomplished without adding to the critical refueling path time. The separator is unbolted using the shroud head bolt wrench es furnished. The fuel pool slot plugs are then removed.

The separator sling assembly is lowered into the vessel and attached to the separator lifting lugs. The water in the reactor well and in the dryer-separator storage is raised to fuel pool water level, and the separator is transferred to its allotted storage place in the equipment pool.

9.1.4.2.10.2.3.4 Fu el Bundle Sampling. During reactor operation, the core offgas radiation level is monitored. If a rise in offgas activity indicating a fuel failure has been noted, the reactor core will be sampled during a refueling outage to locate any leaking fuel assemblies.

9.1.4.2.10.2.4 Refueli ng and Reactor Servicing. The gates isolating the fuel pool from the reactor well are now removed thereby interconnecting the fuel pool, the reactor well, and the dryer-separator storage pool. The actual re fueling of the reactor can now begin.

9.1.4.2.10.2.4.1 Refueling. The refueling pattern will depend on various factors such as fuel performance, plant load ing, and fuel design.

Detailed procedures will be de veloped for various refueling task s such as fuel receipt, fuel inspection, fuel transfer from vault to pool, fuel movements within the pool, fuel movements within the reactor, or between the spent fuel pool and the reacto r, as well as procedures for handling of other core components.

During a normal equilibrium outage, approxima tely 33% of the fuel is removed from the reactor vessel, 33% of the fuel is shuffled in the core (generally from center to peripheral locations), and 33% new fuel is installed. The actual fuel handling is done with the fuel grapple which is an integral part of the refueling platform. The platform runs on rails over the fuel pool and the reactor well. In addition to the fuel grapple, the refueling platform is equipped with two auxiliary hoist s which can be used with vari ous grapples to service other reactor internals.

To move fuel, the fuel grapple is aligned over the fuel assembly , lowered and attached to the fuel bundle bail. The fuel bundle is raised out of the core, moved through the refueling slot to

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-52 the fuel pool, positioned over the storage rack and lowered to storage. Fuel is shuffled and new fuel is moved from the storage pool to the reactor vessel in the same manner.

9.1.4.2.10.2.5 Vessel Closure. The following steps, when perf ormed, will return the reactor to operating condition. The procedures are the reverse of those described in the proceeding sections. Many steps are performed in pa rallel and not in th e sequences listed.

a. Core verification. The core position of each fuel assembly must be verified to ensure the desired core conf iguration has been attained;

b. Control rod drive tests.

The control rod drive timi ng, friction and scram tests are performed;

c. Install inner fuel pool gate;

d. Replace separator and drain dryer-separator storage pool and reactor well;

e. Bolt separator and remove the four steam line plugs;

f. Replace steam dryer;

g. Install reactor vessel head;

h. Decontaminate reactor well;

i. Open drywell vents, install vent piping;

j. Replace slot plugs;

k. Replace vessel studs;

l. Install vessel head piping and insulation;

m. Replace dryer-separator canal plugs;

n. Hydrotest vessel, if necessary;

o. Install drywell head;

p. Install reactor well shield plugs; and

q. Startup tests. The reactor is returned to full power operation. Power is increased gradually in a series of steps until the reactor is operating at rated C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-53 power. At specific steps during the approach to power, the in-core flux monitors are calibrated.

9.1.4.2.10.3 Departure of Spent Fuel From the Reactor Building. Spent fuel is removed from the Reactor Building in an NR C licensed shielded cask. The reactor building and cask handling facilities (see Section 9.1.4.2.2 for a description of the cr ane) are designed to handle casks up to 125 tons.

The following description of the spent fuel departure is based on the licensed Holtec HI-STORM 100 System which includes a transfer cask of 125 tons maximum weight with a capacity of 68 BWR fuel bundles.

After the cask has been inspected and prepared for lifting, it is transferre d to the reactor operating floor and placed onto the cask decontamination pad where it is prepared for placement into the spent fuel cask loading area. The cask provides shielding for a canister contained within the cask. The canister, when sealed, is a confinement for the fuel.

The cask is next raised and transferred into the spent fuel cask loading ar ea. The cask lifting yoke is lowered until it is dise ngaged from the cask trunnions.

Spent fuel is transferred under water from st orage in the fuel pool to the cask using the telescoping fuel grapple mounted on the refueling pl atform. When the cask is filled with spent fuel, the canister closur e head is placed on the cask and th e lift yoke engaged with the cask trunnions. The loaded cask is raised and tr ansferred to the cask decontamination pad.

The cask is checked by health physics personnel and decontamination is performed with high pressure water sprays, chemical s, and hand scru bbing, as required, to clean the cask. The canister is sealed, dewatered and inerted. The ca sk lid is installed and the cask prepared for lifting.

The sealed canister is then transferred to a storage cask. The storage cask and canister are moved to the spent fuel storage area.

9.1.4.3 Fuel Handli ng Safety Evaluation Stresses in all structural and mechanical parts of the reacto r building crane system are far below the endurance limits for infi nite life of the various materials for both the rated crane capacity and the test load of 125% of rated load. All loads to be handled are below rated crane capacity. Therefore, stresses should never reach allowable working stresses. Loads on the structural parts vary but do not reverse. The only critical part s with stress re versals are the rotating parts, and these are pr ovided with single failure protect ion. Since the crane is to operate under normal temperature conditions and since the stress levels are below the C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-54 endurance limits for infinite li fe, testing of the crane to 125

% of rated capacity provides reasonable assurance that th e crane will not fail while handling a spent fuel cask.

As described in Section 9.1.4.2.2.2, sufficient redundancy is provided in the reactor building crane so that no credible postulated failure of any crane component will result in the dropping of the fuel cask. Therefore, the consequenc es of a cask drop accide nt are precluded. In addition, crane travel over the spent fuel pool is controlled by travel path (Figure 9.1-5) and interlocks that prohibit the crane from traveling over the spent fuel racks. At no time while being transported does the cask pass over any safe shutdown items.

The objectives of Regulatory Guide 1.13 are met.

Furthermore, when the crane is carrying the cask over the refueling floor area, the clear distance between the bottom of the cask and the refueling floor is minimized by good operating practice. Should any crane failure occur while the crane is moving the cask over the refueling floor area, the cr ane drop of less than one inch described in Section 9.1.4.2.2.2 prevents the cask from physically contacting the floor. See Section 9.1.4.2.2.2 for safety features which ensure cr ane stability during tornado and seismic excitation.

Jib cranes MT-CRA-9A and 9B are designed and equipped to meet the Class A1 (standby) service requirements of CMAA Specification #70. Each crane is capable of raising, lowering, holding, and transporting a test work load of 12 5% of rated work load without damage to or excessive deflection in a part and without inducing permanent deformation in any crane element. Cables are of the nonrotating type to prevent rota tion of load during raising or lowering operations.

The following factors of safety are used in design and ar e based on the maximum stress produced by worst load combinations and the average ultimate strength of the material, unless otherwise stated herein:

a. All load carrying components except st ructural members and hoisting ropes are designed to have a minimum safety factor of five;

b. All mechanical parts subject to dynamic strains, such as ge ars, shafts, drums, blocks, and other integral parts, have a minimum safety factor of five;

c. All hoisting ropes are designed to have a minimum safety fact or of five based on the published breaking strength of the rope;

d. Components of the jib cr anes are adequately proportioned to limit the overall deflections of the crane to safe limits under any position of the loaded trolley hoist. Maximum vertical static deflec tion of the boom with nameplate rated hook load is less than 0.6 in.; and

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-55 e. Structural members/shapes not covered by the above safety factors are designed in accordance with CMAA Specification No. 70, and if not covered by CMAA Specification No. 70, in accordance with the AISC Specification for the design, fabrication, and erection of structural steel for buildings except that the allowable stress shall not exceed 80%

of the AISC allowable design stress.

Jib cranes MT-CRA-9A and 9B are designed to Seismic Categor y I requirements (with lifted load).

The cranes are equipped with switches to limit their lifted load to 1000 lb.

The cranes are exempted from the NUREG 0612 requirement because the definition of "Heavy Loads" at CGS is 1200 lb. The cranes are also not required to hold their lifted load during a seismic event since the maximum load of 1000 lb (or 1200 lb) is enveloped by analysis that indicates the high density spent fuel storage racks can withstand a drop (both from a structural and criticality standpoint) of 1510 lb from the height of the fuel pool water surface. Licensee Controlled Specifications limit th e loads over spent fu el in the spent fu el pool to 1500 lb whenever irradiated fuel assemb lies are in the spent fuel storage pool.

Crane operations with loads will be suspended in the sp ent fuel storage pool area if ther e is less than 22 ft of water over the top of fuel assemblies seated in the storage racks. Based on the analysis presented in Section 15.7 , the radiological conseque nces of a fuel bundle (with or without a channel) dropped by the jib crane are bounded by the fuel handling analysis wh ich assumed a bundle dropped from 34 ft into the reactor core. This resulted in a much higher kinetic energy level.

In comparing the jib crane drop to the reactor core drop, it is not possible to cause the release of more fission gases from the damaged fuel than assumed in the Section 15.7 analysis due to the shorter drop distance and resulting lower kinetic energy.

The cranes are designed so as to be capable of operating w ithin the following tolerances:

a. With all brakes adjust ed for normal operation, it is possible to control the vertical movement to within 0.25 in. under all conditi ons of loading,

b. Cranes operate through full hook lift w ithout noticeable rotation of load, and

c. With hook carrying 100% of rated load and lowering at fu ll speed, the motor does not exceed 125% of synchronous speed.

Since jib cranes MT-CRA-9A and 9B are to be used on and around the fuel pool for handling new and spent fuel and other components in the work area, the following safety features are provided:

The hoist is motorized with a motorized boom and jib. To preclude the possibility of inadvertently raising radioactive materials out of the water, a specially prepared sling or

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-56 handling tool will be used when handling materials not intended to be raised out of the water.

The length of the sling or tool will be administratively controlled such th at with the crane at full mechanical up travel the top of the active fuel will be at a minimum of 7 ft 6 in. below water level.

The unit is equipped with two full capacity brakes. A mechanical force gauge is supplied to automatically stop the hoist on an overload signal of 1000

+50 lb. Two additional microswitches are wired in parallel and the thr ee leads are brought out with the power leads for connection to a platform receptacle. The two additional switches are designed to open at 400 +50 lb (these switches are not normally used for any function).

Safety aspects (evaluation) of the fuel servicing equipment are discussed in Section 9.1.4.2.3 and safety aspects of the refueling equipment are discussed throughout Section 9.1.4.2.7. A description of fuel transfer, including appropriate safety feat ures, is provided in Section 9.1.4.2.10. In addition, the following summary sa fety evaluation of the fuel handling system is provided below.

The fuel prep machine can remove and install ch annels with all parts re maining under water.

Mechanical stops in use when handling irradiated fuel preven t the carriage from lifting the fuel bundle or assembly to a height where water shielding for the active fuel is less than 7.5 ft.

Irradiated channels, as well as small parts, such as bolts and springs, are stored underwater. The spaces in the channel storage rack have center posts whic h prevent the loading of fuel bundles into this rack.

There are no nuclear safety problems associated with the hand ling of new fuel bundles, singly or in pairs. Equipment and procedures prevent an accumulation of more than two bundles in

any location.

The refueling platform is designed to prevent it from toppling into the pools during a SSE.

The refueling mast has normal up limit switches to prevent raising the top of active fuel to a height of less than 7.5 ft below the water su rface. The grapple is hoisted by redundant cables inside the mast and is lowered by gravity. A digital readout is displayed to the operator, showing him the exact coordinate s of the grapple over the core.

The mast is suspended and gimbal ed from the trolley, near its top, so that the mast can be swung about the axis of platform travel to remove the grapple fr om the water for servicing and storage.

The grapple has two mechanically coupled hooks, operated by a si ngle air cylinder.

Engagement is indicated to the operator.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-57 In addition to the main hoist on the trolley, there is an auxiliary hoist on the trolley, and another hoist on its own mono rail. These three hoists are precluded from operating simultaneously because control power is available to only one of them at a time.

The two auxiliary hoists have electrical interlocks which will prevent lifting the top of the active fuel on lifted fuel asse mblies higher than 7.5 ft under wa ter. Adjustable mechanical jam-stops on the cables back up these interlocks.

In summary, the fuel handling system complies with Regulat ory Guide 1.13, Revision 1, GDC 2, 3, 4, 5, 61, 62, and 63, and applicable por tions of 10 CFR Part 50.

The safety evaluation of the new and spent fuel storage is presented in Sections 9.1.1.3 and 9.1.2.3.

9.1.4.4 Testing and In spection Requirements

9.1.4.4.1 Testing and Inspection of Cranes

The main and auxiliary hooks of the reactor building crane are proof tested in the vertical, direction with a total uniformly applied load equal to 150% of their rated capacity. Tests on the main hook are made with a load suspended.

After the load tests, the hooks are checked by magnetic particle inspection and for any dimensional ch ange. When completely assembled, the crane components (except for wire ropes), are ope rated to ensure the accuracy of fabrication and the quality of workmanship.

After erection in the reactor build ing, the crane is statically te sted to 125% of rated capacity (156.25 tons for the main hoist and 18.75 tons for the auxiliary hoist). The tests are performed in accordance with written proced ures which include movement in all positions of hoisting, lowering and trolley and bridge travel. After completion of the static test, a full performance test with 100% of the design rated load is performed.

Operational tests and visual inspections are ma de at periodic interval s during the life of the crane to demonstrate it s ability to safely perform its f unctions. The crane hooks are to be inspected by the magnetic particle or liquid penetrant methods as applicable. These inspections and tests will be scheduled to precede major fuel handling activities.

Jib crane hooks (9A and 9B) shall be either ma gnetic particle inspected in accordance with ASTM E-109 (July, 1975) or liquid penetrant inspected in accordance with ASTM E-165 (July, 1975). Standard shop tests shall be performed on all equi pment. Field tests shall be performed upon each piece of equipment, after completion of er ection and installation, in order to verify the overall performance of the equi pment against requirements and to check the mechanical performance of the equipment with regard to app licable specifica tion requirements.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-58 Periodic operational tests and inspection will be performed on the jib cranes to demonstrate their ability to perform their safety function.

9.1.4.4.2 Inspection and Tes ting Requirements of Refueli ng and Servicing Equipment

9.1.4.4.2.1 Inspection. Refueling and servicing equipment is subject to the strict controls of quality assurance, incorporating the requirements of 10 CFR Part 50, Appendix B.

Components defined as im portant to safety, such as the fuel storag e racks and refueling platform, have an additional se t of engineering specified "qua lity requirements" that identify safety-related features which require specific quality assurance verifi cation of compliance to drawing requirements.

For components classified as ASME Section III, the shop operation must secure and maintain an ASME "N" stamp, which requires the submitta l of an acceptable AS ME quality plan and a corresponding procedural manual.

Additionally, the shop operation must submit to frequent ASME audits and component inspections by resident state code inspectors.

Prior to shipment, every com ponent inspection item is re viewed by quality assurance supervisory personnel and combined into a summary product quality checklist (PQL). By issuance of the PQL, verification is made that all quality requirements have been confirmed and are on record in the pr oduct's historical file.

9.1.4.4.2.2 Testing. Qualification testing is performed on refueling and servicing equipment prior to multiunit production. Test specifications are defi ned by the responsible design engineer and may include; sequence of operations, load capacity and life cycles tests. These test activities are performed by an independent test engineering group and, in many cases, a full design review of the product is conducted before and after the qualification testing cycle.

Any design changes affec ting function, that are made afte r the completion of qualification testing, are requalified by test or calculation.

Functional tests are performed in the shop prior to the shipment of production units and include electrical tests, leak tests a nd sequence of operations tests.

When the unit is received at th e site, it is inspected to insure no damage has occurred, during transit or storage. Prior to use and at periodic intervals, each piece of equipment is again tested to ensure the electrical and/or mechanical functions are operational.

Passive units, such as the fuel storage r acks are visually inspected prior to use.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.1-59 There is an operation and main tenance instruction manual for each tool that additionally requires a series of functional checks each time the unit is operated for reactor refueling or servicing.

Fuel handling and vessel servicing equipment preoperational tests are described in Section 14.2.12.1.12.

9.1.4.5 Instrumentation Requirements 9.1.4.5.1 Instrument Requirements - Cranes Operation of the reactor building crane is from the cab or from the floor by radio control. Control at any one time is from one point only. Sufficient electrical circu itry in the system is provided such that no single credible electrical component failure will cause the load to drop. The reactor building crane and its safety features are described in Section 9.1.4.2.2.

The refueling platform contains a position-indicating system that indicates the position of the fuel grapple over the core. Interlocks on both the fuel grapple hoist and the auxiliary hoists are provided for safety purposes.

The refueling interlocks ar e described and evaluated in Section 7.7.1.13.

For a description of jib crane safe ty-related interlocks, see Section 9.1.4.3. Except for jib cranes MT-CRA-9A and MT-CRA-9B, limit switches are provided for all electrically operated hoists to prevent overhoisting. Pushbutton pendant controls are provided with stepped controls for multispeed motions.

9.1.4.5.2 Instrumentation Requirements - Refueling and Servicing Equipment

The majority of the refueling and servicing equipment is manually ope rated and controlled by the operator's visual observations.

This type of operation does not necessitate the need for a dynamic instrumentation system.

However, there are several com ponents that are essential to prudent operation that do have instrumentation and control systems.

The refueling platform has a non-safety-related X-Y-Z position i ndicator system that informs the operator which core fuel cell the fuel grappl e is accessing. Interlocks and control room monitor are provided to prevent the fuel grapple from operating in a fuel cell where the control rod is not in the proper orientation for refueli ng. Refer to Section 7.7.1.13 for discussion of refueling interlocks.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-027 9.1-60 Additionally, a programmable logic controller (PLC), which is a part of the refueling platform control system, provides indications to the operator for grapple limits, hoist load conditions, and confirmation that the grapple ho ok is either enga ged or released.

The load for the main hoist, auxiliary frame hoist, and mono rail hoist is determined by electronic load cells and strain gauge transmitters.

The load for the main hoist and auxiliary frame hoist is input to the PLC. The PLC provides hoist loaded and ho ist overload signals for both the main hoist and auxiliary frame hoist, and provides slack cabl e signal for the main hoist. Setpoint modules monito r the monorail hoist load and provide hoist loaded and hoist overload interlock signals.

Automatic shutdown occurs whenever threshold limits are exceeded on either the main hoist, the auxiliary frame hoist, or the monorail hoist.

9.1.4.5.3 Fuel Support Grapple

Although the fuel support piece gr apple is not important to safe ty, it has an instrumentation system consisting of mechanical switches and indicator lights.

This system provides the operator with a positive indication that the grapple is properly aligned and oriented and that the grappling mechanism is either extended or retracted.

The Westinghouse MLT, also used to grapple a nd remove fuel support pieces, is designated Safety Class 3. The instrumentation system for the MLT consists of flag-type indicators directly linked to each grapple mechanism to sh ow when the grapple is engaged to the fuel support piece. The MLT also incorporates internal guide tubes for underwater cameras, allowing the MLT operator to view and verify proper seating and alignmen t of the fuel support piece on its guide pin.

9.1.4.5.4 Other

See Table 9.1-7 for additional refueling and servicing equipment.

9.1.4.5.5 Radia tion Monitoring

The radiation monitoring equipment for the refueling and servicing area is discussed in Sections 11.5.2.1.2 and 12.3.4. 9.1.5 CONTROL OF HEAVY LOADS 9.1.5.1 Introduction/Licensing Background Columbia Generating Station controls heavy load lifts by implementing NUREG-0612 recommended guidelines in its plant procedures.

Columbia's heavy load s program procedures focus on areas where a load drop could impact stored spent fuel, fuel in the reactor core, or equipment that may be require d to achieve safe shutdown or permit continued decay heat

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-027 9.1-61 removal. Columbia reduces the likelihood of dropping heavy loads by implementing criteria for establishing safe load path s, procedures for load handlin g operations, training of crane operators, design, testing, in spection, and maintenance of cranes and lifting devices, and analyses of the impact of heavy load drops.

Columbia also mitigates the consequences of a heavy load drop with the use of a single-failure-proof crane in certain applications for increased handling system reliabili ty. In other areas where a single-failure-proof crane is not used, load drop and consequence analyses are performed for asse ssing the impact of dropped loads on plant safety and operations.

9.1.5.2 Safety Basis The recommended guidelines esta blished in NUREG-0612 are the safety basis that provides a defense-in-depth approach to controlling the handling of heavy lo ads near spent fuel and safe shutdown equipment. These guidelines are reflected in Columbia's heavy loads handling procedures. These procedures incorporate NUREG-0612 requirements except for those deviations approved by the NRC.

Columbia's heavy loads handli ng procedures ensure that the risk associated with load hand ling failures is acceptably low, based on meeting requirements of Phase 1 of NUREG-0612, Sec tion 5.1.1, and based on incr easing the handling system reliability by meeting NUREG-0612 Phase II guidelines with the utilization of a single-failure-proof crane where a load drop could impact on stor ed spent fuel and fuel in the reactor core. An exception to NUREG-0612 Phase I, Section 5.1.1(1) was approved to not require load paths marked on the floor for monorail hoists because the single available path cannot vary.

Additionally, load paths for the Reactor Building Crane loads are not required to be permanently marked on the refueling floor because the multiple paths would overlap in places and would be difficult to follow. Furthermore, protectiv e coverings on the refueling floor would cover the paths for many lifts. The main Reactor Building crane (MT-CRA-2) has been approved as single-failure-proof crane that meets the criteria es tablished in either NUREG-0554 or NUREG-0612 Appe ndix C. Additional defens e-in-depth analyses and evaluations are performed for hea vy load lifts not associated with the use of a single-failure-proof crane in areas where a load drop could impact equipment that may be required to achieve safe shutdown or permit continued decay heat removal.

9.1.5.3 Scope of Overhead Heavy Load Handling Systems At CGS, overhead handling systems are used to handle heavy loads in the area of the reactor vessel or spent fuel in the spent fuel pool. Additionally, loads ar e handled in other areas where their accidental drop may damage safe shutdown systems or systems that permit continued decay heat removal. The following is the committed heavy lifts along w ith their associated cranes/hoists.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-027 9.1-62 Committed Heavy Lifts Committed Heavy Load Lift Committed Heavy Load Lift 1 Vessel Cavity Shield Plug (MT-CRA-2) 21 MPC Lid (MT-CRA-2) 2 Dry/Separator Pool Plugs Top (MT-CRA-2) 22HI-STORM Lid (MT-CRA-2) 3 Dry/Separator Pool Plugs Bottom (MT-CRA-2) 23 MPC with contents (MT-CRA-2) 4 Spent Fuel Gates (MT-CRA-2) 24 MPC without contents (MT-CRA-2) 5 Fuel Pool Shield (Slot) Plugs (MT-CRA-2) 25Sea Van Container & other loads during plant operation (MT-CRA-1) 6 "Cattle" Chute (MT-CRA-2) 26Standby Service Water Pumps (MT-HOI-6A & 6B) 7 Space Frame (w/head insulation) (MT-CRA-2) 27RHR A&B Shield Blocks (MT-HOI-6) 8 Drywell Head (MT-CRA-2) 28RHR A&B Pumps (MT-HOI-6) 9 Reactor Head (MT-CRA-2) 29RHR C Shield Blocks (MT-HOI-8) 10 Steam Dryer (MT-CRA-2) 30 RHR C Pumps (MT-HOI-8) 11 Moisture Separator (MT-CRA-2) 31 RCIC Shield Blocks (MT-HOI-7) 12 Reactor Vessel Service Platform (MT-CRA-2) 32 RCIC Pump & Turbine (MT-HOI-7) 13 RHR HX Floor Plugs (MT-CRA-2) 33 LPCS Shield Blocks (MT-HOI-9) 14 New Fuel Pit Floor Plugs (MT-CRA-2) 34 LPCS Pump/Motor (MT-HOI-9) 15 FPC Skimmer Surge Tank Plugs (MT-CRA-2) 35 HPCS Shield Blocks (MT-HOI-10) 16 Refueling Basket (MT-CRA-2) 36 HPCS Pump (MT-HOI-10) 17 RPV Head Stud Tension/Carousel (MT-CRA-2) 37RRC Motors (MT-HOI-16) 18 Mating Device (MT-CRA-2) 38 Steam (Pipe) Tunnel Shield Blocks (MT-HOI-18) 19 HI-TRAC Cask with or without MPC, lid, etc. (MT-CRA-2) 20 HI-TRAC Top Lid (MT-CRA-2)

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-027 9.1-63 9.1.5.4 Control of Heavy Loads Program Columbia's Control of Heavy Loads Program consists of the following:

1. Commitments to NUREG-0612, Phase I General Requirements.

2. Commitments to NUREG-0612, Phase II with the utilization of a single-failure-proof crane for Reactor Pressure Vessel Head (RPVH) lifts and fo r spent fuel cask lifts over the spent fuel pool. Additional defense-in-depth analyses and evaluations are performed for heavy load lifts not associated with the use of a single-failure-proof crane in areas where a load drop could impact equipment that may be required to achieve safe shutdown or permit continue d decay heat removal.

9.1.5.4.1 Columbia Genera ting Station Commitments in Response to NUREG-0612 Section 5.1.1, Phase I Guidelines 9.1.5.4.1.1 Safe Load Path. Safe load paths are defined in Columbia's heavy load lifting procedures and some temporary paths may be marked when it can help assure safe load handling. A person in ch arge, other than the operator, is responsible for inspecting the load path for potential interferences. Load paths are not marked on the floor for monorail hoists because the single available path can not vary. Load paths for the Reactor Building Crane loads are not permanently marked on the refueling floor because the multiple paths would overlap in places and would be difficult to follo

w. Protective covering s on the refueling floor would cover the paths for many lifts. The a bove exceptions are approved NRC deviations.

9.1.5.4.1.2 Procedures. Procedures are implemented to prevent personnel e rrors consistent with the requirements of Sec tion 5.1.1(2) of NUREG-0612.

9.1.5.4.1.3 Crane and Hoist Operators. Plant-wide training and qualification procedures have been implemented and meet the requirement found in Section 5.1.1(3) of NUREG-0612.

9.1.5.4.1.4 Speci al Lifting Devices. Special lifting devices satisfy the guidelines of ANSI N14.6-1978 are identified, maintained and controlled by plant procedures. These requirements are consistent with Section 5.1.1(4) of NUREG-0612. Dynamic loads determined as a percentage of static loads are added to the static load to obtain the design load.

9.1.5.4.1.5 Lifting Devices that are not Specially Designed. Lifting devices that are not specially designed are used in accordance with ANSI B30.9-1971.

Usage is restricted to properly trained and qualified si te rigging personnel c onsistent with the requirements found in Section 5.1.1(5) of NU REG-0612. Dynamic loads determined as a percentage of static loads are added to the static load to obtain the design load.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-027 9.1-64 9.1.5.4.1.6 Crane Inspecti on, Testing and Maintenance. Cranes are insp ected, tested, and maintained in accordance with the guidance in Chapter 2-2 of ANSI B30.2-1976 by inspection, maintenance and testing procedures. This is consistent with the requirements found in Section 5.1.1(6) of NUREG-0612.

9.1.5.4.1.7 Hoist and Crane Design 9.1.5.4.1.7.1 Hoist Design. In the absence of specific guidance from NUREG-0612, hoists are designed to applicable re quirements of industry standards CMAA 74, ANSI B30.11 and ANSI B30.16.

9.1.5.4.1.7.2 Crane Design. Cranes are designed according to applicable criteria and guidelines of Chapter 2-1 of AN SI B30.2-1976. These crane re quirements are consistent with Section 5.1.1(7) of NUREG-0612.

9.1.5.4.2 Reactor Pressure Vessel Head Lifting Procedures Columbia Generating Station RPV handling procedures are us ed to control the lift and replacement of the reactor pressure vessel head. These procedures rely on the use of the main Reactor Building Crane (MT-CRA-2) which is a single-failure-proof crane.

9.1.5.4.3 Single-Failure-Proof Cranes for Spent Fuel Casks The Spent Fuel Casks lifts and movements are controlled by CGS procedures. These procedures rely on the use of the main Reactor Building Crane (MT-CRA-2) which is a single-failure-proof crane.

9.1.5.4.4 Other Analyses Additional defense-in-depth anal yses and evaluations are perfor med for heavy load lifts not associated with the use of a single-failure-proof crane in areas where a lo ad drop could impact equipment that may be require d to achieve safe shutdown or permit continued decay heat removal. 9.1.5.5 Safety Evaluation The heavy loads handling activities at CGS are controlled in a manner consistent with the requirements of Phase I of NUR EG-0612, except as allowed by the NRC, and with industry standards to ensure defense-in-depth is maintained during h eavy load movements. A single-failure-proof crane is used for Reactor Pressure Vessel Head and Spent Fuel Cask movements.

Evaluations are performed for hea vy load lifts not associated with the use of a single-failure-proof crane and appropriate measures are implemented as necessary to ensure adequate safety

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-029 9.1-65 is maintained. Risks associated with heavy lo ad lifting are identified, assessed and managed through usage of CGS lifting procedures.

9.

1.6 REFERENCES

9.1-1 EMF-2837(P), Revision 0, "Columbia Generating Stati on New Fuel Storage Vault Criticality Safety Analysis for ATRIUM-10 Fuel," Framatome ANP, December 2002.

9.1-2 WAPD-TM-678 PDQ-7 Refe rence Manual by W. R. Ca ldwell, Bettis Atomic Power Laboratory, January 1967.

9.1-3 NUS-894 NUMICE-2, A Spectrum Depe ndent Non-spatial Ce ll Depletion Code by Y. S. Kim, NUS Corpor ation, March 1972. Nu mice is NUS' version of LEOPARD.

9.1-4 ORNL-4938 KENO-IV An Improved Monte Carlo Criticality Program by L. M. Petrie and N. F. Cross, ORNL, November 1975.

9.1-5 ORNL-TM-3706 AMPX-L Modular Code System for Generating Coupled Multi-Group Neutron Gamma Libraries from ENDF/B by N. M. Greene et al., ORNL, November 1974.

9.1-6 MRI/STARDYNE3 (Version 3), developed by Mechanics Research, Inc., Los Angeles, California.

9.1-7 ANF-91-069(P), "Critical ity Safety Analysis Wash ington Public Power Supply System WNP-2 Spent Fuel Storage Rack with 9 x 9-9X Fuel," March 1991.

9.1-8 CE NPSD-786-P, Revision 1, "WNP-2 SV EA-96 Spent Fuel Storage Criticality Safety Evaluation," September 1998.

9.1-9 EMF-2874(P), "Columbia Generating Station Spent Fuel Storage Pool Criticality

Safety Analysis for ATRIUM-10 Fuel

," Framatome ANP, February 2003.

9.1-10 MCNP4A - A General Monte Carlo N-Pa rticle Radiation Transport Code, Version 4A, LA-12625, March 1994.

9.1-11 GEH-0000-0075-4920, GE14 Fuel Design Cycle-I ndependent Analyses for Energy Northwest Columbia Generating Sta tion (most recent ve rsion referenced in the COLR).

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 9.1-66 9.1-12 NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, Resolution of Technical Activity A-36, July 1980.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 9.1-67 Table 9.1-1

9 x 9-9X, SVEA-96, and ATRI UM-10 Assembly Parameters Parameters 9 x 9-9X SVEA-96 ATRIUM-10 GE14 Enrichment (wt % 235 U) Bundle average up to 4.0 Bundle average up to 3.77 Lattice average up to 4.6 Lattice average up to

4.9 Pellet

diameter (in.)

0.3665 0.3224 0.3413 0.345 Pellet density

(% TD) 95 95.8 95.85 97.0 Average fuel length (in.) Infinite (tie plates not

modeled) 95.8 Infinite Infinite Clad I.D./

O.D. (in.)

0.374/0.433 0.329/0.379 0.

3480/0.3957 0.352/0.404 Gd 2 O 3 content None None 10 rods at 2 wt % Gd 2 O 3 Lattice average Gd 2 O 3 content up to a resultant in-core peak reactivity of 1.33 or a beginning-of-life reactivity of 1.31 Fuel rods per assembly 72 96 91 92 Internal water channel 1 1 1 2 internal water rods Rod pitch (in.) 0.569 0.488 0.510 0.510 C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Table 9.1-2 Calculated Dose Rate During Refueling 9.1-68 Refueling 5 Ft Above Fuel Pool Water Level (mrem/hr) 1 3.5 2 4.3 3 4.8 4 5.3 5 5.7 6 6.0 7 6.2 8 6.4 9 6.6 10 6.8 11 6.9 C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Table 9.1-3 Fuel Pool Cooling and Cleanup System Equipment Data 9.1-69 Fuel pool heat exchangers Number 2 Type Tube and shell Material tube/shell SS/CS Capacity, Btu/hr/heat exchanger 4.0 x 10 6 Cooling water flow, gpm/heat exchanger 575 Code and standards ASME/III, Class 3, and TEMA, Class R Seismic Category I Fuel pool circulation pumps Number 2 Type Horizontal, centrifugal

Material SS Flow, gpm 575

Head, ft of H 2 O 160 Motor size, hp 50 Seismic Category I Code ASME/III, Class 3 Fuel pool filter demineralizer Number 2 Design flow rate, gpm 1000 Design pressure, psig 150 Design temperature, °F 150 Material CS-plastic lined Code ASME/III, Class 3 Seismic Category II C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Table 9.1-3 Fuel Pool Cooling and Cleanup System (Continued)

Equipment Data 9.1-70 Piping and valves Design pressure, psig 150 HX tu be side/300 HX shell side Material CS/SS Code ASME/III, Class 3 Seismic Category Fuel pool cooling portion I Cleanup portion II

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Table 9.1-4 Spent Fuel Pool Projected Heat Loads 9.1-71 Table 9.1-4 is deleted. The cycle-specific inform ation provided in this table is obsolete. The heat load in the spent fuel pool during normal refueling activities will vary based on outage-specific activities a nd the transfer of spent fuel to the Independent Spent Fuel Storage Installation.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Table 9.1-5 Fuel Pool Cooling and Cleanup System Performance Data 9.1-72 Design heat load, Btu/hr 8.0 x 10 6 Maximum heat load, Btu/hr 44.3 x 10 6 (see note) System design pressure, psig 150 System design temperature, °F 150 normal, 175 maximum Fuel pool water volume, gal 350,000 Dryer-separator pool water volume, gal 293,600 Reactor well water volume, gal 210,000

NOTE: The maximum heat load is the decay heat of one full core load of the fuel plus the remaining decay heat of previously discharged fuel assembli es. The heat load and the fuel pool cooling system will be controlled such that the temperature limits shown in Table 9.1-6 are maintained.

Table 9.1-6 Bounding Fuel Pool Cooling Events C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORTDecember 2009 9.1-73 FPC Loops RHR Loops in FPC Assist Mode Fuel Pool Heat Load/Scenario Fuel Cycle Length (Months)

Available

Credited Available

Credited Acceptance Criteria a (°F) Normal refueling b (non-transient condition) 24 2 2 1 0 150 Normal refueling c (anticipated operational transient condition) 24 1 1 1 0 155 Full core offload refueling d 24 2 0 1 1 145/175 Normal Operations e (non-transient/non-accident condition) 24 2 2 0 0 125 Normal Operations f (anticipated operational transient condition) 24 1 1 0 0 155 Normal Operations g (design basis LOCA condition) 24 2 1 0 0 125 (pre-accident) 175 (post-accident)

Table 9.1-6 Bounding Fuel Pool Coo ling Events (Continued)

Performance Data C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORTDecember 2009 9.1-74 Table Notes:

a Outage-specific/Cycle-specific calculations are performed, as needed, to ensure acceptance criteria limits are maintained.

b A normal refueling does not involve a full core offload. The normal refueling temperature limits go into effect upon placing t he plant in cold shutdown and remain in effect until plant start-up. The 150°F acceptance criterion is applicable for the case of both FPC loops available.

c A normal refueling does not involve a full core offload. The 155°F acceptance criter ion is applicable for an anticipated operational transient of the loss of one FPC loop.

d A full core offload is a non-routine evolution. Although no single failure is postulated to occur dur ing this evolution, only credit for RHR/FPC assist is taken. During this evolution, sufficient cooling is provided to maintain th e fuel pool temperature at le ss than 145°F. The licen sing basis limit is 175°F.

e Normal Operations is defined as any plant condition other than a refueling outage. The norma l operation temperature limits become applicable upon plant start-up follo wing a refueling outage. The 125°F value supports the initial (i.e., pre-accident) fuel pool temperature assumption in th e design basis LOCA analysis.

f Normal Operations is de fined as any plant condition other than a refueling outage. The 155°F value is applicable for an anticipated operational transient involving the loss of one FPC loop. If the fuel pool heat load is such that the resulting temperature transient exceeds 125°F , procedural guidance is in place to restore the fuel pool temperature to < 125°F.

g Normal Operations is defined as any plant condition other than a refueling outage. The 125°F and 175°F values support the initial fuel pool temperature a ssumption and the calculated peak fuel pool temperatures, respectively, in the design basis LOCA analysis. For this event, RCC is assumed to be lost and SW is manually aligned to the FPC system heat exchangers. An equilibrium temperature of 90°F is assumed for the SW.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Table 9.1-7 Tools and Servicing Equipment 9.1-75 Fuel Servicing Equipment In-Vessel Servicing Equipment Fuel preparation machines New fuel inspection stand Channel bolt wrenches

Channel handling tool Fuel pool sipper Channel gauging fixture General purpose grapples Fuel bundle inspection fixture

Servicing Aids Pool tool accessories Actuating poles General area underwater lights Local area underwater lights Drop lights Underwater TV monitoring system

Underwater vacuum cleaner

Viewing aids Light support brackets In-core detector cutter

In-core manipulator Reactor Vessel Servicing Equipment Reactor vessel servicing tools Steam line plugs Shroud head bolt wrenches

Head holding pedestals Instrument strongback Control rod grapple Control rod guide tube grapple Fuel support grapple Grid guide Control rod latch tool

Instrument handling tool Control rod guide tube seal In-core guide tube seals Blade guides

Fuel bundle sampler Peripheral orifice grapple Orifice holder

Peripheral fuel support plug Fuel bail cleaner Dummy fuel assembly

Multi Lift Tool (MLT)

Refueling Equipment Refueling equipment servicing tools Refueling platform Storage Equipment Spent fuel storage racks Channel storage racks In-vessel racks New fuel storage rack C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Table 9.1-7 Tools and Servicing Equipment (Continued)

LDCN-09-001 9.1-76 Reactor Vessel Servicing Equipment (Continued)

Head stud rack

Dryer-separator sling

Head strongback Steam line plug/installation tool Auxiliary work platform

Head stud tensioner carousel/strongback Under Reactor Vessel Servicing Equipment Control rod drive servicing tools

Control rod drive hydraulic system tools

Spring reels Control rod drive handling equipment Equipment handling platform Thermal sleeve installation tool

In-core flange seal test plug Keybender

Columbia Generating StationFinal Safety Analysis Report Fuel Location Blocked Location Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.New Fuel Storage Vault Fuel Loading Pattern 900547.33 9.1-1 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.960690.63 New Fuel Storage 9.1-2 Columbia Generating StationFinal Safety Analysis Report Figure Amendment 59 December 2007 Form No. 960690 LDCN-06-000 Draw. No.Rev.960690.64 9.1-3 Columbia Generating Station Final Safety Analysis Report Typical Spent Fuel Rack 12 x 16 Array 104.75" 78.75" 6.50" on Center Module Spacing 169.00" 6.50" 1/8" Stainless Steel Can0.21" Thk. B C 4

Figure Form No. 960690Draw. No.Rev.Amendment 53November 1998 960690.65Spent Fuel Pool Arrangement 9.1-4Turnbuckle (Seismic Restraint)

(4) 12 x 16 Cell Racks Shipping Cask Storage Area 48.50" 408" Inside Fuel Pool(2) 11 x 16 Cell Racks (2) 12 x 16 Cell Racks (1) 12 x 13 Rack (1) 8 x 13 Rack (1) 7 x 18 Rack 16.75" 36.00" 32.50" 23.50" 480" Inside Fuel Pool (4) 12 x 16 Cell Racks 17.25" Columbia Generating StationFinal Safety Analysis Report Figure Not Available For Public Viewing Amendment 61December 2011 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.1-6.1 98 M526-1 Fuel Pool Cooling and Clean-Up SystemRev.FigureDraw. No.

Amendment 59December 2007 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.1-6.2 1 M526-2 Fuel Pool Cooling and Clean-Up SystemRev.FigureDraw. No.

Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.960690.66Fuel Preparation Machine Shown Installed in Facsimile Fuel Pool 9.1-7 Carriage Upper Frame Lower Frame Columbia Generating StationFinal Safety Analysis Report Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.960690.68 New Fuel Inspection Stand 9.1-8 Upper Clamps Fuel Support

Structure Hydraulic Inspection

Stand Lower Swivel

Clamps Columbia Generating StationFinal Safety Analysis Report Detent Leaf Springs Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.960690.69Channel Bolt Wrench 9.1-9 Shaft Housing Shaft Drive Socket Knob Guide CableSocket Wrench Columbia Generating StationFinal Safety Analysis Report Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.960690.70Channel Handling Tool 9.1-10 Bail Handle Jaw Channel Stop Actuating Shaft Actuating Knob Guides(Typical of 2)

Columbia Generating StationFinal Safety Analysis Report Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.960690.71Fuel Inspection Fixture9.1-11 1.00 5.76 0.06 SqA0.12 T OT 10.00 Sq 1.31Typ 0.56 0.02Typ Lower Mtg Surface0.50 Thk Reqd Upper Mtg Surface 0.02 x 0.50 WD Spring Fingers 1.20 Fuel ASM 0.78 0.95 0.98 0.50 0.02 Dia Hole In Lifting Eye 0.16 0.02 x 45 2 Seat 3.80 0.01 (Free Width) to Fit in+ 0.032- 0.000 Dia Hole 3.75 162.00 Max.118.00 Min.

12.26 3.29 5.562 Dia.3.68 0.02 DiaTyp Approx Wt 8.7 lb 1 Fuel ASM A Columbia Generating StationFinal Safety Analysis Report 7.12 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.960690.72 General Purpose Grapple 9.1-12 Stud Receptacle (7/16-14 UNC)

StudView PointTorsion Bar

HousingTorsion Bar Set Screw Grapple Hooks Columbia Generating StationFinal Safety Analysis Report Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.960690.73Plant View of Refueling Facilities 9.1-13 Core Hatchway ShippingCask Area Refueling PlatformTrack Refueling PlatformTrack New Fuel StorageVault New Fuel Inspection Stand X Y Z A X Y Channel Storage Racks Fuel Preparation Machine (2)

Channel Handling Boom Z Fuel Storage Racks Fuel Pool Columbia Generating StationFinal Safety Analysis Report Appx Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.960690.74 Simplified Section of New Fuel Handling Facilities (Section X-X) 9.1-14 DJib Crane or 15 Ton Auxiliary Hoist of Reactor Building Crane New Fuel Storage Racks New FuelStorage Vault Fuel Pool Fuel Storage Racks New Fuel Inspection Stand C 5 Shipping Crate F E Fuel Columbia Generating StationFinal Safety Analysis Report Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.960690.67 Simplified Section of Refueling Facilities(Section Y-Y) 9.1-15 Refueling PlatformTelescoping Mast ReactorWell Fuel PoolTelescoping Mast Fuel Prep Machine Fuel Storage Racks K H J Core F H A Channel Handling Boom Columbia Generating StationFinal Safety Analysis Report Figure Amendment 57December 2003Draw. No.Rev.960690.75 Simplified Section of Fuel Shipping Facilities (Section Z-Z) 9.1-16 Building Crane Refueling Platform Open Hatchway P M P Spent Fuel Cask Form No. 960690 LDCN-00-042 Columbia Generating StationFinal Safety Analysis Report

9.1-18 Amendment 58 December 2005 Cavity In-Vessel Service Platform (CISP) Sectional View 910402.44 Columbia Generating Station Final Safety Analysis ReportDraw. No.Rev.Figure Form No. 960690FH LDCN-05-032 C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-04-020,04-052, 05-030 9.2-1 9.2 WATER SYSTEMS

9.2.1 PLANT

SERVICE WATER SYSTEM

9.2.1.1 Design Bases

a. The plant service water (TSW) system is designed to provide cooling water for removal of heat rejected from auxiliary (nonessential) equipment, including turbine generator and reactor auxiliaries located throughout the plant, and

b. The plant service water system is de signed to remove the maximum expected heat load of the equipment under normal operating conditions.

9.2.1.2 System Description

The plant service water system consists of two 100% capacity pumps (TSW-P-1A, TSW-P-1B) taking suction from the circula ting water intake structure an d supplying cooling water to equipment located throughout the plant (see Figure 9.2-1

). Plant service water is returned to the circ ulating water tunnel fo r heat removal by the circulating water system cooling towers.

The plant service water system is designed to function continuously during all modes of operation except during loss-of-coolant accident (LOCA) with loss of offsite power. Following loss of power under nonLOCA conditions, the plan t service water pumps will be supplied from standby power buses (see Section 8.3.1.1.1). The plant service water radiation monitor is continuously recorded in the control room and will alarm if the water becomes contaminated except during loss of offsite power when plant service water will be manually monitored for radiation. The circulating wa ter system blowdown line radiati on monitor serves as backup (see Figure 10.4-4

). On detecting radiation in the blow down line, it automatic ally isolates the blowdown to the river except during pl ant outages as described in Section 11.5.2.2.2.4.

In addition to the biocide treated circulating water supply utilized by the plant service water system, the plant service water system is equipped with a biocide treatment system to retard biological growth. Additional chemical treatment capability is provided to minimize silt deposition, scale formation, corro sion, and consequent fouling of heat transfer surfaces.

Required makeup to the plant service water system is included as part of the overall circulating water system makeup requi rements. The makeup flow can be directed in to the circulating water bay or to one or both of the plant service water pump su ctions. This is accomplished by means of a weir box and sluice gate arrangement in the circulating water mixing bay (see Section 10.4.5).

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-05-030 9.2-2 Water for plant service water pump bearing lubrication is supplied from the plant service water pump during operation of the pump and is filtered before bei ng injected to lubricate the bearings. The Fire Protection system provides TSW pump bearing lubrication water for initial system startup after both pum ps have been secured (o r tripped) (see Appendix F.2 , Section F.2.4.1).

The plant service water system piping is de signed in accordance with ANSI B31.1. The two service water pumps are designed in accordance with the Standard s of the Hydraulic Institute, Centrifugal Pump Section.

9.2.1.3 Safety Evaluation

The operation of the plant servic e water system is not required to ensure any of the following conditions:

a. The integrity of the reactor coolant pressure boundary (RCPB),

b. The capability to shut down the react or and maintain it in a safe shutdown condition, or

c. The ability to pr event or mitigate the consequen ces of accidents which could result in potential offsite exposures in excess of the guideline exposures of 10 CFR Part 100.

A single-failure analysis has not been provided for the plant service water system since this system serves only nonessential sy stems and is not required to pe rform a safety function. See Section 3.6 for a discussion of piping failures. The system, howev er, incorporates features that ensure continuity and reliability of operation. Plan t service water pumps will be alternately operated to minimize wear, and th e standby pump is availa ble as a replacement during maintenance of th e normally operating pump.

All piping, valves, and associated components of the plant service water system are classified Seismic Category II except in the reactor building and control building where system hangers are designed for Seismic Cate gory I loads (seismic 1M).

9.2.1.4 Testing and In spection Requirements The components used in this system were tested and inspected at the manufacturer's plant for conformance with specifications. After installation, major components were checked and the system hydrostatically tested to ensure leak tightness prior to plant startup. Preoperational testing included testing of auto matic controls for actuation at the proper setpoi nts, calibration of instruments and alarms. Routine visual inspection of system components, instrumentation, and trouble alarms verify system operability and performance.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-05-030 9.2-3 9.2.1.5 Instrumentation Requirements

Suitable alarms and indicators are provided to monitor and control the system throughout the system to allow the control room operator to take corrective action as required.

Each pump is manually started from the main control room. The standby pump is automatically started by actuation of a low pr essure switch in the operating pump discharge line. The pressure at the discharge of each pump is indicated locally and also in the main control room. A pressure switch provides ann unciation in the main control room on low pump discharge pressure signal.

9.2.2 REACTOR

BUILDING CLOSED COOLING WATER SYSTEM

9.2.2.1 Design Bases

a. The RCC is designed to provide ade quate cooling water to auxiliary plant equipment in the reactor and radwaste buildings during all normal modes of operation, and

b. The RCC is designed to provide a closed cooling water loop between nonessential systems which are potentially contaminated and the plant service water which is used for cooling. Th is loop provides an additional barrier between the potentially contaminated systems and the plant service water discharged to the circulating water system.

9.2.2.2 System Description

The RCC is a closed loop system which provides parallel-flow cooling to auxiliary equipment in the primary containment, the reactor building, and radwaste bu ilding. The system consists of pumps, heat exchangers, tanks, piping, valves, and in strumentation as shown in Figure 9.2-2. Each of the three pumps and three heat exchangers are ha lf capacity based on maximum normal cooling requireme nts. Heat is removed from the RCC system by the plant service water system. The plant service water is passed through the tube side of the RCC heat exchangers and the closed-loop wa ter is passed through the she ll side. A 550-gal surge tank accommodates volume changes from thermal expans ion and contraction.

Makeup water to the system is supplied to the surge ta nk by the demineralized water system.

The RCC cools the following equipment:

a. Reactor recirculation pumps,

b. Drywell fan unit cooling coils,

c. Drywell equipment drain condenser, C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 9.2-4 d. Reactor water cleanup (RWCU) no nregenerative h eat exchangers, e. Reactor building equipment drain heat exchanger, f. Control rod drive pumps,

g. Fuel pool heat exchangers,

h. RWCU recirculating pumps, and

i. Offgas glycol refrigeration machines.

The RCC piping inside primary containment is constructed in accordance with ASME Section III, Class 3, and Seismic Category II requirements. The R CC piping outside of primary containment is construc ted in accordance with ANSI B31.1. In the reactor building and inside primary containment, RCC piping is supported to Seismi c Category I. In the radwaste building, RCC piping is supported to Seismic Categor y II requirements. System primary containment penetrations and isolation valves are constructed to Seismic Category I, and ASME Code Section III, Cla ss 2, requirements. The RCC pumps are constructed to Seismic Category II, and RCC heat exchangers are constructed to ASME Code Section VIII and the Standards of the Tubular Exchanger Manufacturers Associa tion, Class R. The system design pressure is 195 psig. The design temper ature is 150°F for all piping except the RWCU nonregenerative heat exchanger di scharge piping. The piping fr om the RWCU heat exchanger to the 18-in. return line has a design temperatur e of 175°F. The system operates at pressures and temperatures below the stated design values.

Equipment Design Parameters Closed Cooling Water Pumps (R CC-P-1A, RCC-P-1B, RCC-P-1C)

Quantity 3

Driver 200 hp motor - ac

Design capacity 2500 gpm

Head 200 ft

Speed, rpm 1775 Closed Cooling Water Heat Exchangers (RCC-HX-1A, RCC-HX-1B, RCC-HX-1C)

Quantity 3

Heat duty 25 x 10 6 Btu/hr C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-04-052 9.2-5 Shell side Tube side

Flow rate 2500 gpm 3333 gpm

Inlet temperature 115°F 85°F

Outlet temperature 95°F 100°F

Pressure drop 11.2 psig 5.0 psig

A chemical addition tank (RCC-TK-2) and metering pump (RCC-P-

2) were included in the original design for the addition of corrosion inhibiting chemicals if required. This chemical addition equipment is not normally used.

The RCC pumps supply water to the specified a uxiliary plant equipment during all modes of operation except under LOCA conditions. Fo llowing loss of normal power and under nonLOCA conditions, the RCC pumps ar e supplied from standby power.

9.2.2.3 Safety Evaluation

The RCC provides a barrier betw een nonessential, potentially cont aminated systems, and the plant service water discharged to the circulating water system. A radiat ion monitor is provided in the RCC to detect inleakage to this system from any contaminated system.

Leakage is also detected by pressure instrumentation in the system supply headers or by monitoring system surge tank level.

Portions of the RCC which penetrate the prim ary containment are provided with containment isolation valves which can be remotely actuated by the operator in the main control room and close automatically on the high drywell pressure (F) and the reac tor vessel low water level (A) containment isolation signals.

A pressure bypass line exists around the inboard isolation valve on the return line to prevent overpressurization of the piping between the inboard and outboar d isolation valve. The bypass line has a check valve that acts as an inboard isolation valve in parallel with the normal inboard isolation valve.

A single-failure analysis has not been provided for the RCC since this system serves only nonessential systems and is not required to perform a safety function. See Section 3.6 for a discussion of piping failures.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 9.2-6 9.2.2.4 Testing and In spection Requirements Pumps in the RCC are pr oven operable by their continuous us e during normal plant operation.

The motor-operated isolation valves are tested to ensure that they are capable of opening and closing by operating manual switches in the main control room and observing the position lights. The check valve in th e bypass line across the inboard is olation valve on the return line is tested to ensure that it is capable of ope ning and closing by manual means during an outage.

Routine visual inspection of the system compone nts, instrumentation, and trouble alarms is adequate to verify system operability.

9.2.2.5 Instrumentation Requirements

The RCC is a balanced constant-flow system. Local pressure and temp erature gauges and test points are provided throughout the system. Temperature in the primary containment is

transmitted to the main control room, as is pressure downstream of the closed cooling water heat exchangers. A temperature switch downstream of the heat exchangers actuates an alarm in the main control room on high water temperature. When an operating pump trips, the standby pump is automatically brought on line.

Alarms are actuated in the main control room following a decrease in pressure sensed by pressure switches on each pump discharge.

The system water volume is main tained from the plant deminera lized water system by a level control valve actuated by level switches in the surge tank.

9.2.3 PLANT

MAKEUP WATER TREA TMENT AND DEMINERALIZED WATER SYSTEMS

9.2.3.1 Design Basis

a. The plant potable water system provi des the supply water for the plant makeup water treatment system. A vendor furn ished and maintained trailer-mounted demineralizer provides the purified water to the demi neralized water system using the potable water source.

b. The demineralized water system is designed to utilize the demineralized effluent from the plant makeup water treatment system to provide demineralized water to the condensate storage tanks and to di stribute demineralized water throughout the plant.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-03-046 9.2-7 9.2.3.2 System Description

9.2.3.2.1 Plant Makeup Water Treatment System

The plant makeup water treatme nt system is shown in Figure 9.2-3. Raw water is normally supplied from the Columbia River through a br anch line of the tower makeup system (TMU).

From this line a branch is rout ed to the service bu ilding basement and is isolated. Another branch line continues to pump house 3, where raw water is processed into potable water. Here the water is chemically deflo cculated and transferred to a 300, 000-gal potable water storage tank. Sodium hypochlorite is in jected into the potable storage tank influent line. In case the TMU is temporarily not in servic e, water is supplied from a cr oss tie with the WNP-1 potable water system.

The trailer-mounted demineraliz er is supplied with water from the potable water pump PWC-P-103 which runs continuously. Demineralized water is pumped from the trailer demineralizer to tank DW-TK-1 in the basement of the service building through a tank level control valve.

Demineralized water is stored in a 50,000-gal epoxy-lined car bon steel demineralized water storage tank (see Figure 9.2-4 and Section 9.2.3.2.2).

The effluent from the trailer-mounted demineralizer is continuously recorded with associated alarms.

9.2.3.2.2 Demineralized Water System

The demineralized water system shown in Figure 9.2-4 consists of a 50, 000-gal steel, epoxy-lined storage tank (DW-TK-1) with two 200-gpm transfer pumps with a 265-ft head (DW-P-1A, DW-P-1B). Among the components being supplied by the pumps are the following:

a. Condensate storage tanks,

b. RCC surge tank,

c. Diesel generato r expansion tanks, d. Auxiliary boiler condensate return tanks, e. Chemical addition tanks, f. Stator cooling makeup,

g. Refueling floor service boxes,

h. Decontamination facilities,

i. Flushing and washdown services,

j. Sealing water,

k. Laboratory water,

l. Reactor containment services, C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 9.2-8 m. Instrument flushing water, and n. CJW surge pipe.

During operation one demineralized water transfer pump runs con tinuously to keep the system headers pressurized to s upply demineralized water on demand. The operating pump recirculates a minimum flow back to the stor age tank to prevent dama ge to the pump under low flow conditions.

9.2.3.3 Safety Evaluation

The demineralized water and plant makeup treatment systems are not required for a safe shutdown of the plant and are ther efore, not safety related. A failure of these systems will not affect the operation of systems that are vital to safe shut down of the plant. Alarmed instrumentation is provided to prevent delivery of out-of-specification wa ter to safety-related systems or components. The pi ping associated with the demine ralized water and plant makeup water treatment systems is designed in accord ance with ANSI B31.1 except for piping that penetrates containment is designed ASME III, Class 2. All demineralized water and plant makeup water treatment system piping is classified Seismic Category II except that piping penetrating or routed inside pr imary containment, which is classified Seismic Category I. In the reactor and diesel generator buildings and inside primary containm ent, piping is supported to Seismic Category I requirements.

9.2.3.4 Testing and In spection Requirements

The demineralized water and pl ant makeup water treatment syst ems are in daily use and as such do not require periodic testing to ensure operability. Grab samples are periodically tested in the laboratory to verify filte r and deminerali zer performance and to ascertain stored water quality. Periodic visual inspection and preventive maintenance are conducted.

The major components of these systems were subject to ma nufacturer shop tests including performance and hy drostatic tests.

Prior to initial operation the piping systems were subject to a hydrotest in accordance with their respective code groups. In addition, the completed systems were subjected to acceptance testing during plant startup as discussed in Section 14.2.

9.2.3.5 Instrumentation Requirements

Instrumentation and controls for the plant ma keup water treatment a nd demineralized water systems are provided to cont rol and monitor the operation of each of the subsystems.

Level controls in the demineralized water storage tank maintain the tank at the proper level by opening and closing a valve in the ma keup demineralizer discharge piping.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 9.2-9 9.2.4 POTABLE WATER AND SANITARY DRAIN SYSTEMS

9.2.4.1 Design Bases

The function of the potable water system is to supply potable water throughout the plant, provide supply water to the plan t makeup water treatment system , and can be used to supply makeup water to the spray ponds.

The potable water system is designed to provide cold and/or hot water to the points of potable water usage such as lavatories, urinals, toilets, showers, eyewashes, sinks, electric drinking water coolers located in the various plan t buildings, and site grounds irrigation.

The sanitary waste syst em is designed to drain the noncontam inated sanitary waste from the various buildings served.

All system fixtures, piping, and component material, as well as the fabrication, erection, and testing comply with appli cable sections of appropri ate codes and standards.

The potable water storage tank is designed and constructed in accordance with ASME Section VIII and Seismic Category II requirements.

9.2.4.2 System Description

9.2.4.2.1 Potable Water System

The potable water system is shown in Figure 9.2-5. Water is supplied to the potable water storage tank (PWC-TK-1). The makeup water treatment system is de scribed in Section 9.2.3. The potable water tank is locate d in the basement of the serv ice building from which potable water is pumped to all buildings serviced.

Potable hot water is provided to the various plant buildings by individual electric hot water heaters.

9.2.4.2.2 Sanitary Drain System The plant sanitary waste system is shown in Figure 9.2-6. All noncontaminated sanitary waste from all buildings is directed in to this system. Potentially contaminated drainage, such as that from the radiochemical laborator y sinks, is routed to the radioactive drainage system as described in Section 11.2.

All sanitary drains in the plant are collected in to a main 6-in. sewage header which leaves the plant building complex, below grade, from the se rvice building. All wast e lines are vented to C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 9.2-10 the atmosphere via roof vents in accordance with applicable codes except those in the reactor building, which are vented to the main r eactor building exhaust system (see Section 9.4.6) and in the main control room are vented to the main control room exhaust system.

All sanitary lines drain by gravity flow into th e main 6-in. sewage pipe which connects to an 8-in. line leading to a centrally located sewage treatment fac ility serving WNP-1, Columbia Generating Station (CGS), WNP-4, the Kootenai Building (Plant Support Facility), and the Department of Energy's 400 Area (FFTF facility

). Low levels of radioactive materials, particularly tritium from the 400 Ar ea, are expected to be present in the treatment facility as a result of processing these waste st reams. The operation of the sanitary waste trea tment facility is regulated by the State of Washington.

9.2.4.3 Safety Evaluation

The potable water and sanitary drain systems are of conventi onal design and ha ve no safety function. All contaminated drains are routed to the liquid radwaste sy stem as discussed in Section 11.2. All system piping except those in the reactor building and radwaste building (control room section) are Seismic Category II as defined in Section 3.2.1. The piping in the reactor building and radwaste building (control room section) is Seismic Category I.

9.2.4.4 Testing and In spection Requirements

Routine monitoring during normal operation verifies the system is performing acceptably.

9.2.4.5 Instrumentation Requirements

The system is provided with suffi cient instrumentation and contro ls to monitor and operate the system to meet the normal operati ng requirements of the facility.

9.2.5 ULTIMATE

HEAT SINK

9.2.5.1 Design Bases

a. The ultimate heat sink (UHS), a spray pond system, supplies cooling water to remove heat from all nuclear plant equipment that is essential for a safe and orderly shutdown of the reactor and to maintain it in a safe condition;

b. The UHS is capable of accomplishing its safety function for a normal cooldown or an emergency cooldown following a LO CA without the avai lability of offsite power. The sink provides this cooling cap ability for a period of 30 days without outside makeup (except following a to rnado). Provisions are made for replenishment of the sink to allow c ontinued cooling capab ility beyond the initial 30-day period. The sink will accomplish its safety function despite the C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-027 9.2-11 occurrence of the most severe site-rel ated natural events including earthquake, tornado, flood, or drought; and

c. Regulatory Guide 1.27 complia nce is described in Section 1.8.

9.2.5.2 System Description

During all normal operating conditions, incl uding normal shutdown as well as emergency conditions, waste heat from the reactor auxiliari es is transferred to the UHS by the standby service water (SW) system.

The UHS consists of tw o concrete ponds. The concrete ponds provide suction and discharge points for the redundant pumping and spray facilities of the SW system. The pond and pump house arrangements are shown in Figure 9.2-7. The ponds and pump houses are designed to Seismic Category I requirements. Standby service water loop A draws water from pond A, cools the Division 1 equipment required for sa fe shutdown, and disc harges through a spray ring into pond B for heat dissipation. Simila rly, SW loop B draws wa ter from pond B, cools Division 2 equipment, and discharges through a spray ring into pond A. The high-pressure core spray (HPCS) SW system draws water from pond A, cools Division 3, a nd discharges without spray into pond A. A siphon between the ponds allows for water flow from one pond to the other.

The spray system illustrated in Figure 9.2-7 consists of two annuli of spray trees, one for each of the SW subsystems. Each annulus is 140.0 ft in diameter and contains 32 spray trees equally spaced (13.75 ft between vertical centerlines) on the circumference. The vertical trees are serviced by the annulus wate r pipe, 20 in. in diameter, mount ed above the water level.

The annulus pipe is fed by the main header fr om each respective pump h ouse. Each spray tree consists of a vertical riser pipe or trunk 8 in. in diameter and horizontal limbs of 1.5-in. pipe. The limbs are attached to the riser at 2 ft-8 in. intervals of heights and are rotated at 90 degree subsequent angles from each other so that the arms resemble a counterclockwise helix with

increasing height. The arms radi al to the annulus are 4 ft-6-7/

16 in. long. The lowermost arm is a tangent arm. The arms tangent to the annul us pipe are 3 ft-6 in. long. Spray nozzles are located at the end of each arm and are connected by fittings. The orientation of every nozzle is radially inward with an angle of 55 degrees upward from horizontal. The nozzles are 1-1/2-CX-27-55 Whirljet nozzles supplied by Sp raying Systems Company. Since each tree nozzle is located at a different elevation, each nozzle pressure is different. The uppermost nozzle water pressure is approximately 17 psig , and the total water flow from a tree is approximately 300 gpm.

The HPCS SW flow is treated as a straight heat dum p in the thermal analyses.

The combined water volume of the spray ponds is adequate to prov ide cooling water for 30 days without makeup. Althoug h the ponds are not routinely used for cooling during normal

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.2-12 operation, some small losses are to be expected due to normal evaporation from the surface, minor leakage from the ponds, and occasional blowdown needed to maintain water chemistry. A gravity makeup line is provided from the circulating water pump house to the spray ponds; however, the required pond level will normally be maintained by makeup directly from the TMU water pumps (see Section 10.4.5) or directly from the potab le water system. The TMU is the makeup source credited in the event of wa ter loss due to a tornad

o. Design parameters for the spray pond are given in Tables 9.2-1 and 9.2-2. An SW pump is located in each spray pond pump house along with its associated equipment so that an accident, such as a fire or pipe break associated with one pump, would not affect the operation of the redundant pump.

The bottom of the pump sump is depressed below the pond bottom.

This ensures th at there is still sufficient submergence fo r the pumps at the lowest possi ble water level in the pond. A sand trap and screen precede the pump sump to prevent heavy debris from entering the pump sump area. A skimmer wall and fixed scr een prevent floating debris from entering the pumps.

A spray ring bypass is provided that allows water to be returned directly to the pond through use of manually operated dump and spray ring isolation valves.

This allows pond temperature to be controlled during cold weather operations, mixing of the pond volumes, minimization of drift losses, and other situations where bypass of the spray rings is desirable. If the spray ring header is bypassed, it will be realigned to allow spray mode of operation before pond temperature exceeds 70°F. Analysis demonstrates that at pond temperatures less than or equal to 70°F, sufficient time is av ailable under accident conditions to permit realignment to the spray mode prior to pond te mperature exceeding 77°F.

To prevent adverse operation during freezing weather, all SW piping and components are either below the frost line, with in heated buildings, h eat traced or draine d when not in use.

To allow for infrequent maintenance of the spray pond piping and pipi ng supports, a cross connection, consisting of a re movable spool piece and two is olation valves, is available between the two redundant loops of SW. A re movable siphon plug is available to prevent one pond siphoning to the drained pond. The cross connecting spool piece and siphon plug are normally not installed and are only used while the plant is shut dow n for refueling. This cross connection allows water to be pumped back directly to the pond from which it was removed, thus allowing for the other pond to be drained.

9.2.5.3 Safety Evaluation

An oriented spray cooling system (OSCS) is utilized for cooling the water inventory of the

UHS. The OSCS was developed as a result of intensive analytic al studies and experimental verification over a period of more than 6 years. Details of the OSCS experimental and

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.2-13 analytical developmental effort s are described in Topical Re port, Oriented Spray Cooling System (OSCS) for Ultimate Heat Sink Application (UHS), I-R 100, which has been submitted for Nuclear Regulatory Commission staff review.

The thermal performance model is based on the correlation of the Canadys test data described in Section 3.1 of the Topical Report, I-R 100. The re sulting KAV/L for this application is 2.66. This includes a 10% dera te of the KAV/L to cover co nservatively the data scatter experienced at Canadys. Since the KAV/L represents the performance of the specified geometry and nozzle pressure, the KAV/L combined with the meteorological data are sufficient to determine the syst em cooling performance. The performance predicted by this model is modified as shown below.

CR = (-0.761 + 0.009 x TWB) +

(0.2677 + 0.004029 x TWB) x CP

+ (0.001179 - 7.14 x 10

-6 TWB) x CP 2

CR = cooling range TWB = wet bulb temperature CP = cooling potential

The system model for both the thermal performance analysis and the mass loss analysis was based on the following assumptions:

a. The pond contains total inventory upon onset of LOCA less 0.5 ft for sedimentation of the pond basin;

b. Water losses result only from drift, leakage, evaporation of the sprayed droplets and evaporation due to heat rejection on the pond surface;

c. All the heat transfer is accomplished by evaporation, none of the heat transfer is accomplished by sensible heat transfer;

d. The first 3 days of the thermal performance analysis assume the worst single day of record conditions (Table 9.2-3

). It was found that repeating the worst day conditions for 3 days resulted in th e maximum peak. The 4 through 30 days are the average meteorological conditions of the worst 30-day period of record (Table 9.2-3

). For conservatism, the root m ean square (RMS) average wind speed during each period is used rath er than the diurnal variation;

e. The 1 through 30 days of the mass loss analysis are the average meteorological conditions of the worst 30-day period of record (see Section 2.3.1.2 and Table 9.2-3). The analysis assumes a mass loss due to drift of 1.02% of the spray flow. This value is an upper bound obtained from results of tests

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-08-032 9.2-14 performed on the spray ponds. Its conservatism was c onfirmed by testing reported in Referenc e 2 and shown in Figure 9.2-8

f. For the design basis mass loss case eval uation, the spray flow is continuous for one spray ring for the first 3 days, with one SW pump in opera tion. The sprays are cycled on and off for the 4 through 30 days depending on pond temperature.

The sprays are turned on wh en the SW temperature in the suction pond reaches 85°F and are turned off when the temp erature drops below 80°F. This method of operation resulted in one cycle on and off per day;

g. Minor leakage was observed in testing of the spray ponds, so test results were used to establish a leakage rate of 3.94 gpm total for both ponds (170,000 gal per 30 days);

h. While not a design basis case, mass loss was also ev aluated assuming two spray rings in continuous operation for 2 days, followed by one spray ring in cyclic mode described in item f;

i. The design basis fuel pool cooling heat load is included;

j. For design basis analysis, offsite power is lost and Division 2 diesel fails to start, resulting in a loss of the pond A sp ray header. (Divisi on 1 heat loads are slightly higher); and

k. The major heat loads considered are reactor core decay heat, sensible heat from both the coolant and the r eactor, fuel pool decay heat, pump work, and the heat removed from the station auxiliaries.

These heat load s are listed in Table 9.2-4 (using the calculated heat load values) and Table 9.2-5. The fuel pool decay heat generation rate, listed in Table 9.2-4 as variable, is found in Table 9.2-5. No credit was taken for heat sinks in the primary containment other than the suppression pool volume.

The RMS average wind speed during the selected 30-day period for the mass loss analysis was 6.91 mph. The drift loss was based on the calculated drift value at the RMS wind speed. The mass loss analysis thus demonstrates that the sp ray ponds contain sufficient water inventory to meet drift losses significantly higher than expected.

Figure 9.2-8 shows the conservatism demonstrated in confirma tory drift loss testing.

The analyses assume an initial temperature of 77°F. This is approximately the highest monthly average temperature expected if the sprays are not operate

d. To maintain the pond temperature below this limit, the spray headers will be operate d and/or river water makeup to the cooling towers will be diverted through th e spray ponds. Analyses were performed which demonstrate that the above operations can maintain the spray pond below 77°F.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-08-032 9.2-15 An analysis was conducte d which verified failure of Division 1 or Division 2 power results in the most severe service water transient. If the failure was postulate d in Division 3 (HPCS) instead of Division 1 or 2, the peak pond temperature is lower. The HPCS SW flow is a straight heat dump; therefore, inasmuch as the spray pond is concerned, it raises rather than lowers the temperature transient.

The resulting peak spray pond temperature, 89.5°F, predicted by the "worst case" analysis is below the 90°F service water te mperature assumed in the anal ysis performed in Section 6.2.1 for containment heat removal, adding further c onservatism to the containment temperature and pressure transients therein presented. The service water temperatur e, however, exceeds the design basis temperature, 85°F, for s hort periods of time as shown in Figure 9.2-9. A temperature of 90°F has been evaluated for those rooms served by emergency heating, ventilating, and air cond itioning (HVAC) equipment, and room temperatures remain within the design limit.

A sensitivity study was performed to determin e the effect of the RHR heat exchanger effectiveness on the spray pond temperature transient. The RHR heat exchanger effectiveness varies with the amount of fouling and with th e flow rates. The RH R heat exchanger flows different from the rated values in Table 6.2-2 are anticipated only if th e operator delays or fails to close the RHR heat exch anger shell side bypass valve as discussed in Section 6.2.2.3. Anticipated variations in fouling were determined to have essentially no affect on the spray pond temperature transient following a design ba sis LOCA. However, the combination of low SW to RHR cooling flow, low overall SW flow , design RHR HX fouling and no containment spray does adversely affect peak pond temperature. For this re ason the temperature case uses low flow/design fouling while the mass loss case uses high flow/clean heat exchnagers.

The results of the design basis mass loss analysis assuming an unfouled heat exchanger is shown in Figure 9.2-10 and is tabulated in Table 9.2-6.

Drift losses following loss of ma keup to the ponds may be controlled during two spray ring operations by bypassing the spray header on one pond whenever sp ray pond temperatures drop below approximately 80°F. Continuous, simulta neous operation of both spray rings is not required after a LOCA. Since the two SW loops are redundant to each ot her, operators will be able to secure any redundant safe shutdown equipment when they determine that the peak temperatures have been passed.

Table 9.2-7 lists the available sources of makeup water to provide continued cooling beyond the initial 30-day period. This table assumes that offsite power is restored within the 30 days.

No credit is taken for the water stored in the cooling tower basi ns. However, it is expected that this water will not be inst antaneously lost and may flow to the pond for the same period of time. Table 9.2-7 also summarizes the effects of natural phenomena and of a LOCA on the water supplies to the spray pond.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 9.2-16 The possibility of a tornado passing over the spray pond and removing a significant amount of water is considered a credible event. For this reason, the makeup water pump house is designed to be tornado proof, with all piping an d electrical power supply between the plant and the pump house underground with adequate soil cover to provide protection from tornado-generated missiles. Since it is not credible to assume an earthquake coin cident with a tornado, this system need not be Seismic Category I. Two 12,500-gpm tower makeup water pumps are provided, one powered from each emergency diesel generator.

Should pond water be lost due to a tornado, one of these pumps will be started to provide makeup. If the spray headers are damaged by a tornado-generated missile, cooling is provided by a feed-and-bleed mode of operation. In the feed-and-bleed mode, cooling water is supplied to th e spray ponds from the makeup water pump house. The service water system take s suction from the spray ponds to provide cooling to safe shutdown equipment. The cooling water is then routed to tornado-protected underground circ ulating water piping and discharg ed to the circulating water basin.

The design basis of the UHS in cludes a 6-in. sedimentation allowance for water inventory considerations. This allowan ce includes all forms of accumulation, such as dust, silt, or volcanic ash. The design basis ashfall is 3 in. which bounds th e Mount St. Helens eruption of May 18, 1980. The spray ponds will be cleaned whenever the sedimentation reaches 3 in., which ensures adequate wate r supply even in the event of a design basis ashfall.

With regard to SW pump operati on following a design basis ashfall, experience has shown that there may be some increased seal leakage as a result of ashfall.

An evaluation was performed which concludes that the pumps would remain opera ble during a volcanic ashfall condition.

Bearing life could be s hortened, and bearings would be examined for wear when possible following a significant ashfall.

While the plant is shut down it may be necessary to drain one spray pond for maintenance using the cross connecting spool piece and siphon plug to provide a flow path back to the same pond from which the water was pumpe

d. For this infrequently used mode, there is less than a 30-day supply of water available in one pond wi thout makeup. However, with the reactor vessel depressurized, flooded up to 22 ft or greater above the r eactor flange, the fuel pool gates removed, and the reactor at a relatively low temperature (less than about 140°F) prior to removal of a pond from service, the risk to public safety is very sma ll. Alternative makeup water sources for the UHS can be made readil y available. Experi ence has shown that temporary pipe lines can be run fr om the river in about 3 days.

Even in the unlikely event of a loss of the siphon plug, the water inventory would not be lost and the system would remain operable. In that event, wate r in the inactive pond could be tr ansferred back to the operating pond via portable pumps.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 9.2-17 9.2.5.4 Testing and In spection Requirements Availability is ensured by peri odic functional tests and inspecti ons as required by the Technical Specifications.

9.2.5.5 Instrumentation Requirements

The spray pond is equipped with redundant level and temperat ure sensors which are alarmed and indicated in the main control room.

In the event that the spray pond level falls below the mini mum level required for 30 days of cooling, an alarm is sounded and makeup to the spray ponds is provided using the TMU.

High and low temperature alarms are provided. If the pond wate r temperature approaches the upper design limit, the spray syst em is manually initiated to lowe r the temperature. On a low water temperature signal, return water is dumped directly into the ponds to prevent spray trees and spray headers from icing.

9.2.6 CONDENSATE

SUPPLY SYSTEM

9.2.6.1 Design Bases

The condensate supply system (COND) is designed to:

a. Store and provide a conde nsate supply to the reactor core isolation cooling (RCIC) system, the HPCS system, and the RHR loops;

b. Maintain an adequate level of condensate in the condenser hotwell;

c. Provide a condensate supply for the control rod drive pumps;

d. Provide makeup water to the spent fuel pool;

e. Provide condensate for va rious radwaste processes;

f. Facilitate testing and/or flushing of the HPCS, low-pressure core spray (LPCS), RHR, and the RCIC;

g. Receive and accommodate a surge volume for condensate returned to the storage tanks after treatment in the liquid radwaste system; and

h. System piping is constructed to ANSI B31.1. Condensate storage tanks are constructed to ASME Secti on III, Class 3 requirements.

System piping inside

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 9.2-18 the reactor building is Seismic Category II, supported to Seismic Category I requirements. All other piping and system pumps are designed to Seismic Category II requirements. The radwaste building condensate supply pump and the condensate filter demineralizer b ackwash pump are constructed to ASME Code Section III, Class 3. The react or building condens ate supply pump is designed to the Standards of the Hydraulic Institute.

9.2.6.2 System Description The demineralized water system and the liquid radwaste system are the primary sources of makeup water to the c ondensate storage tanks.

The COND is shown in Figure 9.2-11. The system consists of two storage tanks (COND-TK-1A, COND-TK-1B) each with a nominal capacity of 400,000 gal and equipped

with electric heaters, a reac tor building condensate supply pump (COND-P-3), a radwaste building condensate supply pum p (COND-P-4), a condensate fi lter demineralizer backwash pump (COND-P-5), and necessary piping and instrumentation. The tanks are manufactured with a design pressure of atmospheric plus full static head and maximum design temperature of 140°F. Minimum operating temperature of th e tanks is 40°F. The tanks are designed to withstand a wind load of 20 psf on the vertical projected area of the tank and a snow load of 20 psf.

The radwaste building condensate supply pump and the condens ate filter demineralizer backwash pump each are designed to supply 1535 gpm at 185 ft total head. The radwaste condensate supply pump has a seco ndary operating point of 500 gpm at 220 ft total head. The reactor building condensate supply pump is desi gned to supply 200 gpm at 220 ft total head.

A minimum inventory of 135,000 gal in the condensate storag e tanks is reserved for the RCIC and HPCS pumps. This ensures the immediat e availability of a su fficient quantity of condensate for emergency core cooling, and reactor shutdown, as discussed in Section 9.2.6.3 and Station Blackout as discussed in Appendix 8A.

Makeup for the condenser hotwell is gravity fed from the storag e tank. Bleedoff water from the condensate system is return ed to the storage tanks from the discharge of the condensate demineralizer.

A separate line is provided to supply the control rod drive pump s with condensate. Condensate is supplied for various reacto r building services, including fu el pool makeup by the reactor building condensate supply pum

p. The condensate storage tank can be drained to the condenser hotwell. Ina dvertent overflow of the tanks is co llected in the concrete retaining basin surrounding the tanks. This water can be drained to the radwaste system for processing if sampling indicates that the water is radioactively contaminated. Rain water collected in the retaining basin can be drained to the radwaste system for processing.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 9.2-19 9.2.6.3 Safety Evaluation

The condensate storage facilities are not required to ensure any of the following:

a. The integrity of the RCPB,

b. The ability to shut down the reactor and maintain it in a safe shutdown condition, or

c. The ability to pr event or mitigate the consequen ces of accidents which could result in potential offsite exposures in excess of the guideline exposure of 10 CFR Part 100.

Although a minimum of 135,000 gal is maintained in the condensate storage tanks as a source of water for the RCIC and HPCS pumps, the supply of water in the suppression pool is the emergency source of water for these pumps. The reserve of water is ma intained by monitoring the level in the condensate st orage tank and by preventing conde nsate transfer when this reserve level is reached. Th e RCIC and HPCS pumps are gr avity fed from the condensate storage tanks.

The condensate storage tanks are Seismic Category II; however, they are located inside a Seismic Category I concre te dike which is designed to retain the condensate from both tanks.

Drainage from the dike is routed to the radwaste system for processing. During precipitation, drainage from the dike is sampled and analyzed for radioactivity before being discharged to a

Turbine Building sump and proce ssed in the radwaste system.

The evaluation of radiological considerations for the COND is presented in Section 2.4.13.3. Since the dike is designed to contain any condensate from a postulated tank rupture in conjunction with the heaviest recorded precipitation, the resulti ng offsite dose rate from this occurrence would be no greater than the value presented in Section 2.4.13.3.

For corrosion protection, the tanks are made of carbon steel with a 1/16-in. corrosion allowance and lined with a modifi ed phenolic coating. Quality control for the application of the coating was in acco rdance with ANSI N101.4. The interior surfaces of the tanks were blasted to white metal in accordance with SSPC-SP-5 prior to the application of a minimum of 10 mils dry film thickness of Plasite 7155. Coating repairs are made with a compatible amine cured epoxy coating.

9.2.6.4 Testing and In spection Requirements

The components were inspected and cleaned prior to installation into the system. The condensate storage tanks side wa ll, side wall to bottom, and no zzle joints were examined in

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 9.2-20 accordance with ASME Section II I, Subsection NC-5000. The tank thickness was tested by magnetic means in accordance with ASTM E376. Underground portions of the HPCS and RCIC piping welds were 100% radiographed to ensure the integrity of the piping.

Periodic functional tests and insp ections as required by the Tec hnical Specifications ensure system availability for nonrou tine functions. Routine mon itoring during normal operation verifies satisfactory system operation to support plant requirements for normal operation.

9.2.6.5 Instrumentation Requirements

Condensate storage tank level is monitored in the main contro l room. High- and low-level alarms are provided to prevent overflow and to prevent the water level from dropping below the required reserve level for reactor pressure vessel (RPV) ma keup. Level switches provide low-low annunciation and interloc k with the HPCS and RCIC sy stems. Building condensate supply pumps are provided with manual controls for maintaining condensat e supply pressure in the system headers so that condensate is available fo r system process services, based on demand.

The following parameters apply to the condensate storage tanks.

Reserve capacity for RPV make up is provided between the set points of the low-level switch and the low-low-level switch.

The elevation differential provides a reserve capacity of approximately 67,50 0 gal per tank.

Thermostatically controlled tank heaters are prov ided to maintain water temperature in the tanks at or above a nominal 40°F at all times. All above-ground piping th at contains water is heat traced to prevent freezing.

System logic diagrams are given in Chapter 7.

9.2.7 STANDBY

SERVICE WATER SYSTEM

9.2.7.1 Design Bases

a. The SW system is designed to remove heat from plant systems that are required for a safe reactor shutdown following a LOCA;

b. The system is designed to remove reactor decay heat from the residual heat removal system during normal plant shutdown;

c. The system is designed to perform its required cooling wa ter function following a LOCA, assuming a single active failure;

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 9.2-21 d. The system is designed to provide a means of flooding the vessel and containment, if required during the post-LOCA period;

e. The system is designed to provide a long-term coo ling and makeup source to the fuel pool cooling system following a lo ss of the RCC syst em which is the normal source of cooling to fuel pool cooling and cleanup (FPC); and

f. The system is constructed to Seismic Category I and ASME Code Section III, Class 3, requirements with the exception of that portion to and from the plant cooling towers, which is constructed to ANSI B31.1 and Seismic Category II requirements; the spray bypa ss line discharge orifice and mating flange, which are also constructed to ANSI B31.1 a nd supported Seismic Category I, the keep

full subsystem located in the pump house which is constructed to ANSI B31.1 and supported Seismic Category I, and the spray pond crossover piping which is constructed to ANSI B31.1 a nd supported Seismic Category I.

9.2.7.2 System Description

The SW system includes vertical service water pumps located adjacent to the two spray ponds in two separate pump houses de signed to Seismic Category I crite ria. The pumps discharge to three independent piping systems which serv e emergency core cooling system (ECCS) equipment, auxiliary plant equipment, a nd reactor shutdown coo ling equipment (see Figures 9.2-7 , 9.2-12 through 9.2-14).

The SW pumps consist of two independent 100% capacity pum ps each supplying normal and emergency shutdown cooling equipment as shown in Table 9.2-4. A third 100% capacity HPCS service water pump (hous ed in pump house A) supplie s the HPCS system cooling equipment as shown in Table 9.2-4.

Each of the three SW pumps receives its power from independent electrical buses. Each bus is powered from the offsite power supp ly or its own diesel generator.

During the normal and emergency shutdown modes of operation, water is circulated from the spray ponds to the equipment requiring cooling, a nd returned to the ponds. The two SW systems return via the spray rings prior to recycle. The HPCS se rvice water return s directly to pond A with no sprays. Spray po nd operation and a description of its ability to dissipate waste heat from the plant during normal, shutdown, and accident conditi ons are provided in Section 9.2.5.

The system is designed such that on receipt of an ECCS start signal, all three pumps are started automatically.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-054 9.2-22 The two spray ponds are sized to have a combined equivalent storage for at least 30 days of operation, assuming no makeup and maximum evaporation and drift losses (see Section 9.2.5).

The SW system is chemically tr eated to control biological growth and to minimize corrosion.

A small keep-full subsystem is attached to the SW system piping. The keep-full subsystem was originally designed to provide back pressure on the pump discharge valve so that maintenance repairs on the valve could be minimized (i.e., the k eep-full subsystem wa s never intended to keep the system piping complete ly full). The pump discharge valve design and the system start sequence were changed so that back pressure on the pump discharge valve during the start

sequence was no longer necessary.

The SW system is able to st art from a partially drained condition without damage to system compone nts. The keep-full subsystem has been

deactivated and spar ed in place because it no longer provides a useful function.

The spray ponds are provided with makeup water by the TMU system or the potable water system. The makeup water system supplies Colu mbia River water to the cooling towers or spray pond to replace water lost during normal operation due to evaporation and drift. The potable water system replaces normal spray pond ev aporation and drift losse

s. In addition, the makeup system is designed to replace spray pon d water lost during a tornado. To ensure system availability for this m ode of operation, the makeup syst em is designed to withstand a design basis tornado coincident with a loss of offsite power (see Section 9.2.5.2).

The SW system piping is carbon steel designed to 309 psig (some 150 psig), 150°F, and with a corrosion allowance of 0.080 in. Where SW pipi ng degradation due to erosion/corrosion or other degradation mechanisms oc curs, stainless steel material may be substituted to prevent future recurrence.

Standby makeup water to the fuel pool system ca n be supplied through nor mally closed Seismic Category I, Quality Class 1, isolation valves by remote manual opera tion from the control room for SW-V-75A and SW-V-75B a nd by manual operation for SW-V-75AA and SW-V-75BB (see Section 9.1.3.2.3).

EQUIPMENT DESIGN PARAMETERS Standby Service Water Pumps (SW-P-1A, SW-P-1B)

Quantity 2 Driver Motor - ac Design Capacity 10,300 gpm Head 500 ft C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 9.2-23 HPCS Service Water Pump (HPCS-P-2)

Quantity 1 Driver Motor - ac Design Capacity 1200 gpm Head 123 ft

9.2.7.3 Safety Evaluation The SW system provides cooling for plant equipment that is essential to a safe reactor shutdown following a design basis LOCA.

System failure mode and effects analyses of passive and active components of the SW system is presented in Table 9.2-8. Any of the assumed failures of th e SW system is detected in the main control room by indications and/or al armed from the various system instruments.

The SW is routed through the t ube side of the RHR heat excha nger, through the shell side of the FPC heat exchangers, and thro ugh the shell side of the RHR pump seal coolers. The RHR heat exchanger, the FPC heat exchanger, and the RHR pump seal coolers are the only potential sources of radioactive leakage into the standby water system.

A restricting orifice and bypass valve is installed on return lines of loops A and B to maintain sufficient back pressure to ensure flow to all components an d to maintain the pressure in the system at a level such that tube leakage in the RHR heat exchanger or pump seal cooler is from the service water system into the RHR system after the reac tor is depressurized following shutdown or after a LOCA. Liquid effluent radiation monitors are provided in each of the

two RHR heat exchangers outlet lines of the service water system to detect radioactivity resulting from a tube leak in one of the RHR heat exchangers. This condition can occur only for a short time when the RHR system is operated in the shutdown cooling mode in which the service water system pressure may be lower than the RHR system pressure. On detection of radioactive leakage in one of th e subsystems, that subsystem is isolated by operator action in the main control room and the cooling requirements are met by the redundant train.

• Per the CGS emergency procedures the operators are directed to utilize the alternate shutdown cooling mode when a degraded core c ondition has been identified.

The alternate shutdown cooling mode does not have the potential for radiation leakage into the SW system due to the lower RHR pressure. Consequently, radioactivity released to the spray ponds and/or cooling tower basins is minimized.

Under emergency conditions, i.e., loss of RCC sy stem, the SW system can be routed through the shell side of the FPC heat exchangers to provide long-term cooling. The SW system

• Note that the Safety Evaluation Report for CGS (NUREG - 0892) states in 5.2.5 that automatic isolation is provided. The response for CGS has always been by operator action.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 9.2-24 pressure is higher than the FPC system pressure, so that any leakage will be into the FPC system.

An intertie with the RHR system is provide d from the SW system supply header B which contains two remote manually ope rated isolation valves. These valves can be opened from the main control room in the event primary containment flooding is required following a LOCA.

An atmospheric syphon is provided to allow the transfer of water between the two spray ponds. In addition, the SW pumps are provided with double valved, normally closed bypass connections for transferring water from one pond to another.

Temperature controlled and/or ma nually operated throttle valves are located on the return side of all SW serviced coolers and heat excha ngers. The system is balanced for optimum operation and the throttle valves le ft in that position. The RH R heat exchanger service water outlet valve, while open during normal standby lineup, is interloc ked to open on starting of an SW pump to prevent excess flow conditi ons in other portions of the system.

Normal SW return is to the spray pond, with normally closed redundant isolation valves on the SW return line to the cooling towers. These valves are mechanically and electrically prevented from opening to preclude returning SW to the cooling towers during norm al system operation.

This is to ensure that the spray ponds always maintain th e capability to m eet the system requirements descri bed in Section 9.2.5 and allows control of water chemistry by minimizing dilution.

The non-safety-related SW keep-full subsystem ca nnot inhibit the ability of the SW system to meet its safety function. Keep-full piping is small and any break in that piping would result in a flow change to SW that is insignificant. The potential flooding of the valve pit area has been evaluated to ensure continued system performance should a br eak occur in th e deactivated keep-full system.

An orifice is installed on the discharge of each spray bypass line where it enters the spray pond. The purpose of the orifice is to mitigate cavitation damage that has occurred in the past due to the higher flow rate of the system in the spray bypass mode of operation. The orifice balances the flow in the bypass mode so it is appr oximately the same as that in the spray mode.

The orifice and its mating flange serve no safety-related f unction and serve to reduce the potential for long-term effects of cavitation. Failure of the orifice cannot affect the ability of the SW system to fulfill its safety function. Th e elbow to which the orifice is attached directs the return water into the pond after it passes through SW-V-165A/B, and failure of the orifice would not obstruct or limit flow of return water to the spray pond.

During normal plant operation, the SW pumps are not operating. If a LOCA occurs, the diesel generators and respective SW pumps start automatically. Conse quently, no operator action is

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 9.2-25 required following a LOCA to start the SW system and put the system into its LOCA operating mode.

9.2.7.4 Testing and In spection Requirements

The SW system is designed to permit periodic inspection of all active system components to ensure the integrity and capability of the system. For inservice inspection see Section 6.6.

The SW system is periodically tested as re quired by the Technical Specifications to ensure system availability and capability to perform its require d design functions.

Requirements for monitoring of heat exchangers cooled by SW are discussed in Generic Letter 89-13. Energy Northwest response to Generic Letter 89-13 outlines compliance with these requirements.

9.2.7.5 Instrumentation Requirements

Each of the SW discharge lines from the RHR heat exchangers c ontains radiation monitors to detect any radioactivity resulting from a tube leak in the RHR heat exchanger.

Flow indicators and/or switches are provided for most serviced components to indicate low flow. The radiation monitors have transmitters but no indicators due to the use of gravity flow. There are no permanently installed flow instruments for the cont rol room chilled water (CCH) since service water flow through the ch iller condensers is controlled by temperature control valves. Temperature indicators, temperature switches, pressure indicators, and/or pressure switches are located in each system to determine pump, individual cooling coil, cooler, or heat exch anger performance.

To avoid excessive system surge pressures, SW pumps SW-P-1A and SW-P-1B are started only if the associated pump discharge valve is closed. HPCS-P-2 can auto start with its discharge valve open, in order to minimize running the HPCS DG without service water cooling (see Section 9.5.5 for discussion of the diesel gene rator cooling water systems). All three pumps start automatically when the associated diesel generator is started. For each of the three pumps, the pump discharge valve automatically opens when the associated pump is running and the pump discharge pressure is greater than 50 psig. The system instrumentation is shown in Figure 9.2-12. 9.2.8 COMPRESSOR JACKET WATER SYSTEM

9.2.8.1 Design Bases

a. The CJW system is designed to circulate clean water to remove heat from the control air system compressor oil coolers and aftercoolers, C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 9.2-26 b. The CJW system provides clean water to remove heat from the control air system refrigerated dryer condensers, and

c. Heat is rejected from the CJW system to the plant service water system (TSW).

The CJW system piping and valves are designe d in accordance with th e ANSI B31.1 Power Piping Code, using design cond itions of 125 psig and 200

°F, except for the fill and makeup supply piping at the surge pipe, whic h is designed to 150 psig and 150

°F, the same as the demineralized water system. The CJW system is open to the atmo sphere at the surge pipe, and the maximum system pressure at the pump discharge is less than design pressure. The heat exchangers are designed and c onstructed to the ASME Boiler and Pressure Vessel (B&PV)

Code Section VIII for design c onditions of 150 psig and 200

°F.

9.2.8.2 System Description

The CJW system is show n diagrammatically in Figure 9.2-15. Fill and makeup water is supplied from the demineralized water system to the surge pipe through a float operated valve (CJW-LCV-1). An overflow line allows expansion volume to drain to an equipment drain.

The surge pipe is connected to the circulating co olant return header at a high point and is open to the atmosphere to vent free air released from the heated wa ter. A positive suction head of approximately 20 ft is provided at the circulating pump suction by the water level in the surge pipe.

Two full-capacity centrifugal pumps (CJW-P-1A and CJW-P-1B) are provided, permitting the system to remain in operation while one pump is isolated for servicing or repair. The operating pump draws water through the return h eader from the outlets of the compressor(s) and refrigerated dryers(s) in service at that tim

e. This heated water is discharged from the pump toward two full-capacity plate and frame heat exchangers.

The heat exchangers (CJW-HX-1A and CJW-HX

-1B) are designed for one to be in operation while the other is isolated. Heat is rejected from the CJW system to the plant service water system. The plant service wate r is essentially rive r water which may c ontain considerable amounts of particulate solids.

The CJW to TSW heat exchangers are of the plate and frame type that maintain water velocities high through the passes on both sides, and are readily opened to clean the plate surfaces without disturbing the connecting piping. In the event a heat exchanger becomes fouled, it may be isolated and cleaned while both the CJW and TSW systems remain in operation. The cooled wate r from the heat exchangers is piped to the compressor and refrigerated drye r supply header. Coolant flow to the compressors and dryers is controlled by temperature cont rol valves in each unit. The dryers will provide a minimal flow path so the CJW pumps do not deadhead because the dryers are designed for continuous operation.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-06-054 9.2-27 Cooling water at each compressor and refriger ated dryer may be manually isolated when a compressor or dryer is taken out of service for repair, and the c oolant may be drained, without disturbing coolant flow at the other compressors and dryers.

9.2.8.3 Safety Evaluation

Operation of the CJW system is not required for the initiation of any engineered safety feature system, or for safe shutdown of the reactor. However, plant availability depends on an adequate supply of compressed control air to operate nonnuclea r plant systems, and the CJW system includes redundant components to ensure operability of the system under most conditions.

9.2.8.4 Testing and In spection Requirements

Routine surveillance inspecti ons while in operation will discover conditions requiring correction before functional failure of the system will occur.

9.2.8.5 Instrumentation Requirements

Sufficient instruments and controls are provided for system operati on to support normal facility

requirements.

9.

2.9 REFERENCES

9.2-1 "1979 Ultimate Heat Sink Spray Syst em Test Results,"

Washington Public Power Supply System Nuclear Proj ect No. 2 Report (WPPSS-EN-81-01).

9.2-2 "WNP-2 Spray Pond Dr ift Loss Report," Washi ngton Public Power Supply System, July 1985.

9.2-3 "Plant Service Water (TSW) System

," Design Basis Document, Section 350.

9.2-4 "Standby Service Water (SW) System," Design Basis Document, Section 309.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 9.2-29 Table 9.2-1 Ultimate Heat Sink Spray Cooling

Pond Design Parameter Quantity Pond configuration Squ a re (250 ft x 250 ft)

Surface area (two pond s) 125,000 f t 2 Normal water elevation (above msl) 433 ft 6 in.

Maximum operating water elev a tion (above msl) 433 ft 9 in.

Overflow water elevation (above msl) 434 ft 6 in.

Pond bottom elevation (above msl) 420 ft 0 in.

Pump sump bottom elevat i on (above msl) 408 ft 3 in.

Freeboard above normal water level 1 ft 0 in.

Sedimentation allowance 6 in.

Normal pond capacity (two ponds at el. 433 ft 6 in.)

12,620,000 gal

Minimum pond capacity (two ponds at el. 432 ft 9 in.)

11,920,000 gal

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 9.2-30 Table 9.2-2 Spray System Design

Parameter Quantity Number of syste m s 2 Number of trees per sy s t em 32 Diameter of the ring headers 140 ft 0 in.

Circumferential spacing between trees 13 ft 9 in.

Number of horizontal arms per riser 7 Pressure at top nozzle 17.3 a psig Pressure at interface (flange) 24.5 a psig Flow rate per tree at design pressure 321.9 a gpm Total flow rate per system 10,30 0 a gpm Nozzle type Spraying S y stems Company 1-1/2-CX-27-55 Height of vertical riser 17 ft 9 in.

Spacing between arms 32 in. Height of bottom nozzle elevation from system interface (flange) riser pipe diameter

6-11/16 in.

Riser pipe 8 in., Schedule 80 Horizontal branch arm pipe 1.5 in., Schedule 80 a Flow rates and pressures are for final design with fuel pool cooling system upgraded (see Section 9.1.3). Prior to completion of fuel pool cool i ng upgrade, total flow is approximately 9750 gpm.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT Dece m ber 2003 9.2-31 Table 9.2-3 Diurnal Variation in Meteorological Dat a a Data based on July 10, 1975 Ho ur Dry Bulb ( F) Dewpoint ( F) Wet Bulb ( F) Wind Spee d b (mph) Solar Radiation (Btu/hr) Noon 100.91 59.41 72.98 6.60 276.56 1:00 p.m. 103.09 59.69 73.58 6.60 269.47 2:00 105.20 58.91 73.96 6.60 248.68 3:00 105.71 56.00 72.80 6.60 215.58 4:00 104.93 54.11 71.78 6.60 172.40 5:00 102.48 55.88 71.81 6.60 122.04 6:00 101.15 56.05 71.50 6.60 68.05 7:00 98.27 56.13 70.68 6.60 16.46 8:00 96.21 56.59 70.27 6.60 0.00 9:00 90.72 60.53 70.57 6.60 0.00 10:00 91.33 57.68 69.31 6.60 0.00 11:00 91.49 60.48 70.77 6.60 0.00 Midnight 90.91 58.03 69.35 6.60 0.00 1:00 a.m. 85.92 59.17 68.39 6.60 0.00 2:00 84.24 57.28 66.88 6.60 0.00 3:00 80.61 56.21 65.14 6.60 0.00 4:00 80.24 58.48 66.21 6.60 0.00 5:00 78.27 59.55 66.15 6.60 16.46 6:00 83.25 62.99 69.65 6.60 68.05 7:00 86.77 62.91 70.67 6.60 122.04 8:00 90.64 61.09 70.83 6.60 172.40 9:00 92.64 62.00 71.90 6.60 215.58 10:00 95.23 63.36 73.38 6.60 248.68 11:00 98.32 62.40 73.73 6.60 269.47 C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT Dece m ber 2003 9.2-32 Table 9.2-3 Diurnal Variation in Meteorological Dat a c (Continued)

Data ba s e d on average values for the period July 9-August 8, 1961 Ho ur Dry Bulb ( F) Dewpoint ( F) Wet Bulb ( F) Wind Spee d b (mph) Solar Radiation (Btu/hr) Noon 95.40 45.9 65.5 5.50 276.56 1:00 p.m. 96.80 46.1 66.0 5.50 269.47 2:00 97.30 46.1 66.2 5.50 248.68 3:00 96.80 46.2 66.0 5.50 215.58 4:00 95.40 46.2 65.5 5.50 172.40 5:00 93.10 46.0 64.7 5.50 122.04 6:00 90.10 45.6 63.6 5.50 68.05 7:00 86.60 45.6 62.3 5.50 16.46 8:00 82.80 45.6 61.0 5.50 0.00 9:00 79.00 45.2 59.6 5.50 0.00 10:00 75.60 45.6 58.4 5.50 0.00 11:00 72.50 46.0 57.3 5.50 0.00 Midnight 70.20 46.2 56.5 5.50 0.00 1:00 a.m. 68.80 46.0 56.0 5.50 0.00 2:00 68.30 46.3 55.8 5.50 0.00 3:00 68.80 46.1 56.0 5.50 0.00 4:00 70.20 46.2 56.5 5.50 0.00 5:00 72.50 45.8 57.3 5.50 16.46 6:00 75.60 46.0 58.4 5.50 68.05 7:00 79.00 46.6 59.6 5.50 122.04 8:00 82.80 45.8 61.0 5.50 172.40 9:00 86.60 45.6 62.3 5.50 215.58 10:00 90.10 45.8 63.6 5.50 248.68 11:00 93.10 45.8 64.7 5.50 269.47 C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT Dece m ber 2003 9.2-33 Table 9.2-3 Diurnal Variation in Meteorological Dat a d (Continued)

Data ba s e d on average values for the period July 2-August 1, 1960 Ho ur Dry Bulb ( F) Dewpoint ( F) Wet Bulb ( F) Wind Spee d b (mph) Solar Radiation (Btu/hr) Noon 96.40 42.50 64.70 6.91 276.56 1:00 p.m. 98.00 43.50 65.40 6.91 269.47 2:00 98.50 43.50 65.60 6.91 248.68 3:00 98.00 43.50 65.40 6.91 215.58 4:00 96.40 42.50 64.70 6.91 172.40 5:00 93.90 42.00 63.70 6.91 122.04 6:00 90.70 42.00 62.30 6.91 68.05 7:00 86.90 40.50 60.70 6.91 16.46 8:00 82.90 40.00 59.00 6.91 0.00 9:00 78.90 40.00 57.30 6.91 0.00 10:00 75.10 39.00 55.70 6.91 0.00 11:00 71.90 39.00 54.30 6.91 0.00 Midnight 69.40 39.00 53.30 6.91 0.00 1:00 a.m. 67.80 39.00 52.60 6.91 0.00 2:00 67.30 39.00 52.40 6.91 0.00 3:00 67.80 39.00 52.60 6.91 0.00 4:00 69.40 39.00 53.30 6.91 0.00 5:00 71.90 39.50 54.30 6.91 16.46 6:00 75.10 39.00 55.70 6.91 68.05 7:00 78.90 40.00 57.30 6.91 122.04 8:00 82.90 40.00 59.00 6.91 172.40 9:00 86.90 40.70 60.70 6.91 215.58 10:00 90.70 42.00 62.30 6.91 248.68 11:00 93.90 42.20 63.70 6.91 269.47 C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT Dece m ber 2003 9.2-34 Table 9.2-3 Diurnal Variation in Mete orological Data (Continued) a Worst single day of record used to analyze the pond thermal response during first 3 days following LOCA.

b Wind speed is the average wind speed for period.

c Day 4 through 30 used to analyze pond thermal response following LOCA.

d Day 1 through 30 used to an alyze mass loss following LOCA.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-08-032 9.2-35 Table 9.2-4

Flow Rates and Associated Heat Loads Used in the Ultimate Heat Sink Analysis a Equipment Cooled Flow (gpm) Heat Load (Btu/hr)

Division 1 1. Standby service wate r pump house cooler A (PRA-CC-1A) 80 300,0002. Diesel generator A 1,650 13,200,0003. Diesel generator building A coolers (DM CC-11&12) 144 Variable4. LPCS pump motor bearings 4 05. LPCS pump room cooler (RRA-CC-5) 56 400,0006. RHR seal cooler A 7 120,0007. RHR Room Cooler A (RRA-CC-2) 33 200,0008. Motor control center (MCC) dc room cooler (RRA-CC-12) 20 30,0009. MCC (RRA-CC-11) 15 54,00010. Control room chiller (CCH-CR-1A) 161 764,75011. Cable spreading room c ooler (WMA-CC-52A) 40 130,00012. Switchgear room cooler (WMA-CC-53A) 60 420,00013. Hydrogen recombiner MCC room cooler (RRA-CC-13) 11 40,00014. Hydrogen recombiner aftercooler A (DEACTIVATED) 0 015. RHR heat exchanger A 6,750 to 7,400 Variable16. Analyzer room cooler (RRA-CC-15) 10 15,00017. Fuel pool pump room cooler (RRA-CC-20) 35 112,000

18. Fuel pool heat exchanger A 575 Variable TOTAL 9,800 to 10,301 C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 Table 9.2-4 Flow Rates and Associated Heat Loads Used in the Ultimate Heat Sink Analysis a (Continued)

Equipment Cooled Flow (gpm) Heat Load (Btu/hr)

LDCN-08-032 9.2-36 Division 2

1. Standby service wate r pump house cooler B (PRA-CC-1B) 80 300,000 2. Diesel generator B 1,650 13,200,000 3. Diesel generator building B coolers (DMA-CC-21&22) 144 Variable 4. RHR pump seals B 7 120,000 5. RHR pump seals C 0 0 6. RHR room cooler B (RRA-CC-3) 33 162,000 7. RHR room cooler C (RRA-CC-1) 33 190,000 8. RCIC pump room cooler (RRA-CC-6) 12 97,000 9. MCC room cooler (RRA-CC-10) 15 55,000 10. Control room chiller (CCH-CR-1B) 161 764,750 11. Cable spreading room cool er (WMA-CC-52B1) 40 130,000 12. Switchgear room cooler (WMA-CC-53B1) 60 393,000 13. Aftercooler hydrogen recombiner B (DEACTIVATED) 0 0 14. Hydrogen recombiner MCC room cooler B (RRA-CC-14) 11 35,000 15. RHR B heat exchanger B 6,550 to 7,400 Variable 16. Analyzer room cooler (RRA-CC-17) 10 25,000 17. Fuel pool pump room cooler (RRA-CC-19) 35 100,000 18. Fuel pool heat exchanger B 575 Variable TOTAL 9,700 to 10,266

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 Table 9.2-4 Flow Rates and Associated Heat Loads Used in the Ultimate Heat Sink Analysis a (Continued)

Equipment Cooled Flow (gpm) Heat Load (Btu/hr)

LDCN-08-032 9.2-37 Division 3 b 1. HPCS diesel gene rator 850 7,890,000 2. HPCS diesel building coolers 137 Variable 3. HPCS pump room cooler 35 600,000 TOTAL 1022 a Information in this table is data input into the thermal analysis of the UHS calculation. The thermal analysis is constructed to provide the peak pond temp erature for a LOCA. As such, the data in this table was chosen to conservatively bound the conditions in the field and, therefore, does not necessarily represent exact plant conditions.

b The HPCS heat load is a straight dump into the UHS; therefore, the flow rates listed are not used in the UHS analysis and are for reference only.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-08-032 9.2-38 Table 9.2-5

Heat Load Rates Used in Ultimate Heat Sink Analysis I. Core decay heat load a Calculated using BTP ASB 9-2 data.

II. Reactor coolant sensible heat load a The energy (414 x 10 6 Btu referenced to 32 F) of the reactor coolant is accounted for by

starting the suppression pool at 150 F). Time (hr) Rate (10 6 Btu/hr) III. Reactor vessel, piping, and core sensible heat load a t < 24 t 24 8.14 Negligible IV. Metal-water reaction heat load a t < 1 t 1 0.47 Negligible V. ECCS pump work loada,b,c t < 8 t 8 13.168 5.534 VI. HPCS (division 3) service water system heat load c,d t < 8 t 8 8.623 0 VII. Constant division 1 service water system heat load e,f t > 0 19.7 VIII. Fuel pool heat load t > 0 8.2 g a Rejected initially to the suppression pool and subsequently transf erred by the RHR heat exchangers to the UHS.

b RHR pump 1.972 x 10 6 Btu/hr LPCS pump 3.5623 x 10 6 Btu/hr HPCS pump 7.6335 x 10 6 Btu/hr c HPCS system and HPCS SW system shut down after 8 hr. LPCS system and RHR loop A maintain long-term cooling.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-08-032 9.2-39 Table 9.2-5

Heat Load Rates Used in Ultimate Heat Sink Analysis (Continued)

d HPCS service water pump work 0.132314 x 10 6 Btu/hr HPCS coolers (Table 9.2-4) 8.49 x 10 6 Btu/hr Does not include HPCS diesel building coolers.

e Division 1 service water pump work 3.91853 x 10 6 Btu/hr Division 1 diesel generator, 15.784 x 10 6 Btu/hr Coolers and misc. equipment (Table 9.2-4) f Excludes fuel pool, RHR heat exchanger, a nd diesel generator building "A" coolers.

g This heat load is treated as a constant.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-08-032 9.2-40 Table 9.2-6

Spray Pond Water Losses and Content (30 days after desi gn basis LOCA event)

Analysis Quantity (gal)

Drift losses a 3.042 x 10 6 Spray evaporation b 5.523 x 10 6 Surface evaporation 0.649 x 10 6 Hydrogen recombiner (DEACTIVATED) 0 Leakage 0.1728 x 10 6 TOTAL 9.387 x 10 6 Remaining inventory 2.533 x 10 6 Minimum volume required for system operation 2.26 x 10 6 a Based on conservative RHR heat exchanger flow rate of 7400 gpm and fuel pool cooling at 575 gpm.

b Based on inclusion of design basi s fuel pool cooling heat load.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 9.2-41 Table 9.2-7 Source of Spray Pond Makeup Water Water Source Design Basis Earthquake Probable Maximum Flood River Blockage or Dr o ught Tornado LOCA Plant makeup water pumps N/A a N/A a N/A a A b A b Cooling tower basin by

gravity N/A a A b A b A b A b Tank truck or rail A b A b A b A b A b a Not available b Available

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 9.2-42 Table 9.2-8 Standby Se r v ice Water System

Failure Analysis Single Acti v e Failure Analysis Loss of power on one emergency bus due to

failure of a diesel gene rator to start or loss of offsite power.

The service water pump powered by the redundant emergency bus will be

automatically start e d to prov i de the necessary cooling water.

Failure of pump to automatically start. Same anal ysis as above.

Failure of an isolation valve in the service water piping to the cooling tower.

Redundant valve is oper a ted to perform the

isolation function.

Failure of ECCS pump room air cooling.

Essential plant cool i ng requirements met by the redundant ECCS subs ystems which have their own independently cooled pump

rooms. Failure of a single service water pump during operation.

Essential plant cooli ng requirements met by the remaining operable redundant ECCS

subsystems.

Failure of system return isolation valve. Essential plant coo ling requirements met by the remaining intact ECCS subsystem which

includes its own indepe ndent return service water headers. Failure of a pump discharge or RHR outlet

isolation valve to open.

Essential plant cooli ng requirements met by the remaining operabl e redundant emergency

core cooling subsystems.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 Table 9.2-8 Standby Service Water System Failure Analys is (Continued)

Single Active Failure Analysis 9.2-43 Failure of any supply or return piping.

Essential plant cooli ng requirements met by the remaining intact ECCS subsystem which includes its own independent supply and return service water headers. Passive failures, i.e., pipe wh ip, missiles, jet loads are discussed in Sections 3.5 and 3.6.1. The SW system is a moderate-energy fluid

system since it is ope rated at less than 275 psig and 200

°F. Failure of RHR heat exchanger. Esse ntial plant cooling requirements met by the remaining intact redundant RHR

subsystem which in cludes its own 100% capacity heat exchanger.

Fire All credible fires and their consequences were evaluated in Appendix F. Because of physical separation, redundancy, and low inventories of combustibles, there are no

credible fires which could impede the SW from performing its function.

Failures of equipment by missile.

Th e HVAC equipment in the SW pump house is arranged in such a manner that would preclude any postulated missile from preventing the plant to be brought to a safe

shutdown. External missiles are discussed in Section 3.5.

Table 9.2-9 Integrated Heat Data - U ltimate Heat Sink Reanalysis Peak UHS Temperature Case Time After LOCA (minutes)

Q Decay a Q Sens b Q Aux 1 c Q Aux 2 d Q Aux 3 e Q Total f Q SW g,h 10 7 Btu 1 1.141 0.014 0.022 0.045 0.015 1.237 0.158 2 2.050 0.027 0.044 0.090 0.031 2.242 0.317 4 3.605 0.054 0.088 0.180 0.062 3.989 0.641 10 7.367 0.136 0.219 0.450 0.155 8.327 1.641 20 12.538 0.271 0.439 0.900 0.310 14.458 3.384 40 20.805 0.543 0.878 1.800 0.621 24.647 7.082 90 36.910 1.220 1.975 4.052 1.399 45.556 17.169 120(2H) 45.085 1.630 2.634 5.404 1.868 56.621 23.616 240(4H) 73.582 3.260 5.267 10.809 3.744 96.662 51.190 360(6H) 98.348 4.880 7.901 16.180 5.593 132.902 80.381 480(8H) 120.603 6.510 10.534 21.514 7.408 166.569 110.210 720(12H) 160.016 9.770 12.748 32.106 7.408 222.048 169.181 960(16H) 195.082 13.02 14.962 43.623 7.408 273.095 225.503 1200(20H) 227.462 16.28 17.175 53.168 7.408 321.493 278.998 C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORTDecember 2011LDCN-08-032 9.2-44 Table 9.2-9 Integrated Heat Data - Ultimate Heat Sink Reanalys is (Continued) Peak UHS Temperature Case (Continued)

Time After LOCA (minutes)

Q Decay a Q Sens b Q Aux 1 c Q Aux 2 d Q Aux 3 e Q Total f Q SW g,h 10 7 Btu C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORTDecember 2011LDCN-08-032 9.2-451440(1D) 258.019 19.54 19.389 62.867 7.408 368.223 329.810 2160(1.5D) 342.440 19.56 26.030 96.003 7.408 491.441 465.612 2880(2D) 419.033 19.56 32.671 128.199 7.408 606.871 588.078 4320(3D) 555.030 19.56 45.954 193.654 7.408 821.606 811.164 5760(4D) 674.009 19.56 59.236 259.407 7.408 1019.620 1017.191 7200(5D) 780.800 19.56 72.518 325.863 7.408 1206.149 1209.436 8640(6D) 878.638 19.56 85.801 391.122 7.408 1382.529 1387.436 11520(8D) 1055.541 19.56 112.365 522.633 7.408 1717.507 1726.759 14400(10D) 1215.309 19.56 138.930 654.180 7.408 2035.387 2047.600 17280(12D) 1363.194 19.56 165.495 785.644 7.408 2341.301 2355.179 23040(16D) 1633.357 19.56 218.624 1048.117 7.408 2927.066 2943.321 28800(20D) 1877.638 19.56 271.753 1310.437 7.408 3486.796 3504.390 34560(24D) 2101.496 19.56 324.882 1572.698 7.408 4026.044 4044.572 43200(30D) 2406.351 19.56 404.576 1967.206 7.408 4805.101 4827.357 Table 9.2-9 Integrated Heat Data - Ultimate Heat Sink Reanalys is (Continued) Mass Loss Case Time After LOCA (minutes)

Q Decay a Q Sens b Q Aux 1 c Q Aux 2 d Q Aux 3 e Q Total f Q SW g,h 10 7 Btu 1 1.141 0.014 0.022 0.045 0.015 1.237 0.241 2 2.050 0.027 0.044 0.089 0.030 2.240 0.484 4 3.605 0.054 0.088 0.179 0.061 3.987 0.979 10 7.367 0.136 0.219 0.447 0.152 8.321 2.505 20 12.538 0.271 0.439 0.894 0.305 14.447 5.156 40 20.805 0.543 0.878 1.788 0.610 24.624 10.728 90 36.910 1.220 1.975 4.024 1.375 45.504 25.524 120(2H) 45.085 1.630 2.634 5.366 1.833 56.548 34.689 240(4H) 73.582 3.260 5.267 10.716 3.658 96.483 71.755 360(6H) 98.348 4.880 7.900 16.025 5.449 132.602 108.096 480(8H) 120.603 6.510 10.534 21.281 7.189 166.117 142.981 720(12H) 160.016 9.770 12.748 31.674 7.189 221.397 207.088 960(16H) 195.082 13.02 14.962 42.054 7.189 272.307 264.650 1200(20H) 227.462 16.28 17.175 52.640 7.189 320.746 317.798 C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORTDecember 2011LDCN-08-032 9.2-46 Table 9.2-9 Integrated Heat Data - Ultimate Heat Sink Reanalys is (Continued)

Mass Loss Case (Continued)

Time After LOCA (minutes)

Q Decay a Q Sens b Q Aux 1 c Q Aux 2 d Q Aux 3 e Q Total f Q SW g,h 10 7 Btu C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORTDecember 2011LDCN-08-032 9.2-471440(1D) 258.019 19.54 19.389 63.466 7.189 367.603 367.578 2160(1.5D) 342.440 19.56 26.030 95.849 7.189 491.068 498.703 2880(2D) 419.033 19.56 32.671 128.533 7.189 606.968 619.615 4320(3D) 555.030 19.56 45.954 194.785 7.189 822.518 840.885 5760(4D) 674.009 19.56 59.236 259.697 7.189 1019.691 1036.572 7200(5D) 780.800 19.56 72.518 325.625 7.189 1205.692 1226.242 8640(6D) 878.638 19.56 85.801 391.033 7.189 1382.221 1404.952 11520(8D) 1055.541 19.56 112.370 522.481 7.189 1717.141 1741.809 14400(10D) 1215.309 19.56 138.93 653.818 7.189 2034.806 2060.639 17280(12D) 1363.194 19.56 165.50 785.138 7.189 2340.581 2367.082 23040(16D) 1633.357 19.56 218.62 1047.679 7.189 2926.405 2953.354 28800(20D) 1877.638 19.56 271.75 1310.168 7.189 3486.305 3513.470 34560(24D) 2101.496 19.56 324.88 1573.252 7.189 4026.377 4056.154 43200(30D) 2406.351 19.56 404.59 1967.899 7.189 4805.589 4836.358 Table 9.2-9 Integrated Heat Data - Ultimate Heat Sink Reanalys is (Continued)

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORTDecember 2011LDCN-08-032 9.2-48 a Q Decay Integrated core decay heat suppression pool.

b Q Sensible Integrated sensible he at rejected by the reactor vessel, piping, and core to the suppression pool.

c Q Auxiliary I Integrated h eat from ECCS pump work rejected to the suppression pool.

d Q Auxiliary 2 Integrated heat from auxiliary systems rejected to division 1 SW system. This heat includes all sources of heat into division 1 SW system except fo r the RHR heat exchanger. The RHR heat exchanger transfers heat from the suppression pool to division 1 SW system.

e Q Auxiliary 3 Integrated heat from HPCS SW system. This heat is a straight heat dump into spray pond A.

f Q Total Sum of Q Decay, Q Sensible, Q Auxilia ry 1, Q Auxiliary 2, and Q Auxiliary 3.

g Q Service Sum of Q Auxiliary 2 and the h eat rejected by the RHR heat exchanger in to division 1 SW system, i.e, the sum of the heat rejected through the spray nozzles.

h The RHR heat exchangers provide suppression pool cooling from 10 minutes through to the end of an accident (30 days). No heat exchanger cooling is assumed for the first 10 minutes of an accident. See Sections 6.2.2.2 and 6.3.2.8 for further information on containment cooling.

Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.2-1.1 124 M508-1Plant Service Water SystemRev.FigureDraw. No.Amendment 61December 2011 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.2-1.2 22 M508-2Plant Service Water SystemRev.FigureDraw. No.Amendment 61December 2011 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.2-2.1 83 M525-1Closed Cooling Water System - Reactor and Radwaste BuildingsRev.FigureDraw. No.Amendment 61December 2011

Amendment 60December 2009 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 09.2-03 76 M516Makeup Water Treatment SystemRev.FigureDraw. No.

Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.2-4 102 M517Demineralized Water SystemRev.FigureDraw. No.Amendment 61December 2011 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.2-5.1 75 P541-1Potable Hot and Cold WaterReactor, Turbine Generator & Radwaste Bldgs.Rev.FigureDraw. No.Amendment 61December 2011 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.2-5.2 7 P541-2Potable Hot and Cold Water Service Building and YardRev.FigureDraw. No.Amendment 61December 2011 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.2-5.3 7 P541-3Potable Hot and Cold Water Site BuildingsRev.FigureDraw. No.Amendment 61December 2011POTABLE WATER SUPPLY TO SITE BUILDINGS

Amendment 61December 2011 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.2-7 14 M782 Composite Piping Plan, Section and Details, Spray PondsRev.FigureDraw. No.

Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.960690.76Drift Loss Test Results 9.2-8012345678910111213141516171819202122 Wind Speed, mph Curve Fit Drift Loss, Percent of Flow Legend No Heat Load With Heat Load Safety Analysis Drift Loss 1 2 3 4 5 6

7 0 Columbia Generating StationFinal Safety Analysis Report 0 80 84Days Degrees (°F) 88 86 78 82 90 89.523456789101112131415161718192021222324252627282930 1 Figure Form No. 960690Draw. No.Rev.Temperature Response Following DesignBasis LOCA (Standby Service Water) 960222.21 9.2-9 Columbia Generating StationFinal Safety Analysis Report LDCN-FSAR-08-032 Amendment 61December 2011 Amendment 61December 2011 Figure Form No. 960690Draw. No.Rev.Water Inventory in UHS Following Design Basis LOCA 960222.22 9.2-10 2,000,000Time Inventory (Gallons)(Days)023456789101112131415161718192021222324252627282930 1 4,000,000 6,000,000 8,000,000 10,000,000 12,000,000 0 2,533,000 Columbia Generating StationFinal Safety Analysis Report LDCN-FSAR-08-032 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report9.2-11.1 101 M527-1 Condensate Supply SystemRev.FigureDraw. No.Amendment 61December 2011

Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.2-12.1116 M524-1Standby Service Water SystemRev.FigureDraw. No.Amendment 61December 2011 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.2-12.2 106 M524-2Standby Service Water SystemRev.FigureDraw. No.Amendment 61December 2011 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.2-12.3 18 M524-3Standby Service Water SystemRev.FigureDraw. No.Amendment 61December 2011 Figure Not Available For Public Viewing Figure Not Available For Public Viewing

Draw. No.

Rev.Figure C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-07-049 9.3-1 9.3 PROCESS AUXILIARIES

9.3.1 COMPRESSED

AIR SYSTEMS

The compressed air systems consist of the control and service air systems (CAS and SA) and the containment instrument air (CIA) system.

9.3.1.1 Design Bases 9.3.1.1.1 Control and Service Air Systems

The CAS and the SA system are supplied with oil-free compressed air from three rotary screw compressors, two refrigerated dryers, two sets of filters, and three air receivers located in the turbine generator building and by one rotary screw compressor and refrigerated filtered dryer in the radwaste building. The CA S receivers are maintained at a pressure between 110 psig and 120 psig by the compression load ing control devices. Norma lly, the SA compressor and one CAS compressor will meet the plant demand for compressed air. Should the air pressure in the three CAS air receivers, which are interconnected, drop to less than the running CAS compressor setpoint, a second and if necessary, a third CAS compressor will st art. If pressure in the CAS supply header drops to less than 80 psig, the SA system will be automatically isolated from the CAS air receivers to conserve air for use in the CAS.

The CAS provides instrument quality air, oil-free, maximum particle si ze of 1 micron, and dried to an atmospheric dewpoint of -40°F, throughout the plant for pneumatic instrumentation, controls and actuators. Air is drawn from the air receivers, is passed through prefilters, a desiccant dryer and afterfilters, to distribution piping.

A complete loss of control air will result in a plant shutdown, but none of the reactor systems require control air to achieve and maintain a safe shutdown condition.

The SA system provides oil-free compressed air from the CAS air receivers and/or from the SA receiver in the radwaste building, for use in process functions such as backwashing filters or demineralizers, for genera l services and maintenance uses throughout the plant, and as a source of breathing air.

The CAS and SA system are generally designed to Safety Class G, Seismic Category II requirements; however, piping and equipment in the reactor building, diesel generator building, and the standby service water (SW) pump houses are suppor ted to Seismic Category I requirements. Pressure vessels are designed and cons tructed in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code Section VIII. Piping and valves are desi gned and constructed in accordance with the ANSI B31.1 Power Piping Code, except for piping or valves performing a primary containment function and the local cont rol piping at the outer main steam isolation valves (MSIV). These

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 9.3-2 portions of the piping are Quality Class I, Seismic Category I, and are designed and constructed in accordance with ASME B&PV Code Section III, Class 2 and Class 3, respectively.

9.3.1.1.2 Containment In strument Air System

The CIA system is designed to deliver clean, dr y, compressed gas, nitr ogen or air, to the following valve actuator accumulato rs and valve pilot controls inside primary containment. Seven individual accumulators for the seven main steam relief valves (MSRV) dedicated to the automatic depressurization system (ADS) function are supplied with gas through two separate isolable headers at a nominal pressure of 180 psig. The four inboard MSIV actuator accumulators, 18 MSRV actuator accumulators for the power-assisted pressure relief function, and two reactor recirculation cooling (RRC) pump seal staging dr ain valve pilot control valves are supplied with gas at a reduced nominal pressu re of 100 psig. The CI A system also provides nitrogen to the set pressure verification device (SPVD) system. For normal plant operation, nitrogen is supplied from the containment nitrog en (CN) system cryogeni c storage vessel which is also the source of nitrogen for inerting the primary containment atmosphere. Should the normal nitrogen supply become unavailable, the gas supply piping to the ADS function accumulators will automa tically isolate from the cryogenic nitrogen supply, and the ADS accumulator backup compressed gas manifold subs ystems will provide a nominal pressure of 180 psig nitrogen from banks of high pressure compressed n itrogen cylinders. The designed pneumatic supply to the ADS accumulator is such that, following a failure of the safety-related pneumatic supply to the accumulator, at least two valve actuations can occur with the drywell at 70% of design pressure.

The 100 psig line does not have an automatic backup supply, but provision has been made for supplying air from the CAS system through the CIA dryer and filters via manually actuated intertie valves in lieu of the normal CN supply.

The piping for normal supply of compressed nitrogen or air, that is outboard of the primary

containment isolation valves, is designed to the ANSI B31.1 Po wer Piping Code, and pressure vessels are constructed to the rules of the ASME B&PV Code Section VIII. Piping penetrating primary containment and inside of primary c ontainment, from the outermost containment isolation valves to the point of use valve connect ions, is constructed to Safety Class 2, ASME B&PV Code Section III, Class 2, Seismic Category I, requirements; except that the flexible connectors at the four inboard MSIV actuator suppl y valves are equivalent to but are not ASME Section III stamped items. Piping located in the reactor building and the diesel generator building is supported by Seismic Category I supports, although most of the non-ADS piping itself is designed to Seismic Category II loadings. The ADS function accumulator backup compressed nitrogen supply subsys tems, from the normal supply is olation valves and from the outboard containment isolation va lves, to, but not including, the compressed nitrogen cylinders, pressure regulators and overpressure relief valves, are cons tructed to ASME B&PV Code Section III Class 3 rules includi ng Seismic Category I load cond itions. An exception to this

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 9.3-3 ASME statement occurs at the pressure control valve stations in the di esel generator building corridor, where the sections of line containing th e pressure control valve in each station is non-ASME (due to the unavailability of ASME pressure control va lves), but is isolable and bypassable via ASME manually operable valves. The co mpressed gas cylinders are U.S. Department of Transpor tation approved commercial shipping containe rs. The cylinder valves, connectors, pressure regulators, and overpressure re lief valves are approved by the Compressed Gas Association. The supports for the compressed gas cylinders, cylinder valves, connectors, pressure regulators, and overpressure relief valves are designed to Seismic Category I requirements.

9.3.1.2 System Description

9.3.1.2.1 Control and Service Air System

The CAS a n d SA system are shown schematical l y in Figure 9.3-1. Three compress o rs and three interconnected CAS air receivers in addition to one SA compressor and SA air receiver supply both control a nd SA requirements.

Each CAS compressor has a run-st andby-stop local selector switch. When the selector switch is set in the run position, the compressor runs conti nuously and loads and unloads to maintain receiver pressure. When the selector sw itch is in standby pos ition, the standby CAS compressor will automatically st art when the CAS supply header pressure falls below the running CAS compressor setpoint. Normally one CAS and the SA compressor are running with at least one CAS compressor on standby.

The CAS compressors take suct ion from the room through 10 filters. The air is filtered down to 0.1 before entering the air receivers. The CA S compressors are cool ed by the compressor jacket water (CJW) system. The CJW system is a closed loop syst em that transfers heat from the CAS system to the plant SW system. Air quality is mon itored by three ca rbon monoxide (CO) monitors.

The SA compressor takes suction from the outside to ensure that the air, which can be used for breathing purposes, is free from c ontaminants. A CO monitor alar ms if CO leve ls rise above safe levels.

The CAS and SA compressors pump an air/oil mixt ure to an internal air receiver/oil separator and then the air enters an internal aftercooler for cooling and moisture separation. The air is then discharged to a refrigerat ed dryer for further moisture removal. The air then enters a series of filters to remove any remaining particles and oil vapor from the air before entering the main air receivers.

Control air for station instrume ntation and controls is processed through a refrigerated air dryer, a filter array, one of two 100%-capacity air dryer sk ids. Each skid has two 100%

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 LDC N-9 8-1 1 7 9.3-4 capacity prefilters and after filters arranged in parallel pairs to allow for replacement of the filter cartridge without interruption of air flow. The dryer on each skid is a heatless regenerative twin tower desiccant type which dries through on e tower while the other is regenerating. A timer switches between towers and a purge economizer allows reduction in purge usage when the dryer is not operated at full capacity. There are two parallel refrigerated dryer and filter arrays, one or both of which may be operating at a given time.

Service air is distributed from the header to quick-disconnect hose connections where it is used for pneumatic service equipment and maintenance throughout the plant. Service air is also distributed for plant services such as demineralizer resin mixing and filter and demineralizer backwashing (see Figure 9.3-1

).

Table 9.3-1 presents the major characteristics of the air compresso rs, receivers, and the air dryer for the control and SA.

9.3.1.2.2 Containment In strument Air System

The CIA system i s primarily a pres s u rized nitrogen system as shown in Figure 9.3-2. The location of components in this system in the Seismic Category I reactor building is shown in Figures 1.2-7 , 1.2-8 and 1.2-9. Du r i ng normal reactor operation, the CN system, d i scussed in Section 6.2.5.7, supplies pressurized nitrogen from an 11,000-gal (approximately 1 million-scf) cryogenic storage tank to meet the requirement s of the following valves inside primary containment:

a. Full supply pressure (180 psig nominal) loads,

1. Seven ADS function MSRVs and their individual ADS accumulators, b. Reduced pressure (100 psig nominal) loads,

1. Four inboard MSIV accumulators,

2. Eighteen main steam safety/relief valves (SRVs) and their individual accumulators to be used for the power assisted pressure relief function, and

3. Two RRC pump seal staging drain valve pilot control valves.

For a discussion of the function and operation of the ADS, see Sections 6.3 and 7.3. In case the cryogenic nitrogen source does not maintain supply pressure to the ADS accumulator supply headers, two backup nitrogen cylinder bank subsystems are provided to automatically supply a nominal pressure of 180 psig nitrogen. A bank of 15 nitrogen cylinders supplies three of the ADS function accumulators, and a separa te bank of 19 nitrogen cylinders supplies the

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 LDC N-0 2-0 3 6 9.3-5 other four ADS function accumulators (see Figur e 9.3-2). Of these backup subsystems, only 14 nitrogen bottles for three ADS accumulators and 17 nitroge n bottles for the four ADS accumulators are necessary to provide a 30-day supply of nitrogen for the ADS function during a postulated loss-of-coolant acci dent (LOCA) condition. The backup subsystems also provide for manual SRV operation during station blackout (Appendix 8A

). A remote nitrogen cylinder connection is provided to each subsystem to permit supplementing the cylinder banks through manual connection of portable nitrogen cylinders, and thus main tain the ADS function for at least 100 days following a postulated LOCA even

t. The remote cyli nder connections are located in the diesel generator building corridor, adjacent to the reactor building, outside the secondary containment bounda ry, permitting personnel access to the connections under postaccident conditions.

Remote pressure control stations, each consisting of a bypassable pressure contro l valve, are schematically located downstream of the junction of the backup nitrogen cylinders supply with the remote backup nitrogen cyli nder supply, and are physically lo cated in the diesel generator building corridor. Thus, any problems with the pressure control valve could be accommodated by isolating it, to allow maintenance to be performed, while th e ADS function header pressure was being maintained with the manual bypass valve. The pressu re control valve allows the cylinder regulators to be set at a higher, but broad range, pressure.

The backup nitrogen cylinder banks are located in the reactor building vehicle air lock (railroad bay) and are accessible during normal reactor operation. Cylinders are valved to the supply piping in a sequential manner by a pre ssure-controlled programm er for each bank, so that only the number of cylinders necessary to maintain pressure in the ADS accumulator supply lines are drawn on. During normal reactor operation, minimum cylinder pressure is maintained as required by the Technical Specifications. The ga s cylinders used may only be charged to 3000 psig, a limitati on imposed by the system piping.

Once opened, the ADS valves are not expected to be cycled during the postaccident period; nevertheless, the air s upply was conservatively sized to allo w for extra cycles , since they may be used for alternate shutdown cooling.

In the event the cryogeni c nitrogen source totally fails to supply the system requirements, the backup nitrogen cylinder banks will supply their respective ADS supply headers as described above and the headers will be automatically isolated from the common CIA supply line. The reduced pressure loads will then be without a source of pr essurized gas until the CAS intertie valves are opened. A normally open vent between the two in-lin e series block valves could then be closed (the vent is there to guaran tee no backleakage from CIA to CAS, which could lower the oxygen content of CAS/

SA and result in an adverse condition should a breathing-air tap be in use). The relatively low pressure CAS supply would still be inadequate for the ADS headers, though, and they would still auto matically be supplied as discussed above.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-05-002 9.3-6 9.3.1.3 Safety Evaluation

9.3.1.3.1 Control and Service Air System

Operation of the CAS and SA systems are not required for the initiation of any engineered safety feature systems or for safe shutdown of the reactor.

A complete loss of the CAS system during power operation resu lts in a plant shutdown with the following expected to occur: The SA system is automatically isolated as the CAS system pressure decreases. Control rods drift into the core as the control rod drive (CRD) scram outlet valves begin to slowly open. The outboard MSIVs close causing an automatic reactor scram. Operator actions based on CAS failure would be taken in accordance with plant procedures to scram the rector, control reactor pressure vessel (RPV) pressure/water level, and to mitigate the effects of individual air-ope rated valve failure modes.

9.3.1.3.2 Containment In strument Air System

Since each of the two backup nitrogen cylinder banks and th e cryogenic nitrogen supply are independent of each other, a si ngle component failure in one will not affect the operational function of the other. The two ADS header tie line isolation valves are each powered from a different division of the critical power supply.

During normal operation, the cryogenic nitrogen supply will main tain pressure in the inboard MSIV, the power-assisted MSRV , and the ADS function accumulators. The cryogenic nitrogen supply piping and the CAS suppl y piping are not assumed to be serviceable under accident conditions. In such an event, the local accumulators at the MSIVs and MSRVs provide a short-term source of pressure for actuating these valves. Further discussion of the effects of loss of pressure to the MSIVs and the normal function MSRVs is presented in Sections 5.2.2.4 , 5.4.5 , and 7.3.1.1.10. The backup nitrogen cylinder bank subsystems w ill supply operating nitrogen pressure to the ADS MSRV accumulators at any ti me the normal supply does not function.

The solenoid valves on the ADS function MSRVs c ould be pressurized to a pressure greater than their qualified pressure under certain accident conditions if the ADS header pressure were allowed to approach the piping design pressure; therefore, the ASME relief valves are set at a lower pressure to preclude this possibility. They are located in the diesel generator building to provide access (post-LOCA) shoul d one open and stick open.

9.3.1.4 Testing and In spection Requirements

A channel functional test of the ADS accumulator backup compressed gas system alarm is conducted no less than once every 115 days, (whe n ADS is required operable) with a channel

calibration of the alarm performe d no less than once every 22.5 months that verifies an alarm C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 9.3-7 setpoint of greater than or e qual to 135 psig (decreasing). At a frequency no less than once per 22.5 months, the nitrogen capacity in at least two accumulator bottles per division within the backup compressed gas system is verified. The backup compre ssed gas pressure instrumentation is periodically tested and calib rated as required by the Technical Specifications.

Instrument air sampling and testing is performed periodically to ensure that the air quality meets system design requirements. Routine maintenance and inspec tion along with operator observation during operation ensures the system is functioning properly to support normal plant operations.

9.3.1.5 Instrumentation Requirements

9.3.1.5.1 Control and Service Air Systems

Instrumentation and controls are provided in the main control room to control and monitor the operation of the systems. Low receiver air pressure, high air temperature, high filter differential pressure, and other system operating parameters (see Figure 9.3-1) are alarmed in

the control room.

An air-operated control valve isol ates the SA when its header pr essure drops to 80 psig to enable sufficient air flow to the CAS. Nonlubricated control valves, a solid state timer, and a purge economizer regulate the air drying cycle.

9.3.1.5.2 Containment In strument Air System

The instrumentation provided in the CIA system is shown in F i gure 9.3-2. Pressure switches

and alarms are provided in this system to regulate the nitrogen supply from the 15 nitrogen bottles in one bank and the 19 bottles in the seco nd bank. Pressure switc hes in the tie lines between the CIA distribution header, which is it self supplied by either the CN system or the CAS system, and the ADS function lines monitor the supply pressure and cause tie line isolation valves to independently close if the tie line pressu re should drop below the switch setpoint for a time period greater than 3 minutes. A pressure switch on each ADS function header then causes the nitrogen bottles to begin supplying their respective headers after the tie-line isolation valves actua lly close. A redundant pressure switch on each ADS function header set at a slightly lower pressure than th e tie-line isolation setpoi nt, acts as a backup to either the isolation valve closed position switch, or the other pressure switch on the same ADS function header, to initiate supply from the n itrogen bottles. The re dundant pressure switch also initiates an alarm in the control room to alert the operators to a low-pressure condition in the ADS function header.

In addition to the low pressure switches, each ADS function header has a pressure switch that provides control room annuncia tion to alert the operators to a high-pressure condition.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 9.3-8 Local pressure indication is provided just downs tream of each pressure control valve station should the station require isola tion and the ADS function header pressure require regulation via the manual bypass valve.

9.3.2 PROCESS

SAMPLING SYSTEM

9.3.2.1 Design Bases

The plant process sampling system is designed to provide representative samples, under controlled conditions, of plant process streams. Provisions for continuous monitoring of selected systems provide a means of analytical surveillance of system trends and performance during plant operations. Laborator y samples are taken to provide (a) comprehensive analytical information on plant operations, (b) a check on continuous monitoring instrumentation, and (c) regular reports on critical plant systems to ensure safe and proper operation.

The sample system is designe d to the following criteria:

a. To ensure that the samples of the process stream are re presentative of the conditions that exist at the sample tap, the following design practices have been implemented:

1. The shortest practical line length and smallest practical line diameter is used to reduce time lag and sample line plate-out;

2. Sample lines are routed to prevent dead legs and are assembled to avoid traps and dips;

3. Certain sample lines have a continuous purge flow of sufficient velocity to inhibit deposition of suspended solids and to satisfy the requirements of analytical instrumentation;

4. To ensure representa tive sampling of gaseous effluents, special sampling probes are employed which consider the process line size, flow conditions, and flow direction. Where appropriate, isokinetic probes are used to ensure representative sampling of the particulates;

5. Where precision measurements are to be made of high purity (low conductivity) water, the process sample flow stream is conditioned by constant temperature baths before in-line conductivity measurement to

maintain the internal temperature co mpensation within the valid range. Hot process fluids are cooled by heat exchangers before entering grab sample hoods or analysis equipment to avoid measurement errors and the potential for injury; and

C OLUMBIA G ENERATING S TATION Amendment 56 F INAL S AFETY A NALYSIS R EPORT December 2001 LDC N-0 0-0 8 7, 0 1-008 9.3-9

6. Sample line pressures are reduced at sample racks to protect in-line instruments and minimize the potential for personnel injury. Pressure regulators with relief valves are used to control and limit sample pressure.

7. Where tubing samples are taken fr om high temperature and pressure lines, flow will be isolated and the temperature allo wed to decay to ambient before the system is depressurized, drained and then the sample removed. b. In addition to limiting sample temperat ures and pressures, the following safety precautions are taken:

1. The sample lines are made of continuous seamless tubing with a minimum of joints or connections, and

2. Process sample lines containing high ly radioactive fluids are routed to minimize personal radioactive exposure.

9.3.2.2 System Description

The sampling system consists of continuous fl ow in-line analytical in struments, plus numerous grab sample points which are eith er local or routed to a centr alized sampling station. The sampling system is shown in Figu r e s 9.3-3 , 9.3-4 , 9.3-5 and 9.3-17.

A sample station typically consis ts of a sample rack and a fume hood. Process fluids enter the conditioning rack and, where nece ssary, pass through one or more heat exchangers to reduce temperatures to 105°F or less. Pressure is reduced, where necessary, to limit sample pressure to 40 psig or less. If sampling criteria does not allow the temperature and pressure to be reduced at a sample rack, the flow will be isolated and the system brought to a safe temperature and pressure befo re sample removal to minimize the potential for injury to personnel. After conditi oning, the process fluid is routed to the chemical fume hood where grab samples are taken. In th e case of continuously monitore d process streams, a portion of the fluid is diverted through a constant temperature bath, through conductivity cells and rotameters, and then to a drai n line. Selected process flui ds are continuously monitored by sample line sensors fo r conductivity and other pa rameters. Electric out puts of the sensors go to monitors and continuous recorders. Electrodes located in RWCU piping and inside of the RPV provide electro-chemical pot ential (ECP) signals to continuous recorders on PSR-CAB-1.

Liquid effluents from the sample racks and chemical fume hoods are routed through closed drains to the liquid radwaste system.

C OLUMBIA G ENERATING S TATION Amendment 56 F INAL S AFETY A NALYSIS R EPORT December 2001 9.3-10 The systems that require interm ittent or continuous an alytical sampling are arranged to permit sampling during normal and shutdown conditions. Many of the sample lines flow continuously. For those that do not, written procedures specify the flow rate and time required for sample line purge before taking the sample.

This ensures that representative samples are obtained.

9.3.2.2.1 Sample Locations

The available process liquid sampli ng system points are described in Table 9.3-2 (not all available sample points are used). The samplin g system for radiation monitoring equipment is described in Section 11.5.

9.3.2.2.2 Liquid Samp le Taps and Probes

Sample line connections are, wher e possible, located after a strai ght run of pipe. Since process piping is designed for turbulent flow conditions, sample taps for nominal line sizes 3 in. and less can be taken from tees or welded-pipe connections. For nominal pipe sizes 4 in. and larger, sample probes are installe d into the pipe line. Sample connections to process pipes are a minimum of 0.75 in. and have the root valves as close to process lines as possible. The root valve and connection are the same quality and code group as that of the process pipe.

The samp l e probes are fabrica t ed in accordance with Figures 9.3

-6 , 9.3-7 , and 9.3-8 and are designed to avoid failure through vibration or wear.

Sampling taps, where possible, are located at the sides of horizontal process lines rather than at the top or bottom.

9.3.2.2.3 Steam Samples

Two sample taps for main steam are installed, one of which is a spare.

The sample line is stainless steel tubing of sufficient total length (from sample point to sample rack) to allow for decay of 16N to minimize radiation dose rate at the sample station. The steam sample is condensed near the process line tap and the resultant c ondensate is piped to the sample rack.

9.3.2.2.4 Sample Piping Design

The sample piping, from the root valve at the pr ocess tap to the sample conditioning rack in a sample room, is small diamet er heavy wall stainless steel tubing to minimize corrosion or contamination of the sample. The small diameter also provides a passive flow restriction in the event of a sample line break.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 9.3-11 Sample lines are sized to ensure turbulent flow conditions at the required flow rate. This turbulence alleviates stratifi cation or plate-out of par ticles in the sample line.

Continuous flow lines containing fluids over 150°F are insulated to protect personnel. When warranted by high anticipated radi ation levels, sample lines containing radioactive fluids are routed or shielded as described in Section 12.3.

9.3.2.2.5 Fume Hood De sign and Grab Samples All liquid sample lines entering the sample rooms have provisions for grab sampling. Each fume hood is an enclosed sheet metal structure (sta inless steel) with manua l grab sample valves positioned over a stainless steel sink, all located w ithin the hood. The face of the hood has a transparent sash opening for hood access. Ambient air enters the opening to remove any gases or droplets released from the liquid sample. Air is exhausted from the top of the hood into the building ventilation exhaust system. The hood sink collects the samp le liquids not collected in the grab sample and pipes the effluent to a clos ed drain system. A demi neralized water faucet and a compressed air nozzle is located near the si nk in the hood to flush and clean sampling apparatus.

9.3.2.3 Safety Evaluation

The process sampling system has no direct pr ocess control functions. Each continuously monitored sample line is provided with local in dication and annunciation.

This includes a high alarm, a low alarm where appropriate, and a c ontinuous recorder. Remote indication is provided on certain samples.

Components of the process samplin g system which form part of the reactor coolant pressure boundary (RCPB) or containment isolation system are designed in accordance with Seismic Category I requirements and other code and quality requirements as described in Section

3.2. Sample

lines which form part of the contai nment isolation are provided with automatic fail-closed isolation valves bot h inside and outside of contai nment. The components of the process sampling system that do not form part of the RCPB or containment isolation system and are not in the reactor building are not de signed to Seismic Category I requirements since they are not necessary to en sure any of the following:

a. The integrity of the RCPB,

b. The capability to shut down the reactor, and

c. The capability to prevent or mitigate the consequence of ac cidents which could result in potential offsite exposure in excess of the limits of 10 CFR 50.67.

C OLUMBIA G ENERATING S TATION Amendment 56 F INAL S AFETY A NALYSIS R EPORT December 2001 LDCN-00-087 9.3-12 The operation of the liquid sampling system is not necessary for plant safety. Therefore, in the unlikely event of an accident, all sample lines which pass through the containment are automatically isolated by fail-closed valves. Electrical power for the sampling subsystem is from a nonessential bus.

The routing of high temperature and high pressure sample lines outside the containment is not considered hazardous because of limited flow and limited stored en ergy as discussed in Section 3.6. The location of sample temperature a nd pressure reduction equipment outside the containment allows maintena nce during plant operation.

The sample station is a closed system with grab samples taken under the safety of a chemical fume hood to minimize radiological hazard. The hood maintains a constant air velocity of approximately 100 ft/minute through the working face to ensure that airborne contamination will not enter the room.

Pressure control valves are installed in high-pressure sample lines so that operators will not be exposed to high sample pressure when obtaining grab samples.

At PSR-SR-49 sample rack, tubi ng samples are taken from high temperature and pressure lines that will utilize double isolation valves to protect the worker from harm due to the operating conditions of the high-energy line. This sample rack will be isolated th en cooled to ambient temperature and then depressurized. The rack is then draine d before the tubing sample is removed. This rack is located in the RWCU heat exchanger room, which minimizes exposure or contamination of personnel and maintains contamination c ontrol in case of a spill.

9.3.2.4 Tests and Inspections

The sample nozzles and associated piping, tubing, fittings, and va lves are tested and inspected in accordance with the requirements of the main process pipes from which the samples are taken.

The process sampling system is proved operabl e by its use during norm al plant operation. Grab sampling is provided for laboratory anal ysis for verification of proper operation and calibration of continuous analyzers.

9.3.2.5 Instrume nt Requirements Continuous analyzers and associated record ers monitor conductivity and dissolved oxygen at selected points in the sampling system. Alarms annunciate in the main control room or the radwaste control room when the variables are out of specifica tion to alert the control room operator that correctiv e action is required.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-05-007 9.3-13 9.3.3 EQUIPMENT AND FLOOR DRAINAGE SYSTEMS

9.3.3.1 Design Bases

The equipment and floor drainage systems ar e designed to collect and convey the various operational waste liquids from their points of origin to their points of ultimate disposal under controlled conditions. The following design bases are used to ensure the system integrity during normal plant operation and preclude any danger to the h ealth and safety of plant personnel, the environs, and the general public.

Drainage systems which carry radioactive waste are isolated from drai nage systems which do not carry radioactive waste. Radioactive wastes are collected separately in tanks or sumps, based on their classification as floor drainage (low purity), equipment drainage (high purity), chemical drainage (nonneutral) and detergent drainage, to faci litate their treatment in the radwaste building.

Floor and equipment drains in the diesel generato r building, service buildi ng, isolated areas of the turbine building, and storm water drainage are not intended to be operated as radioactive systems.

The majority of radioactive drai nage piping 1-in. and above is of butt-welded construction to avoid the collection of radioactive solids, see Section 12.3.1.3.2. Drain lines are sloped to ensure complete drainage of piping. Appropri ate shielding is used in locations which could result in increased radiation exposur

e. Cover plates are an integral part of all floor drains to prevent solids from entering drain pi ping and causing subsequent clogging.

9.3.3.2 System Description

Equipment and floor drainage systems are prov ided to handle radioactive and nonradioactive wastes in separate systems. Radioactive wastes are collect ed in the building sumps and transferred to the radwaste system (see Section 11.2) for treatment, sampli ng, and disposal or reuse within the plant. Roof drains are drained by gravity or pumped to the storm drain system.

9.3.3.2.1 Radioactive Equi pment Drainage System

9.3.3.2.1.1 Reactor Building Drains. Reactor building equipment drains are collected in two separate s ubsystems (see Figure 9.3-9

). One subsystem handles drainage from all equipment drains located in the primary cont ainment. The other handles drainage from equipment drains located in remaini ng portions of the reactor building.

The primary containment equipment drain subsystem starts at funnel drains located at pieces of equipment, collects in br anch lines, and discharges to the drywell equipment drain sump. The C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-05-007,05-017 9.3-14 drywell equipment drain sump is the collecting point for valve leakoffs, the inner refueling

bellows seal support drains, reac tor recirculation pump seal lea koffs, and equipment vents and drains within the drywell.

The drywell equipment drain sump is drained th rough a 3-in. line penetrating the containment wall to the reactor building equipment drain sump. This drain line includes two isolation

valves located outside the dryw ell and a flow meter for monitoring of the leakage into the drywell sump during reactor operation. Leakage into the drywell equipment drain sump is considered identified leakage (see Section 5.2.5.2). This drain line is provided with a loop seal to prevent gas flow between the drywell and the reactor building during normal operation.

The containment isolation valves in the drywe ll sump drain line will cl ose on a high pressure signal in the drywell to prevent blowing out the water seals. All equipment drainage piping in the drywell has been designed to Seismic Category 1M requirem ents. The drywell equipment sump drain line is Seismic Category I and is constructed to ASME Section III Class 2 requirements from the sump outlet to th e second containment isolation valve.

The equipment drains for the remainder of the reactor building start at funnel drains located at pieces of equipment, collect in branch lines, and discharge to the reactor building equipment drain sump. The reactor building equipment dr ain sump collects drainage and leakage from the following major sources:

a. Fuel pool gate drainage,

b. Residual heat removal system equipment,

c. Fuel pool cooling system equipment,

d. CRD system equipment,

e. Reactor water cleanup system equipment,

f. High-pressure core spray (HPCS) system equipment,

g. Low-pressure core sp ray system equipment,

h. Reactor core isolation cooling system equipment,

i. Condensate filter demi neralizer backwash pump and auxiliary condensate pumps, and

j. Drywell equipment drain sump.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-05-017 9.3-15 Two sump pumps are installed in the reactor building equipment drain sump to transfer the collected water to the selected collector tank located in the radw aste building. Only one pump is in service at a time. A manual selector switch is located near the pumps to select the active pump.

A level switch in the sump st arts and stops the active pump and timers monitor the fill and pump out rate and energize an alarm(s) in the control room if the settings on the fill and pump out timers are exceeded. A temperature sensor in the sump starts th e active sump pump and directs the sump water through a heat exchanger to cool the wate r prior to being pumped to the radwaste system. All drainage piping in the reactor building has been designed to Seismic Category 1M requirements.

To minimize the release of radioactive contaminants, the reactor building equipment drain sump and drain headers are main tained at a negative pressure and vented through a filter system (see Section 9.4.2).

9.3.3.2.1.2 Turbine Building Drains. The turbine building equi pment drain sumps serve as the collection point for equipment drains from all floors of the building (see Figure 9.3-10

).

The equipment drains start at f unnel drains located at pieces of equipment, collect in branch lines, and discharge to one of two turbine building equipment drain sumps. The turbine building equipment drain sumps collect drainage and leakage collection includes the following major sources:

a. Low-pressure feedwater heaters,

b. High-pressure feedwater heaters,

c. Gland steam evaporators,

d. Reactor feed pumps,

e. Instrument racks,

f. Steam jet air ejector condensers,

g. Mechanical vacuum pumps, and

h. Main condenser.

Two sump pumps are installed in each turbine building equipmen t drain sump since the pumps are not accessible during plant operation due to their location in a high radiation area. These pumps transfer the collected water to the waste collector tank located in the radwaste building.

Level switches are provided in each sump to start and stop the sump pumps at predetermined levels. Drainage piping in th is system is designed to Se ismic Category II requirements.

9.3.3.2.1.3 Radwas te Building Drains. The radwaste building equipment drainage is collected in a separate equipm ent drain sump (see Figure 9.3-11

). The subsystem collects drainage from components containing high purity water. The sump contains a sump pump which transfers the

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-11-005 9.3-16 water collected to the waste co llector tank. The pump is cont rolled by a level switch in the sump.

To minimize the release of radioactive contaminants, the radwaste building equipment drain sump and drain header are maintained at a ne gative pressure and ve nted to the radwaste building ventilation system (see Section 9.4.3).

The radwaste building chemical waste sump coll ects wastes from equipm ent drains and floor drains associated with the chemical waste system (see Section 11.2.2) in a separate sump (see Figure 9.3-11). These wastes are collected and pumped to the chemical waste tank. The pump is controlled by a level switch located in the sump.

Drainage piping in these systems is desi gned to Seismic Cate gory II requirements.

9.3.3.2.2 Radioactive Fl oor Drainage Subsystem

9.3.3.2.2.1 Reactor Building Floor Drains. Reactor building floor drainage is collected in two separate systems (see Figure 9.3-12

). One handles drainage from the drywell which collects leakage from piping and eq uipment. The other handles dr ainage from all floor drains located in the remaining porti ons of the reactor building.

The drywell floor drains subsystem collects leakage in the drywell from piping, valves, and equipment in the drywell floor drain sump. In addition, drains from the drywell coolers are routed to this sump.

The floor drain sump system is a gravity flow feed from the drywell floor drain into one of the reactor building floor drain sumps.

The floor drain sump geometry is such that the gravity feed is connected to a perforat ed standpipe in the center of th e sump. Approximately 50 gal of liquid is required to fill a completely dry sump to the first row of holes in the perforated pipe outlet. A flow transmitter continuously measures flow from the drywell floor drain and supplies this flow information to a flow totalizer in the control room. Drywell floor drain flow is also provided to a recorder whose function is to actuate a control room alarm if flow exceeds a pre-established limit.

The drywell floor drain sump is drained through a 3-in. line penetrating the containment wall to one of the reactor building fl oor drain sumps. This drain lin e includes two isolation valves located outside the containment and an in-line flow meter for monitoring the flow from the drywell sump during reactor operation. Leakage into the drywell floor drain sump is considered unidentified leakage (see Section 5.2.5.2). The drain line is provided with a loop seal to prevent gas flow between the drywell and the reactor building during normal operation.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-05-002,05-007, 05-017 9.3-17 The containment isolation valves in the drywe ll sump drain line will cl ose on a high-pressure signal in the drywell to prevent blowing out the water seal.

All floor drainage piping in the drywell has been designed to Seismic Category 1M requirements. The drywell floor drain sump pipi ng is Seismic Category I and is constructed to ASME Section III Class 2 requirements from the sump out to the second containment isolation valve except 3 ft of 2-in. piping from the diaphragm floor seal, which does not require hydrotest or code stamp.

The floor drain system for the remainder of the reactor building contains four independent sumps. Each sump is located n ear one of the four corners of the building and collects drainage from roughly one quarter of the building.

The four sumps collect water fr om the following typical sources:

a. Floor drains thro ughout the building;

b. Drains from electrical trenches, refueling services boxes, valve boxes, cable reel pit and track drains on the refueling fl oor, pool liner drains, and the new fuel storage vault; and

c. Equipment drains from equipment containing low-purity water, such as reactor building closed cooling water system and standby gas treatment system.

As shown in Figure 9.3-12 , the floor drain piping in the r eactor building drains to one of four sumps listed below.

Floor Drain Sump Room Locations Rooms Served FDR-SUMP-R1 RHR A pump room RCIC RHR A FDR-SUMP-R2 RHR B pump room RHR B FDR-SUMP-R3 HPCS pump room HPCS CRD

FDR-SUMP-R4 RHR C pump room LPCS RHR C Each sump is equipped with level instrume ntation which (a) controls the sump pumps, (b) alarms in the control room (on high sump le vel), and (c) initiates cl osure of the isolation valves in the piping between inte rconnected rooms. Class 1E leve l instrumentation is installed

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 9.3-18 just above floor level in each emergency core cooling system (ECCS) pump room. This instrumentation alarms in the control room.

The floor drain system is analyzed against potential sources of fl ooding as desc ribed in Section 3.6.

The effects of passive failures in the ECCS during post-LOCA l ong term cooling is addressed in Section 6.3. Two of the reactor building floor drain sumps are provided with one sump pump (FD-SUMP-R4 and FD-SUMP-R2). The other two sumps have two pumps, one of which is for backup. The two sumps (FD-SUMP-R3 and FD-SUMP-R1) which have a backup pump, serve rooms containing the reactor core isolation system, residual heat removal system loop A pumps, and the CRD pump, condensate supply pumps, and the HPCS pump.

Floor drain sump FDR-SUMP-R3 al so receives th e leakage from the dryw ell floor drain sump. Each FDR sump draining more than one pump r oom has an isolation va lve installed in the drain header between the conn ected pump rooms. The valve will close on high water level signal from the sump. Thus, except for the sump valve, room doors and piping/cable penetrations, drainage in any pump room, exceeding the capacity of the sump pump, will be confined to that room with the exception of RCIC and CRD pump rooms which are connected

by an unisolable sump pipe.

The reactor building floor drain pumps transfer the collected wate r to the floor drain collector tank located in the radwaste building. Level switches are provided in each sump to start and stop the sump pumps at predetermined levels.

A timing switch in each sump monitors the fill-up rate in the sump. Abnormal sump fill-up rate is alarmed in the main control room.

In addition, wall-mounted level sensing instrumentation is pr ovided in each ECCS pump room to detect passive failures in the ECCS during po st-LOCA long-term cooling and to alarm in the control room. As discussed in Section 6.3.2.5, ample operator time is available after detection of the leak to identify and isolate the source before the leak has a ny adverse effect on the ECCS.

To minimize the release of radioactive contaminants the reacto r building floor drain sumps and drain headers are maintain ed at a negative pressure and vented through a filter system (see Section 9.4.2).

9.3.3.2.2.2 Turbine Building Floor Drains. Turbine building radioactive floor drains are collected in two sumps located in the turbine building (see Figure 9.3-10). Each sump is equipped with two sump pumps. Floor drain water collected in these sumps is pumped to the floor drain collector tank in the radwaste building for processing.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 LDC N-0 2-0 0 5 9.3-19 Level switches are provided in each sump to start and stop the sump pumps at predetermined levels. A mechanical alternator installed in each sump alternates the operation of each pump.

Drainage piping in this system is desi gned to Seismic Category II requirements.

9.3.3.2.2.3 Radwaste Building Floor Drains. Three sumps collect radioactive floor drains and equipment drains and overflows from equipment containing water of low purity (see

Figure 9.3-11

). Two of these sumps pump to t h e floor drain collector tank. The remaining sump, which collects drainage fr om the solid waste handling area, discharges to the waste sludge phase separator.

Level switches are provided in each sump to start and stop the pump at predetermined levels.

Drainage piping in this system is designed to Seis mic Category II requirements and supported to Seismic Category I requirements.

9.3.3.2.3 Nonradioactive Water Drainage System

9.3.3.2.3.1 Turbine and Service Buildings. Equipment and floor drains from normally uncontaminated areas o f the turbine building are col l ec t ed i n three sumps (see Figure 9.3-13

). All these sumps are routed to the radwaste system for processing (see Section 11.2).

Level switches are provided in each sump to start and stop the pumps at predetermined levels.

Area floor drains and equipment drains collected by this system are as follows:

a. Operating floor (el. 501 ft 0 in.)

Air handling units (equipment drains)

Air handling wash pumps (equipment drains)

Instrument panels (equipment drains)

Air washers (equipment drains)

Clean area drains

b. Mezzanine floor (el. 471 ft.0 in.)

Clean area drains Turbine oil coolers (cooling water drains)

HP fluid reservoir (cooling water drains)

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 LDC N-0 0-0 1 2 9.3-20 c. Ground floor (el. 441 ft 0 in.)

Clean area drains

H 2 seal oil unit (cooling water drains) Stator cooling water unit (equipment drain)

Main condenser wate r box (equipment drain) Elevator pit drain Auxiliary boiler blowdown tank (equipment drain)

Auxiliary condensate tank (equipment drain) Service and instrument air compressors (equipment drain)

Equipment and floor drains in the service building are collected in a single sump containing two sump pumps. Water collect ed in the service building floor drain sump is pumped to the storm water drainage system.

Equipment and floor drains in the diesel gene rator building are routed to the storm water drainage system. Building roof drains are collec ted in branch lines whic h empty into a header prior to discharge to the storm water system.

Water collected by the storm wate r drainage system is conveyed by a concrete pipe to a point approximately 1500 ft northeast of the plant. The pipe discharges to an earthern channel that carries the water to a small un lined evaporation/percolation pond. Grab samples of the discharge are analyzed as part of the radiological environmental monitoring program (REMP).

A composite water sampler wa s installed in late 1992.

9.3.3.2.3.2 Miscella neous Drainage System. Liquid waste from curb ed oil equipment areas are directed to separate oil su mps for collection and disposal. These areas are as follows:

a. Turbine lube oil storage and conditioning area, and

b. H 2 seal oil room.

Curbs are provided around fuel oil tanks, auxili ary boiler, and other equipment to contain spillage and prevent oil from entering drainage systems.

9.3.3.3 Safety Evaluation

Each of the sumps in the reac tor, radwaste, turbine generato r, and service buildings are provided with high-high level alarm signals. The sump pump st arts operating on actuation of the high water level switch a nd shuts down on actuation of the low level switch. On sumps containing two pumps the standby pump is started when a high-high wa ter level switch is actuated and an alarm is actuated. Both pumps shut down on actuation of the low level switch.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 9.3-21 Sumps in the reactor building are provided with timers which initiate an alarm when the fill-up rate of a sump exceeds a preset value. These sumps serve as leakage mon itors for the CRD seals, the drywell equipment, the RCIC system, ECCS system, and other systems that support and protect the nuclear steam supply system.

Containment integrity is maintained in the transfer of liquid wa ste from the drywell sumps to the reactor building sump by a loop seal and isolation valves located outside the containment. These valves close on high containment pressure or low reactor vessel water level. In the reactor building, floor drain sump s that serve more than one ECCS equipment room isolate the rooms from one another on a high-high level sump signal that clos es sump influent isolation valves.

In addition, if failure of the normal sump level detection and isolation system in the ECCS pump rooms should occur, a Class 1E leak de tection system is in stalled (see Section 6.3).

9.3.3.4 Testing and In spection Requirements

The equipment and floor drainage systems are in daily use and as such do not require periodic testing to ensure operability. Testing and calibration of the leak detection system components are discussed in Section 5.2.5.

9.3.3.5 Instrumentation Requirements

Sump pumps are equipped with run lights and elapsed time meters located on the radwaste control panel. High-high level alarms for the turbine and radwaste building sumps are also annunciated in the radwas te control room. Reactor building equipment drain sump temperature and reactor building sump level are m onitored in the main control room.

9.3.4 CHEMICAL

AND VOLUME CONTROL SYSTEM

Not applicable to BWRs.

9.3.5 STANDBY

LIQUID CONTROL SYSTEM

9.3.5.1 Design Bases

The standby liquid control (SLC) syst em meets the following design bases:

a. Backup capability for reactivity control is provided, independent of normal reactivity control provisions in the nuclear reactor, to shut down the reactor if the normal control ever becomes inope rable. The SLC system equipment essential for injection of neutron absorber solution into the reactor is designed to Seismic Category I requir ements (see Section 3.7). The system piping and

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 9.3-22 equipment are designed in accordance with the requirements stated in Sections 3.2 and 3.9;

b. The system has the capac ity for controlling the reactivity difference between the steady-state operating conditi on of the reactor with voids and the cold shutdown condition, including shutdown margin, to ensure complete shutdown from the most reactive condition at any time in core life;

c. The system has th e capability to inject sodium pe ntaborate solution into the RPV in response to a LOCA to control the pH in the suppression pool. Re-evolution of iodine from the suppr ession pool water ca n be minimized by maintaining the suppression pool pH level greater than 7.0.

d. The system meet s the requirements of 10 CFR 50.

62 for response to anticipated transients without scram (ATWS). Syst em modifications have been done in conformance with the criteria presented in the BWR Owner's Group ATWS Licensing Topical Report, NEDE-31096-P, and in Generic Letter 85-06;

e. The time required for actuation and effectiveness of the SLC system is consistent with the nuclear reactivity ra te of change predicted between rated operating and cold shutdown condition. This system is not safety related and is not designed to provide a fast scram of th e reactor or operation of fast reactivity transients;

f. The functional performan ce capability of the SLC system components can be verified periodically under test conditi ons consistent with system operating parameters. In addition, demineralized water, rather than the actual neutron absorber solution, can be injected during cold shutdow n or refueling into the reactor to test the operation of all redundant components of the system;

g. The neutron absorber is dispersed with in the reactor core in sufficient quantity to provide a reasonable margin for leakage or imperfect mixing; and

h. The possibility of unintentional or acci dental shutdown of the reactor by this system is minimized.

9.3.5.2 System Description

The SLC system (see Figure 9.3-14) is manually initiated through two keylock switches from the main control room to pump a boron neutr on absorber solution into the reactor if the operator determines the reactor cannot be shut down with the control rods or suppression pool pH control is required to mitigate the dose consequences of a LOCA.

The key locked switches prevent inadvertent injection of neutron absorber by the SLC system.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-11-001,11-013 9.3-23 The two key locked switches are provided to ensure positive action from the main control room should the need arise. Standard power plant procedural controls are applied to the operation of the key locked control room switches.

The boron solution tank, the test water tank, the two positive displacement pumps, the

two explosive valves, the two motor-operated pump suction valves, and associated local valves and controls are located in the reactor building.

The solution is piped into the reactor vessel and discharged into the core via the HPCS spray header so it mixes with the cooling water (see Section 5.3). In the event of a LOCA, the mixed solution and cooling water flows to the suppression pool by flowing through the break and downcomers.

The SLC system can deliver enough sodium pentaborate solu tion into the reactor (see Figure 9.3-15) to ensure reactor shutdown. As well, the SLC system can inject sufficient sodium pentaborate solution to ensure the pH in the suppression po ol is maintained greater than

7.0 following

a LOCA. The sodium pentaborate solution is prepared in the SLC tank. The sodium pentaborate enrichment is verified prior to addition to the SLC tank by isotopic analysis. The required sodium pentaborate concentration and solution volume in the SLC tank to ensure reactor shutdown capability is included in the Technical Specifications. An air sparger is provided in the tank for mixing.

To prevent system plugging, the tank outlet is raised above the bottom of the tank. Sample s for analyses are take n in accordance with approved procedures.

The saturation temperature of th e solution is 67°F at the maximum concentration of 15.0% (see Figure 9.3-16

). An automatic electrical resistance heater system provides heat to maintain the solution temperature greater than saturation conditions for sodium pentaborate to prevent precipitation during storage. The pump suction piping from the storage tank to the pump suction valves is also electrically heat tra ced. Administrative controls limit the solution temperature to less than 150°F during heatup a nd mixing using manual temperature control.

Liquid level in the storage tank is alarmed at either high level or low level.

The positive displacement pumps are sized to inject the solution in to the reactor in approximately 1 hr with both pu mps operating. The system de sign pressure between the pump discharge and the explosive valves is 1400 psig, at which pressure the two relief valves are set.

To prevent bypass flow from one pump in case of relief valve failure in the line from the other pump, a check valve is installed downstream of each relief valve line in the pump discharge pipe.

The two explosive-actuated injection valves provide assurance of opening when needed and

ensure that boron will not leak into the reactor even when the pumps are being tested. Each

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 9.3-24 explosive valve is closed by a plug in the inlet chamber. The plug is circumscribed with a deep groove so the end will readily shear off when pushed with the valve plunger. This opens the inlet hole through the plug. The sheared end is pushed out of the way in the chamber, and it is shaped so it will not bl ock the ports after release.

The shearing plunger is actuated by an explosive charge with dual ignition primers inserted in

the side chamber of the valve. The two ignition circuits of each explosive valve are monitored for continuity. If either circuit opens, a bypass and inoperable st atus indication (BISI) display occurs in the main control room.

The SLC system is actuated by two key locked switches on the main control room console.

This ensures that switching from the "off" position is a deliberate act. Operating either switch starts one of the injection pumps, actuates both of the explosive valves, opens both pump suction motor-operated valves, and closes the RWCU system outboard isolation valve to prevent loss or dilution of the boron. (This is the sole purpose of this RWCU isolation. This isolation is not related to RWCU isolation to ensure containment integrity.)

A green light in the control room indicates that power is available to the pump motor contactor and that the contactor is deenergized (pump not running). A red light indicates that the contactor is closed to energize.

Storage tank liquid level, tank outlet valve position, pump discharg e pressure, flow indication, and loss of continuity on the explosive valves indicate that the system is functioning. Cross piping and check valves ensure a flow path through either pump and either explosive valve.

Pump discharge pressure and system flow is indicated in the control room.

Equipment drains and tank overflo w are not piped to the radwaste system but to separate containers (such as 55-gal drums) that can be remove d and disposed of to prevent any trace of boron from inadvertentl y reaching the reactor.

Instrumentation consisting of solution temperature indication and control, solution level, and heater system status is provided locally at the storage tank. The SLC system is seismically qualified from the storage tank (including the tank) to the injecti on point to the HPCS piping. Seismic category and quality class are included in Table 3.2-1. Principles of system testing are discussed in Section 9.3.5.4.

9.3.5.3 Safety Evaluation

The SLC system is a reactivity control system and is maintained in an operable status whenever the reactor is critical. The SLC syst em may also be used to maintain suppression pool pH above 7.0 following a LOCA and is ma intained operable when critical or in hot shutdown.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-11-001 9.3-25 The system is designed to bring the reactor from rated power to a cold shutdown at any time in core life. The reactivity compensation provi ded will reduce reactor power from rated to zero level and allow cooling the nuclear system to below 20 0°F with the control rods remaining withdrawn in the rate d power pattern. It includes th e reactivity gain s that result from complete decay of the rated power xenon inventory. It also includes the positive reactivity effects from eliminating steam voids, changing water density from hot to cold, reduced doppler effect in uranium, reducing neutron leakag e from boiling to cold, and decreasing control rod worth as the moderator cools.

The minimum average concentration of natural boron in the reactor to provide adequate shutdown margin, after ope ration of the SLC system, is 780 ppm. Calculation of the minimum quantity of sodium pentaborate to be injected into the reactor is based on the required 780 ppm average concentration in the reactor coolant, including recirc ulation loops, at 70°F and reactor normal water level. The amount is increased by 25% to allow for imperfect mixing and leakage. An additional 275 ppm is provided to accommodate d ilution by the RHR system in the shutdown cooling mode. This concentration is achieved when the so lution is prepared as described in Section 9.3.5.2 and maintained above saturation temperature.

The saturation temperature of the maximum co ncentration solution is 67°F. To ensure complete solubility of the solution, a tank operati ng heater is provided wh ich turns on when the temperature drops below approximately 80°F. Th e tank heater turns off when the temperature increases to approximately 90°F.

Cooldown of the nuclear system requires a mi nimum of several hours to remove the thermal energy stored in the reactor, cooling water, and associated equipment. Use of the main condenser and various shutdown cooling systems requires 10 to 24 hr to lower the reactor vessel to room temperature (70°F

); this is the condition of maximum reactivity and, therefore, the condition that requires the maximum concentration of boron.

To mitigate the consequences of an ATWS, the SLC system is capable of injecting the equivalent in reactivity control of 86 gpm at 13.0% (by weight) of sodium pentaborate solution per the requirements of 10 CFR 50.62. Each SLC pump injects at not less than 41.2 gpm at 13.6% concentration which is math ematically equivalent to 86 gpm at 13.0%. At this injection rate with only one pump, the solution is injected into the reactor in approximately 2 hr which provides negative reactivity insertion considerab ly quicker than the re activity increase caused by the cooldown. The SLC injection path via th e HPCS spray ring provid es sufficient mixing so that any power oscillations are precluded.

This same amount and con centration of boron, at the flow rate of a single pump is adequate to maintain the suppression pool pH greater than 7.0, when used in a LOCA.

The SLC system has redundant el ectrical components requiring AC power to actuate for sodium pentaborate solution inj ection. These divisional compone nts are powered by separate safety-related AC divisions that are back ed by onsite emergenc y diesel ge nerators.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 9.3-26 The SLC system and pumps have sufficient pressure margin, up to the system relief valve setting of approximately 1400 psig, to ensure solution injection into the reactor above the normal pressure in the reactor. The nuclear system safety/relief valves begin to relieve pressure at approximate ly 1100 psig. Therefore, the SLC system positive displacement pumps cannot overpressurize the nuclear system.

The SLC system is evaluated against the app licable General Design Criteria as follows:

Criterion 2

The SLC system is located in the area outsi de of the primary cont ainment and below the refueling floor. In this location, it is protected by walls from external natural phenomena such

as earthquakes, tornadoe s, hurricanes, and floods and also from the effects of internal postulated accident events.

Criterion 4

The SLC system is designed for the expected environment in th e compartment in which it is located. In this compartment, it is not subject to the cond itions postulated in this criterion such as missiles, whipping pipes, and discharging fluids. The SL C system components are qualified in accordance with the CGS Equipment Qualification Program to operate in a post-LOCA environment for a minimum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Criterion 26

The SLC system is a backup reactivity contro l system for the normal reactivity control systems.

Criterion 29

Although GDC 29 is not a design ba sis, certain valves outboard of the isolation valves are redundant. Two suction valves a nd two injection valves are arra nged and cross-tied such that operation of either one of a pair results in successful operation of the system. The SLC system also has test capability.

A special test tank is supplied for pr oviding test fluid for the injection test.

Pumping capability and suction valve operability may be tested at any time. A trickle current continuously monitors continuity of the firi ng mechanisms of the in jection squib valves.

Regulatory Guide compliance is described in Section 1.8.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-11-001 9.3-27 This system is used in the special plant cap ability demonstration ev ents, "Reactor Shutdown and Cooldown - ATWS" and "Reactor Shutdown and Cooldown Without Control Rods."

These events are extremely low probability non design basi s postulated accidents. Various single failure analytical exer cises can be examined to show additional capabilities to accommodate further plant system degradations. See GE Topical Report NEDO-20626, dated October 1974, Studies of BWR Design for Mitigation of ATWS.

9.3.5.4 Testing and In spection Requirements Operational testing of the SLC system is performed in accordance with applicable Technical Specifications. The SL C system injection train components and piping are included in the CGS Inservice Inspection (ISI) and Inservice Testing (IST) Programs.

The reactor must be in mode 4 or 5 before changing from the SLC syst em standby mode to the injection test mode.

The concentration of the sodium pentaborate in the solution tank is determined periodically by chemical analysis using samples taken directly from the tank. The sodium pentaborate enrichment is verified prior to add ition to the SLC tank by isotopic analysis.

Should the boron solution ever be injected in to the reactor, either intentionally or inadvertently, the boron can be removed from th e reactor coolant system by flushing, followed by operating the RWCU system. There is practically no effect on reactor operations when the

boron concentration has been redu ced below approximately 50 ppm.

The SLC system preoperational te st is described in Section 14.2.

9.3.5.5 Instrumentation Requirements

The instrumentation and control system for the SLC system is designed to allow the injection of neutron absorber solution into the reactor and to maintain th e neutron absorber solution well above the saturation temperature. A further di scussion of the SLC system instrumentation is provided in Section 7.4.1.2.

9.

3.6 REFERENCES

9.3-1 "Compressed Air Systems," Design Basis Document, Section 305.

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 Table 9.3-1 Equipment Charac t eri s tics 9.3-29 Control air system

a. Air compressors

1. Quantity 3 2. Rated output (e ach) 387 +/- 4% scfm

3. Rated discharge pressure 115 psig (after filters) b. Air receivers

1. Quantity 3 2. Volume (each) 96 ft 3 c. Air desiccant dryers 1. Quantity 2 2. Type Heatless regenerative

3. Desicca n t Activa t ed a l umina

4. Rated inlet drying capacity (each skid) 750 scfm 5. Dewpoint at outlet -40°F

6. Cycle of operation 4 or 10 minutes d. Air dryer (refrigerated) 1. Quantity 2

2. Type Freon R-22 3. Maximum capacity 830 scfm (each) 4. Dewpoint 39°F +/- 4°F

5. Operation Continuous Service air system

a. Air compressor 1. Quantity 1

2. Rated output 630 scfm at 100 psig

3. Pressure modulating range 100 psig to 110 psig C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 Table 9.3-1 Equipment Characteristics (Continued) 9.3-30 b. Air Receiver 1. Quantity (main header) 1

2. Volume 140 ft 3 c. Service air dryer 1. Quantity 1 2. Type Refrigerated (Freon R-22) 3. Maximum capacity 650 scfm

4. Dewpoint at outlet +40°F

5. Operation Continuous Containment instrument air system

a. Air receiver 1. Quantity 1

2. Volume 34 ft 3 b. Air dryer 1. Quantity 1

2. Type Twin tower, heat regenerative

3. Media: active buffer Silica gel Activated alumina absorbent 4. Peak inlet drying capacity 84 scfm (minimum)

5. Dewpoint at outlet -40°F

6. Cycle of operation "As needed" basis (moi sture sensor)

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 Table 9.3-2 Available Sample Locations

Sample Point a Description b Sampling Location c Activit y d Analysi s e LDC N-0 1-0 6 6 9.3-31 SP-1 Reactor water, recirculating inlet manifold SR-9 (SR-48) H Conductivity (dissolved oxygen

dissolved hydrogen)

SP-2A,B Main steam SR-1 H Grab sample SP-3 Feedwater, after heaters HP-6A,B SR-1 L Conductivity, dissolved oxygen, corrosion p r oduct SP-4 Condensate, after condensate pumps 1A,B,C SR-1 L Grab sample SP-5 Condensate after gland seal steam condenser Local L Grab sample SP-6 Condensate after ejector condensers Local L Grab sample SP-8 Influent to cleanup filter demineralizers A,B SR-9 (SR-48) H Conductivity (dissolved oxygen

dissolved hydrogen) SP-9A,B,C Condensate after low pressure heater 1A,B,C Local L Grab sample SP-10 Waste collection filter EDR-DM-9 outlet SR-8 L Grab sample SP-11 Condensate after condensate pumps 1A,B,C SR-1 L Conductivity SP-12A,B,C Condensate after low pressure heaters 4A,B,C SR-1 L Grab sample SP-13A Condensate after low pressure heater 5A SR-1 L Grab sample C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 Table 9.3-2 Available S a mple L o cations (Continued)

Sample Point a Description b Sampling Location c Activit y d Analysis e LDC N-9 9-0 6 2, 0 2-008 9.3-32 SP-13B Condensate after low pressure heater 5B SR-1 L Grab sample SP-14A Condensate after filter demineralizer 1A SR-9 L Conductivity SP-14B Condensate after filter demineralizer 1B SR-9 L Conductivity SP-14C Condensate after filter demineralizer 1C SR-9 L Conductivity SP-14D Condensate after filter demineralizer 1D SR-9 L Conductivity SP-14E Condensate after filter demineralizer 1E SR-9 L Conductivity SP-14F Condensate after filter demineralizer 1F SR-9 L Conductivity SP-15 Combined condensate after filter demineralizers SR-9 L Conductivity, dissolved oxygen SP-17A Condenser hotwell SR-1 Spared in place SP-19A RCIC pump P-1 discharge SR-6 Spared in place SP-19B RCIC water to containment SR-6 Spared in place SP-20 HPCS pump P-1 discharge SR-6 Spared in place SP-21 LPCS pump P-1 discharge SR-6 Spared in place SP-22A After RHR heat exchanger A SR-6 H Grab sample C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 Table 9.3-2 Available S a mple L o cations (Continued)

Sample Point a Description b Sampling Location c Activit y d Analysis e LDC N-9 9-0 6 2 9.3-33 SP-22B After RHR heat exchanger B SR-6 H Grab sample SP-23 After RCC heat exchangers 1A,B,C Local L Grab sample SP-24 Inlet to fuel pool demineralizers SR-9 L/H Conductivity SP-25A Outlet from fuel pool demineralizer A SR-9 L Conductivity SP-25B Outlet from fuel pool demineralizer B SR-9 L Conductivity SP-26 Drywell equipment cooling water return SR-6 Spared in place SP-27A After clean u p filter A SR-9 H Conductivity SP-27B After clean u p filter B SR-9 H Conductivity SP-28 Nonregenerative heat exchanger shell outlet Local L Grab sample SP-29 Waste collector pump EDR-P-11, discharge SR-8 L Conductivity SP-30 Waste surge pump EDR-P-15, discharge SR-8 L Grab sample SP-31 Floor drain collector pump FDR-P-16 discharge SR-8 L Conductivity SP-32 Floor drain filter FDR-FU-10, discharge SR-8 L Grab sample SP-33A Waste demineralizer EDR-DM-29 inlet SR-8 L Conductivity

C OLUMBIA G ENERATING S TATION Amendment 56 F INAL S AFETY A NALYSIS R EPORT December 2001 Table 9.3-2 Available S a mple L o cations (Continued)

Sample Point a Description b Sampling Location c Activit y d Analysis e 9.3-34 SP-33B Waste demineralizer EDR-DM-29 outlet SR-9 L Conductivity SP-34A Waste sample pump EDR-P-14A, discharge SR-8 L Grab sample SP-34B Waste sample pump EDR-P-14B, discharge SR-8 L Grab sample SP-35A Floor drain demineralizer FDR-DM-111, inlet SR-8 L Conductivity SP-35B Floor drai n demineralizer FDR-DM-111, outlet SR-9 L Conductivity SP-36 Floor drain sample pump FDR-P-21, discharge SR-8 L Grab sample SP-37 Effluent from centrifuges PWR-FU-94A,B Local L Grab sample SP-38A Chemical waste pump MWR-P-26A, discharge SR-8 L Grab sample SP-38B Chemical waste pumps MWR-P-26B, discharge SR-8 L Grab sample SP-43 Detergent drain pumps MWR-P-20A,B discharge Local L Grab sample SP-44 Cleanup decant pump RWCU-P-27 discharge SR-8 H Grab sample SP-46 Condensate decant pump CPR-P-24 discharge SR-8 L Grab sample C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 Table 9.3-2 Available S a mple L o cations (Continued)

Sample Point a Description b Sampling Location c Activit y d Analysis e LDC N-0 0-0 1 2 9.3-35 SP-47 Condensate sludge discharge mixing pump

CPR-P-25 discharge Local L Grab sample SP-48A Auxiliary boiler fee d water Local N Grab sample SP-48B,C Auxiliary boiler blow down Local N Grab sample SP-59 Demineralized water

system supply SR-5 N Grab sample SP-61 Potable water supply SR-5 N Grab sample SP-62 RCC heat exchangers 1A,B,C, service water

outlet Local N Grab sample SP-63A RHR heat exchanger 1A service water outlet Local N Grab sample SP-63B RHR heat exchanger 1B service water outlet Local N Grab sample SP-65A Condensate storage tank 1A Local N Grab sample SP-65B Condensate storage tank 1B Local N Grab sample SP-67C Polishing demineralizer MWR-DM-4D, outlet SR-8 L Conductivity SP-69 Fuel pool demineralizer outlet SR-9 L Conductivity SP-70A,B Carbon filter FW-FU-2A,B inlet Local N Grab sample C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 Table 9.3-2 Available S a mple L o cations (Continued)

Sample Point a Description b Sampling Location c Activit y d Analysis e 9.3-36 SP-71 Condensate filter demineralizer, inlet header SR-9 L Conductivity SP-115 COND-P-5 suction combined CSTs Local L Grab sample SP-1088 COND-P-4 suction combined CSTs Local L Grab sample a Sequential numbering of sample points.

b Indicates type of system being sampled, i.e., reactor water, main steam, condensate, etc., and general location of the process line tap.

c Sample is taken from either a local process line tap or in a sample stat i on rack, i.e

., TBSR Turbine building sample rack, SR-1 RWSR Radwaste building sample racks, SR-8, SR-9 RBSR Reactor building sample rack, SR-6 SBSL Service building sample laboratory, SR-5 d The sample is expected to have H (H igh), L (Low), or N (No) radioactivity.

e Type of continuous monitor and/or a grab sample.

Amendment 59December 2007 Form No. 960690ai Columbia Generating StationFinal Safety Analysis ReportRev.FigureDraw. No.9.3-1.1 80 M510-1Control and Service Air System Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.3-1.2 36 M510-2Control and Service Air SystemRev.FigureDraw. No.Amendment 61December 2011 Amendment 60December 2009 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 09.3-01.3 14 M510-2AControl and Service Air SystemRev.FigureDraw. No.

Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.3-1.4 24 M510-3Control and Service Air SystemRev.FigureDraw. No.Amendment 61December 2011 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.3-1.5 9 M510-4Control and Service Air SystemRev.FigureDraw. No.Amendment 61December 2011 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.3-2.1 50 M556-1Containment Instrument Air SystemRev.FigureDraw. No.Amendment 61December 2011

Draw. No.

Rev.Figure Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.3-3 49 M607-1 Steam and Liquid Sampling - Turbine and Service BuildingsRev.FigureDraw. No.Amendment 61December 2011 Amendment 60December 2009 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 09.3-04 26 M607-2Steam and Liquid Sampling - Reactor BuildingRev.FigureDraw. No.

Amendment 59December 2007 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.3-5 28 M607-3 Steam and Liquid Sampling - Radwaste BuildingRev.FigureDraw. No.

Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.960690.79General Sample Probe 9.3-6 8 inch Sampling of Process Fluids 22A2708, Rev 5 Reference Specification RootValve (SS)

Reducer (SS)Machine permanent witness marks on pipe to indicate port orientationTubing (SS) 1/2" 0.145D Process Flow 1/8" dia. hole facing upstream End of pipe welded closed without en-larging pipe O.D.

Pipe, seamless 3/4" Stainless Steel Double Extra Strong(ASTM A312 GR TP316 or 304)

C L Pipe Columbia Generating StationFinal Safety Analysis Report D =Process Pipe I.D.

Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.960690.94Sample Probe RFW-SP-3 9.3-7 R = .375"Typ.Existing 1 1/2" Sch. XX-STG Red WeldoletDETAI L I Scale: 2X SAMPLE PROBE SP-3 Scale: Half SizeNote 1: Sample Probe to Weldolet Weld Omitted for ClarityGrind Weldolet End Prep Back to .400".400" 1.338"ø O.D. .03" 9693"ø .005" 1.05".03" R=.375" 1.338"ø.03" DETAI L II Scale: Full 1.50" C L 1/32" Chamfer

.875"ø.25" 1.50"ø Hole Existing

.50" 1.50" 64 125 8" See Det. I 1/4.375"ø Ream After Port

Hole is Drilled See Detail II 1/4 1.338"ø .03" (See Note 1)3/4" Root Valve Center punch on same side of sampling nozzle as

port hole so that sampling

port may be correctly directed facing flow.

Drill Cleanly .125"ø Port (No Chamber Required).06"1 11/16" (ref)

.25" Columbia Generating StationFinal Safety Analysis Report Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.960690.80Feedwater Condensate Sample Probe 9.3-8 8 inch Sampling of Process Fluids 22A2708, Rev 5 Reference Specification RootValve (SS)

Reducer (SS)Machine permanent witness marks on pipe to indicate port orientationTubing (SS) 1/2" Process Flow (3) 1/8" dia.

hole facing upstream End of pipe welded closed without en-larging pipe O.D.

Pipe, seamless 3/4" Stainless Steel Double Extra Strong(ASTM A312 GR TP316 or 304)

D =C L Pipe 0.35D 0.455D 0.145D Columbia Generating StationFinal Safety Analysis Report Process Pipe I.D.

Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.3-9 75 M537Equipment Drain System - Reactor BuildingRev.FigureDraw. No.Amendment 61December 2011 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.3-10 35 M538Floor Drains - Turbine Building EquipmentRev.FigureDraw. No.Amendment 61December 2011 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report9.3-11 52 M540Radwaste Building Equipment and Floor DrainsRev.FigureDraw. No.Amendment 61December 2011 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.3-12 85 M539Floor Drains - Reactor BuildingRev.FigureDraw. No.Amendment 61December 2011 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 09.3-13 35 M518Nonradioactive Floor DrainsRev.FigureDraw. No.Amendment 60December 2009 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.3-14 37 M522Standby Liquid Control System - Reactor BuildingRev.FigureDraw. No.Amendment 61December 2011 Solution Concentration Weight % Na 2 B 10 O 16 2 O Figure Amendment 61 December 2011 Form No. 960690Draw. No.Rev.960690.77Sodium Pentaborate Decahydrate VolumeConcentration Requirements 9.3-15 Columbia Generating StationFinal Safety Analysis ReportSodium Pentaborate Decahydrate (Na 2 B 10 O 16 2 O)Volume Concentration RequirementsRegion of ApprovedVolume Concentration Net SLC Tank Volume (gal.)

Minimum Concentration 13.6%Low Level Limit (4787 gal.)

Over flow (5150 gal.)

15.0 16.0 14.0 13.0 12.04600470048004900500051005200The minimum required volume to ensure reactor shutdown is 4587 gal. The Low Level Limit (4587 + 200 gal) includes 200 gal process margin to minimize air entrainment in the pumps. Operators may shutoff the pumps at 200 gal net SLC

Tank Volume or less.

LDCN-11-013 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.960690.78Saturation Temperature of Sodium Pentaborate Solution 9.3-16 140 130 120110 100 90 80 70 60 50 400510152025303540Temperature ( F)Percent Sodium Pentaborate by Weight of Solution Columbia Generating StationFinal Safety Analysis Report

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 LDC N-9 8-1 4 2 9.4-1 9.4 HEATING, VENTILATING, AND AIR CONDITIONING SYSTEMS

The various heating, ventilati ng, and air conditioni ng (HVAC) systems serving the plant are designed to provide suitable environmental conditions throughout the plant for personnel

comfort and/or equipment operation.

The following performance objectives are implem ented in the design of the HVAC systems:

a. Maintain appropriate ambient temper ature and humidity conditions for station operating personnel and equipment, b. Control and monitor all potentially radioactive airborne releases from the plant so that releases are within the limits of 10 CFR Part 20, c. Control and limit airborne radioactive contaminants within the plant structures by inducing air flow from areas of low contamination potential into areas of progressively higher contam ination potential, and

d. Remove all potentially explosive gases, noxious fumes, or smoke from the plant.

In addition, a number of HVAC systems are requi red to ensure a safe shutdown of the reactor or to mitigate the results of the design basis accident. These systems are designed to continue operation in the event of any or all of the following events:

a. Safe shutdown earthquake (SSE), b. Loss of offsite power,

c. Single failure of any active component, and

d. Design basis accident.

The effect of a loss of normal ventilation during a station blackout is discussed in Appendix 8A.

The following areas containing engineered safety features (ESF) equipment are serviced by critical HVAC systems:

a. Main control room/cable spread ing room/critical switchgear area, b. Diesel generator building,

c. Standby service wa ter (SW) pump houses, d. Reactor building em ergency pump rooms, e. Reactor building critical elec trical equipment rooms, and f. Diesel generator cable area corridor.

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 9.4-2 The balance of the plant structures are serviced by noncritical HVAC system

s. The following outdoor design cond itions are used in the design of H VA C systems having ESF:

Summer: 105°F d r y-bulb, 71°F w e t-bulb Winter: 0°F dry-bu l b The extreme outdoor conditions are:

Summer: 115°F d r y-bulb Winter: -27°F

The e f fect o f the s e extreme ou t door conditions on main control room/cable spreading room and critical switchgear area temperatures is negligi b le for the f r equency and duration of these

conditions since the rooms are interior rooms and the total load changes due to fresh air (maximum 5% of total air flow for control room and maximum 10% of total air flow for

critical switchgear are a) are within equipment capacity li m its for maintaining inside design

conditions. The mecha n ical equipment rooms h ousing these three sys t ems are of e x tra heavy exterior wall and roof construct i on and the ef f e ct of extre m e outs i de conditions is considered to be negligible for the frequency and duration of these ext r emes.

Extreme outside conditions have no effect on reactor building emergency pump rooms and

reactor building critical electri c al equipment rooms since t h ese are interior rooms without any

outdoor air supply during emerg e ncy. During normal operat i on the emergency pu m ps are not

runn i ng and the critical e l ectrical equipment is operating at a reduced load. The battery rooms are provided with room heaters to prevent exce s s ive cooling from SW that could a f fect bat t ery capacity.

The effect o f extreme o u tdoor conditions on t h e diesel generator building ventilation system, diesel generator area cable cooling syste m , and SW pump house ventilation syste m s is discussed in Sections 9.4.7 , 9.4.8 , and 9.4.10 respectively.

9.4.1 MAIN CONTROL ROOM/CABLE SPREADING ROOM

/C RITICAL SWITCH G EAR AREA 9.4.1.1 Design Bases The crit i cal switchgear a r ea, cable s p reading room, and main control room are loca t ed, one above the other, on thr e e success i ve levels of the radwaste b u ilding, with the main control room on the top level. Each level is served by a separate H VAC system. Redundant HVAC systems are prov i ded for the m a in c o ntrol room and the ca b le spreading r oom. These three syste m s are ESF s y stems and all system components, except the r a dwaste building chilled water system (WCH) (see Section 9.4.4) to the control room, cable spreading room and critical

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-11-003 9.4-3 switchgear area systems, and the plant service water (TSW) system to the switchgear area system, are designed to operate under all emergency modes. Du ring an emergency condition, the SW system is used as the cooling medium for the cable spreading room and switchgear area systems. The control room chilled water (CCH) system or SW system is used as the cooling medium for the control room HVAC system during emergency conditions. The SW system provides sufficient cooling for equi pment operability. The CCH system provides additional cooling capacity for personnel comfort.

The three systems are de signed to satisfy the following design criteria:

a. Main control room

During normal operation (radwaste chillers operating/SW not operating) the main control room ambient conditions are normally maintained at 75°F +/-3° dry-bulb temperature. The main cont rol room temperature is verified by surveillance test to be less than 85°F. In the event both ra dwaste chillers are inoperative (emergency condition) the control room temperature will be maintained within the design limit (104°F) by contro l room chilled water or SW. The ingress of smoke or combustion vapors (due to a fire within the plant but external to the control room), or of airborne radioactive contaminants released due to the design basis acci dent, is minimized by pressurizing the control room. During a loss-of-coolant accident (LOCA), the control room emergency pressurization mode through the emergency filter unit maintains a positive pressure with resp ect to its surroundings as measured in the cable spreading room.

Three air intakes are provided from which fresh air can be drawn. One local intake is provided for normal operation and two remote intakes are provided for normal and emergency ope ration. Fire external to the plant and any ingress of smoke or combustion vapors are detected by smoke detectors in the control room fresh air intake ducting, which will alarm in the control room.

Isolation of the control room fresh air intakes would place the control room HVAC in an unfiltered recirculation mode. In the event of a hazardous chemical release, the control room HVAC is manually isolated into the recirculation mode without filtration through the emergency filter units by closing the normal fresh air isolation damper.

b. Cable spreading room

The cable spreading room HVAC system is designed to maintain the cable spreading room and the re mote shutdown room at approximately 80°F during

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-028 9.4-4 normal operation and to limit the temp erature below equipment operability limits during all emergency modes of operation. See Table 3.11-1. c. Critical swit chgear area

The critical switchgear area HVAC system is designed to maintain temperatures in the electrical rooms between 55°F a nd 104°F during normal operation and to limit the temperatures below equipment operability limits during all emergency modes of operation. See Table 3.11-1. The system is also designed to remove any combustible fumes generate d by the emergency batteries.

9.4.1.2 System Description

The HVAC systems of the main control room, cable spreading room, and critical switchgear areas are shown in Figure 9.4-1. The HVAC systems for thes e three areas are located in equipment rooms above the main control room. These two equipment rooms are separated from each other by a missile barrier. Each equipment room houses three separate and independent systems serving the three areas. Equipment details are given in Table 9.4-1. Equipment seismic information is given in Table 3.2-1.

9.4.1.2.1 Main Control Room

Each of the main control room's 100%-capacity HVAC systems are composed of a primary air handling system and an emergency filter system. The two HVAC systems share a common outside air intake system and a common duct distribution system within the main control room.

A single exhaust system, composed of a fan, shutoff damper, and ductwork, discharges air

from the main control room toilet and kitchen.

The exhaust fan opera tes continuously during normal operations.

Each primary air handling system consists of a centrifugal supply fan which blows through an air handling unit consisting of an air filter, two water cooling coils in series (one for radwaste chilled water and one for control room chilled water or SW), and an electric blast coil heater and associated ductwork and dampers.

Separate return air ductwork containing a sound absorber unit is provided from the main control room to each of the primary air handling systems.

During normal operation one air handling system operate s, distributing air to the main control room. The temperature is controlled by electronic controllers located in the main control room which modulate the chilled wate r flow to the cooling coil.

Chilled water is normally supplied to the main control room air handling systems by the WCH.

The WCH, which includes two 100%-capacity chillers and two 100%-capacity pumps is not an

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.4-5 ESF. During emergency condition, control room chilled water or SW is supplied to the air handling units for cooling. The control room can be maintained below 85°F by the control room chilled water, or SW can be used to maintain less than 104°F (shedding of nonessential loads may be required under some conditions). The environmental qualification temperature limit for control room equipment is 104°F and 85°F equivalent temperature for control room personnel habitability.

The two 1000-cfm capacity filter systems are norma lly in standby and opera te in the event of an emergency (F, A, Z signal). Each of the emergency filter sy stems consists of an emergency filter unit, a 5-kW electric heater, bypass and recirculation control dampers, and associated ductwork. Each emergency filter unit consists of a medium efficiency prefilter, high efficiency particulate air (HEPA) filter, ac tivated charcoal filt ers, and direct driv e centrifugal fan, all enclosed in an all-welded sheet metal housing. A de luge water spray syst em is provided to soak the charcoal filter s in the event of high te mperatures in the charco al beds. Check valves are provided on all drain connections from the filt er unit, and the drain header is provided with a deep water seal trap to prev ent inleakage of air during unit operation. The electric heater located in the fresh air duct to each emergency filter limits the relative humidity of the air entering the filter to 70%.

The medium efficiency prefilters are provided to protect and extend the life of the HEPA filters. They have an 80% to 85% dust spot efficiency by ASHR AE Standard 52-68.

Regulatory Guide 1.52 complian ce is described in Section 1.8.

Motor-operated outside air intake (bypass) dampers WMA-AD-51A-1 and 51B-1 are spring-loaded fail-closed type and are provided with limit switches to indicate full open and full closed positions on the main control room panel. Each damper is provided with a remote manual switch in the main control room. Dampers are automatically closed when deenergized by isolation signal or by the ma in control room panel-mounted switch. These dampers have a design leak rate of 0.5% of the rated flow. Main control room supply fan inlet pressure is higher than the emergency filter unit fan; therefore, when the emergency filter unit is operating, a negative pressure is developed on the inlet side of the bypass dampers preventing any contaminated air bypass of the emergency filter unit. In the event that either bypass damper does not close, an alarm will be activated in the main control room.

The three fresh air intakes (one normal and two remote) for the main control room are fitted with two butterfly isolation valves in series. The normal control room fresh air local intake valves are automatic isolation va lves which isolate on an F, A, or Z signal. The normal fresh air intake valves have electrohydraulic operators which are power ed from the Class 1E buses.

All fresh air intakes are connected via ductwork to a common intake header from which both main control room air handling systems and both emergency filter systems draw fresh air. The isolation valves in the purge lines and the normal air intake, position and radiation indication

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.4-6 on all fresh air intakes, the emer gency filter units, and the bypass dampers that direct fresh air through the emergency filter units ar e division oriented electrically.

The remote air intake valves ar e manual, and the design is simp le and reliable enough that the valve may be manually repositioned in a short period of time.

In the event of a LOCA, the main control room is prot ected from potential airborne radioactivity by pressurizing the control room with air supplied via the remote air intakes. The system operates in the following manner:

The isolation valves in the normal fresh air intake are closed by any of the F, A, or Z signals.

These are the same sign als provided for isolating the prim ary and secondary containments.

Both sets of remote air intake isolation valves are normally locked open. Each remote air intake is monitored by its divisional radiation monitor which sens es the radiation level at the remote air intake header. When radiation levels exceed the tr ip setpoint, an alarm annunciates in the main control room to indicate the condition. The cont rol room operator may manually isolate the alarming remote air intake. At leas t one air intake must be maintained open to ensure the control room is pressurized.

Each remote intake h eader is provided with a purge exha ust system to provide continuous radiation monitoring of the remo te intakes while isolation valves are closed. The two purge exhaust systems each consists of two isolation valves in series. One purge valve is equipped with an electrohydraulic operator and is interlocked with its remote intake valve, and the other purge valve is ma intained open.

Both purge exhaust systems utilize battery exhaust fans WEA-FN-53A and WEA-FN-53B (which are both ESF) to purge air from the remote intake headers.

Both battery exhaust fans provide the necessary redundanc y through cross over ductwork between each set of purge exhaust valves.

The emergency filter units are energized by F, A, Z signals and all outdoor pressurizing air is automatically diverted through the filter units. The main contro l room kitchen exhaust fan and its isolation damper are also shut off by F, A, Z signals.

Operating in the above manner ensures that the ma in control room is c ontinuously pressurized with filtered air. The details of the control room dose analysis are discussed in Sections 15.4.9 , 15.6.4 , 15.6.5 and 15.7.4.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.4-7 9.4.1.2.2 Cable Spreading Room

The cable spreading room HVAC system cons ists of two 100%-capac ity air-handling units, each with its own duct distribution system, co mmon distribution system inside the cable spreading room, and one purge exhaust fan. The air handli ng units are similar to those servicing the main control room, i.e., filter, SW coil, chilled water coil, electric heater, and centrifugal fan in the sheet metal housing. Normally one air handling unit operates continuously on a 100% recirculation mode of operation maintaining the cable spreading room and remote shut down room at approximately 80°F.

The cable spreading room purge exhaust fan does not normally operate. In the unlikely event of fire developing in the cable spreading room, the purge fan can be manually operated to remove any smoke from the cable spr eading room prior to personnel access.

If the radwaste chilled water suppl y to the cable spreading room air handling units is lost, SW is supplied to the units for emergency cooli ng. Under this mode of operation the cable spreading room and remote shut down room temperature is limited below equipment operability limits. See Table 3.11-1.

9.4.1.2.3 Critical Switchgear Area

The switchgear and batt eries associated with the redundant emergency electric power systems are located in separate equipment rooms below the cable spreading room. A separate heating and ventilation system is provi ded for each set of equipment rooms. Ventilation of the emergency chiller area at the 525 ft level is also provided by this system. Each of the two heating and ventilating systems consists of an air handling un it, battery room exhaust fans and associated ductwork and controls. The air handling unit consists of a roughing filter, two water coils in series (one for WCH or plant service water, one for SW), an electric blast coil heater, and a centrif ugal fan in a sheet metal housing. The two air handling units have different capacities due to heat load differences between the tw o sets of rooms. Electric heaters are provided in the ducts supplying air to the battery room s to maintain temperature in those rooms above 60°F.

Both heating and ventilating systems normally operate continuously during all modes of operation. They are both partial recirculation systems with fr esh air provided as makeup for the air exhausted from the battery rooms. Th e battery rooms are continuously exhausted (no recirculation) to prevent the possible buildup of combustible ga ses generated by the batteries. During normal operation, either WCH or plant service water is provided to the air handling units as the cooling medium.

Under all emergency modes of operation SW is provided to both units as the cooling medium. The critical switchgear area air handling units also provide the normal and emergency ventilation for the HVAC equipment rooms and th e emergency chiller area at the 525 ft level.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.4-8 Temperatures in the HVAC equi pment rooms are limited to a range of 55°F to 104°F during normal operation and below equipment operability limits during all emergency modes of operation. See Table 3.11-1. 9.4.1.3 Safety Evaluation

9.4.1.3.1 Main Control Room

The reliability of the HVAC sy stems serving the main control room is achieved by providing two 100% redundant systems. Only one of the two redundant syst ems is required to operate to provide adequate dose mitigati on. The two systems are physically separated to preclude simultaneous failure from any one incident. Th e emergency chillers are located in a general area. All components of the two systems, except the radwas te chilled water system, are designed to withstand the effects of an SSE and both are powered from emergency diesel buses. The CCH or SW is used as the cooli ng medium in the event that radwaste building chilled water is unavailable, thus providing acceptable temperatures in the control room under all modes of operation.

The normal fresh air intake is provided with two division oriented valv es (normally open, fail closed) in series to close in the event of an F, A, Z signal. Each remote air intake header is provided with redundant ra diation monitors to alarm in the even t of high radiati on. The valves are a highly reliable butterfly-type w ith the disc keyed to the pivot shaft. If one of the remote air intake isolation valves s hould fail it may be easily repositioned or repaired. One remote intake will always remain open to ensure a pressurized cont rol room and prevent infiltration.

The remote fresh air intakes are used to pressurize the main control room through emergency filter units. This limits infiltration of airborne radioactiv e contaminants and smoke due to a fire within the plant but external to the control room. Infiltration of airborne radioactivity in the main control room is discussed in Section 6.4.

The emergency filter unit starts operating in the event of a LOCA.

The main control room is maintained at 75 +/-3°F dry-bulb temperature under normal conditions. In the event of an emergency, the control room can be maintained below 85°F by the CCH or SW can be used to maintain less than 104°F.

9.4.1.3.2 Cable Spreading Room

The cable spreading room is provided with two 100%-capacity HVAC systems which are physically separated. All components of the tw o systems, excep t the chilled wa ter system, are designed to operate through the SSE and are power ed from emergency diesel buses. As with the control room HVAC system, SW is used as the cooling water for th e cable spreading room

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.4-9 air handling units to maintain an acceptable temperature in the cable spreading room for equipment operation in the event that th e chilled water syst em is inoperable.

9.4.1.3.3 Critical Switchgear Area

The essential electric equipment for each of the redundant emer gency diesel ge nerators is serviced by separate heating and cooling systems. Each system is powered from a Class 1E bus which is supplied by the dies el generator it serves and is designed to operate through an SSE. Standby service water is used as the system cooling medium whenever WCH or plant service water are not available, thus ensuring cooli ng during all modes of plant operation. The two systems are physically separated and arranged in such a manner that the failure of one system can affect only the dies el generator that it services.

9.4.1.4 Testing and In spection Requirements

The performance of the HVAC systems servicing the main control room, cable spreading room, and critical switchgear areas can be verified while the systems are operating. The operability and performance of standby equipment is determined by alternating the duty of redundant systems.

The control room system ductwork was subject to leak tests during erection and was balanced for air flows in accordance with the procedures of the Associat ed Air Balance Control Council (AABC). All system components were subject to preoperati onal testing. All piping systems components were subject to hydro static tests during erection.

The emergency filter housings and filters were subject to both shop and field efficiency tests.

The HEPA and charcoal adsorber filters are periodically te sted as required by the Technical Specifications. Charcoal samples laboratory test results are required within 31 days of removal.

9.4.1.5 Instrumentation Requirements All essential controls for the control room, cable spreading room, and critical switchgear area HVAC systems are electric or electronic except the remote ai r intake isolation valves.

Pneumatic controls are used onl y on nonessential components.

9.4.1.5.1 Main Control Room

The following controls are provided in the main control room in addition to those discussed in Section 9.4.1.2.1.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.4-10 Both air handling systems serving the main contro l room can be started fr om separate selector switches located in the main c ontrol room. When an air handling unit fan is started, all controls associated with that system are energized via electrical interlocks initiating the following operations:

a. The fresh air intake damper is opened, and

b. The electronic thermostat in the main control room modulate s the chilled water valve and energizes the two stages of electric heaters, as required, to satisfy the heating/cooling requirement s of the main control r oom. (The heater breakers are normally locked open in modes 1, 2, and 3.)

In the emergency condition (loss of radwaste building chilled wate r during design basis accident), the cooling coil WMA-CC-51A1 serving air handling unit WMA-AH-51A is supplied with SW. When the Off-Auto control switch in the control r oom, which is normally in the Off position, is set to Auto, the coo ling coil WMA-CC-51B1 will be automatically supplied with emergency chilled water. If necessary, cooling coil WMA-CC-51A1 can be supplied with CCH by manual openi ng or closing of valves in SW and CCH lines to chiller CCH-CR-1A. Also, if necessary, cooling coil WMA-CC-51B1 (whi ch is normally lined up for CCH) can be supplied by SW by manually opening or closing of the appropriate valves in SW

and CCH lines to chiller CCH-CR-1B.

Control switches are provided in the main control room for all fans, local air intake isolation valves, and dampers so that all components can be cont rolled manually as well as automatically. The remote air intake isolation valves are manual valves.

9.4.1.5.2 Cable Spreading Room

The air handling units serving the cable spreadi ng room are started from separate selector switches located in the main c ontrol room. When an air handling unit fan is started, an associated solenoid valve is energized permitting the air hand ling unit pneumatic chilled water control valve and electric heating coils to receive a pneumatic control signal from a temperature controller sensing temperature in the air return duct from the cable spreading room (the heater breakers are nor mally locked open in modes 1, 2, and 3). The starting of the fan also energizes the air handling automatic roll filter contro l circuit permitting the filter drive motor to change media at sele cted preset timer intervals.

Redundant temperature switches located in the cable spreading room will annunciate alarms in the main control room in the event of a temperature rise to 90°F thus alerting the operator of a possible equipment malfunction. A differential pressure switch across the air handling unit filter will alarm in the

event of high differential pressure.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.4-11 Smoke detectors in the cable spreading room return air ducts annunciate alarms in the main control room in the event of smoke so that the operator can activate the fire protection system (see Appendix F

). There are no control valves associated with the SW coils in the cable sp reading room units. Whenever SW is on, there is full water flow through the coil. During normal operation any heat added to the space by the SW coil is co mpensated for by the air handling units chilled water coil.

9.4.1.5.3 Critical Switchgear Area

The two air handling units and the two battery room exhaust fans whic h service the critical switchgear area each have their own selector switch located in the main control room. When an air handling unit fan is starte d, an associated solenoid valv e is energized permitting the air handling unit pneumatic plant service water valve and electric heating coils to receive a pneumatic control signal from a temperature c ontroller sensing temperature in the supply air duct to the critical switchgear area. The temperature contro ller is set to maintain the temperature as described in Table 3.11-1. The starting of the fan also energizes the air handling unit automatic ro ll filter control circuit, permitting the filter drive motor to change media at selected preset timer intervals.

Temperature switches located in each of the electrical equipment rooms serviced by the critical switchgear system annunciate alar ms in the main control room in the event of abnormally high temperatures thus alerting the operator of a po ssible malfunction of the cooling system. Smoke detectors in the main return air ducts to the air handling unit and in th e battery room exhaust ducts will annunciate alarms in the main control room in the ev ent of fire in the switchgear area. Differential pressure switches across each of the battery room exha ust fans and across the filter of the air handling un its will annunciate alarms in the main control room in the event of low differential pressure across fan or hi gh/low differential pressure across filter.

As with the cable spreading room air handling units, there are no control valves associated with the SW coils in the switchg ear area units. Whenever SW is on, there is full water flow through the coil.

9.4.2 REACTOR

BUILDING 9.4.2.1 Design Bases

The reactor building, or secondary containment, is provided w ith a HVAC system designed to meet the requirements for all ge neral areas of the building, the spent fuel pool, potentially contaminated areas and the primary purge as follows.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.4-12 a. To provide fresh, temper ed, ventilating air to the various spaces in the reactor building in sufficient quantity to lim it the temperature during all normal operating conditions and/or during lo ss of ac power as specified in Table 3.11-1 , while providing at least one air change per hour in all areas;

b. To draw air across the surface of the spent fuel pool, reactor well, and dryer/separator pool by exhausting around their pe rimeters to control temperature and to maintain airborne radioactivity within acceptable levels in these spaces;

c. To provide for controlled air movement from areas of potentially low radiation to areas of potentially high radiation. This serves to isolate and segregate airborne radioactivity which may be released due to equipment failure or malfunction. The system is designed to preclude recirculation of reactor building air under normal operating conditions;

d. To maintain the reactor building during normal opera tion at a negative pressure of 0.25 in. w.g. with respect to atmosphere as indicated at the reactor building el. 572 ft to minimize release of air borne radioactive material. During emergency operation, the standby gas tr eatment (SGT) system (see Section 6.5) maintains the reactor buildi ng at a negative pressure;

e. Maintain the reactor building equipmen t drain sumps and h eaders at a negative pressure with respect to the rest of th e reactor building, and provide means to filter the effluent before discharging it to the main reactor building exhaust header so that the release of radioactive contaminants is minimized;

f. Provide means to monitor all effluent from the reactor building prior to release for radioactive contamination and to is olate all ventilation openings in the building in the event that radiation levels so monitored exceed the limits defined in 10 CFR Part 20; and

g. In addition, the reactor building ventilation system is designed to provide for purging of the primary containment. Pr imary containment purge air is supplied and exhausted by the reactor building ventilation system.

Except for the emergency cooling system which is described in Section 9.4.9 , the only portions of the reactor building heating and ventilating system which are designed as ESF are the isolation valves located in the outdoor air intake duct and the exha ust system discharge duct, and the isolation valves on the purge connections to th e primary containment. The reactor building isolation valves are designe d, manufactured, and N-stamped Class 3 in accordance with Section III of the ASME Boiler and Pressure Vessel (B&PV) Code. The primary containment purge isolation valves are N-stamped Class 2 in accordance with

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-028 9.4-13 Section III of the ASME B&PV Code. All is olation valves are Se ismic Category I. The means of protecting the system vents and louvers from missiles is discussed in Section 3.5.

9.4.2.2 System Description

The reactor building heating and ventilating system is shown in Figure 9.4-2. The system is basically a "push-pull" heating and ventilation system providing once-through air flow with no recirculation and consists of the following systems:

a. Supply air,

b. Exhaust air,

c. Sump vent exhaust filter,

d. Emergency cooling (described in Section 9.4.9), and e. Miscellaneous ventilation.

Equipment details are given in Table 9.4-2. Equipment seismic information is given in Table 3.2-1.

9.4.2.2.1 Supply Air System

The supply air consists of a heating and ventilation unit, air distribution ductwork, two isolation butterfly valves on the fresh air intake, and associated cont rols. The heating and ventilation unit is composed of the follo wing elements in an insulated housing:

a. An automatic roll type prefilter,

b. Steam coils with fa ce and bypass dampers,

c. A capillary type, evaporative cooling section with two 100%-capacity spray pumps, and

d. Two 100%-capacity vaneaxial fans with back draft dampers on the fan discharges.

During normal plant operation and shutdown, the supply air syst em isolation valves are open and the ventilation system operates continuously distributi ng tempered, 100% outdoor air throughout the building. One supply fan is nor mally operating with the second fan in standby.

The standby fan starts automatic ally in the event the operating fan fails. During winter months, the steam coil in the ventilating unit heats the supply air. A second steam booster heater in the supply duct to the refueling floor he ats the air to maintain th e refueling floor area. (See Table 3.11-1 for temperature limitations.) During summer months , the capillary type air washer cools the incoming air by evaporative coo ling. Water for this air washer is supplied from the potable water system (see Section 9.2.4).

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.4-14 The reactor building supply air system also provides makeup air to the primary containment during primary containment purge if needed. Du ring purge, supply air is directed into the primary containment through the isolation valves.

The purge supply air fl ow rate is controlled by an air operated directing da mper in the supply system ductw ork which is set manually from the main control room. A b ackdraft damper is provided in this purge line and is counterweighted so that the pressure drop acro ss the backdraft damper in the open position, does not exceed 0.25 in. w.g.

9.4.2.2.2 Exhaust Air System

The reactor building exhaust system draws air from all ar eas with potential radiation contamination and discharges it to the elevated release point. There are two 100%-capacity vaneaxial fans in the system. One fan normally operates wi th the second fan in standby.

Approximately one-half of the reac tor building exhaust air is draw n from the refueling level of the reactor building. Intake ducts to the e xhaust system are embedded around the periphery of the spent fuel pool, dryer/separator pool, and reactor well to remove any potentially radioactive vapors generated in the pools. Volume control da mpers in the ducts from each pool are controlled from a local panel in the reac tor building to permit the operator to vary the flow over each pool during vari ous fuel handling operations.

Primary containment purge can be performed by discharging the purge air through the reactor building exhaust system or through the SGT system (see Section 6.5). Ducts connect the primary containment drywell and suppression pool area with both systems. The reactor building exhaust system is normally used for purge after the SGT system is used for the first 24 hr. The SGT system is also used when an unacceptably high le vel of airborne radioactivity is present inside the primary containment.

Radiation monitors are located ju st outside of the air plenum on the intake side of the two reactor building exhaust fans. In the event that a preset high level limit of radioactivity is exceeded, the radiation monitors will annunciate an alarm in the main control room and transmit an isolation signal to the affected emergency safeguards systems.

In the event of an F, A, or Z signal, the reactor building exhaust fans stop and the isolation valves close. All isolation signals which stop and isolate the reactor building exhaust system also start the SGT system (see Section 6.5.1).

9.4.2.2.3 Sump Vent E xhaust Filter System

All potentially radioactive leaks and/or spills in the reactor building are channeled to the equipment and/or floor drain systems. To minimize the release of radioactive contaminants from the building, the drain system sumps a nd drain headers are maintained at a negative pressure and vented through a filter system. The sump vent e xhaust system is composed of

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.4-15 two, full-capacity, filt er units which draw air from the sumps and drain headers and pass it through filters, a moisture separator, and electric heater before discharge to the main reactor building exhaust system upstream of the radiation monitors.

Each filter unit consists of the following elements in a sheet metal housing:

a. A moisture separator to remove entrained moisture,

b. An electric blast coil heat er to limit the relative humidity of the air entering the filter to 70%,

c. A medium efficiency particulate prefilter,

d. A HEPA filter,

e. A tray type, 2-in.-thick activated charcoal filter, and

f. A centrifugal fan.

All filters in the sump vent exhaust units are designed to satisfy the same efficiency requirements as those of the control room emergency filter units outlined in Section 9.4.1.2.1. One filter unit operates continuously during normal plant operation. The standby unit is started from the main control room in the event that the fan of the operating unit fails. Neither unit operates in the event of a reactor building isolat ion signal. Deluge fire protection is provided for each filter unit (see Appendix F

).

9.4.2.2.4 Miscellaneous Area Ventilation Systems

The following miscellaneous HVAC systems are provided to service local areas in the reactor building:

a. The main steam tunnel is serviced by three fan coil units located within the tunnel. Three fan coil units are compos ed of a water cooling coil supplied by the plant service water system and a dir ect drive centrifugal fan in a sheet metal housing. Two fan coil units normally opera te continuously in the recirculation mode to remove heat generated by the st eam piping in the tunnel. Two fan coil units operating has sufficien t capacity to limit the temp erature in the tunnel to 130°F with one fan coil unit in standby.

b. The reactor water cleanup (RWCU) sample hood is provided with an exhaust filter unit which is operated whenever a sample is taken. The hood is of the air

curtain type with supply air to the hood provided from the reactor building supply air system. An electric duct heater is provided in the supply air duct to

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.4-16 temper the air into the hood to 70°F. Th e exhaust filter unit, which draws air from the hood, is composed of a medium efficiency prefilt er, a HEPA filter, and a centrifugal fan in a sheet metal housing. The exhaust fan discharges air into the reactor building exhaust system upstream of the radiation monitors.

c. The vehicle air lock (railroad bay), on grade elevation, is heated and cooled by an air handling unit which operates in a 100% recirculation mode. The air handling unit is composed of a prefilter, water cooling coil, and centrifugal fan in a sheet metal housing. A 15-kW electr ic blast coil heater is provided in the unit discharge duct for heating. Plant se rvice water is supplie d to the water coil for cooling.

During periods when the large railroad lock door (R106) is open for operations associated with the Independent Spent Fu el Storage Installation (ISFSI), or other activities requiring the door to be open for an extended period, engineered features are available to maintain the railroad lock at temperate conditions. Eight thermostatically controlled heaters are installed near the door, and an air curtain is installed above the door.

These supplemental h eaters, used in conjunction with the air curtain, keep the railroad lock warm during winter conditions when the door is open.

9.4.2.3 Safety Evaluation

The following safety features are incorporated in the design of the reactor building ventilation system:

a. The air distribution within the building provides for cont rolled air movement from areas of potentially low radiation to areas of potentiall y high radiation. Under normal operating conditions, reci rculation of contaminated air is precluded by system design.

b. There are only two penetrations of the reactor building associated with the building ventilation system. Two butterfly valves in series are provided in the duct connections to each of these ventila tion openings. The four valves close, isolating the reactor building, and the ven tilation system fans stop in the event of any of the F, A, Z isolation signals.

The redundant, fail-closed valves on each ventilation opening ensure reactor building isolation in the event of failure of any one valve.

Two fail-closed, air-operate d isolation valves, in seri es, are also provided on all purge connections to the primary cont ainment. These valves are open only

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.4-17 during primary containment purge operations and clos e automatically in the event of any of the a bove isolation signals.

c. The reactor building is ma intained at a negative pressu re of 0.25 in. w.g. with respect to atmosphere by differential pressure c ontrollers which position the blade settings of the exhaust fans.

This atmospheric control limits the exfiltration of radioactive contaminants from secondary containment. See Section 6.5.1 for emergency operation of the SGT system whic h maintains the reactor building at a negative pressure under postulated accident conditions.

Each set of main supply and exhaust fans are powered from different divisions of the Class 1E power supply.

d. Radiation monitors, located just outsi de of the exhaust air discharge plenum, monitor exhaust air discharged from the building and transmit an isolation signal in the event that the radiation level exceeds a preset level. See Chapter 11 for further discussion of radiation monitoring.

e. The equipment drain system sumps and headers are vented through two 100%-capacity filter units. This operation minimizes the release of radioactive material during normal plant operation. Two 100%-capacity redundant filter trains ensure s continuous system operation in the event of any single component failure.

f. Each ventilation head er with embedded exhaust duct connections around the spent fuel pool, reactor well and dryer/

separator pool is equipped with a flow transmitter and flow indicator to ensure that the designed quantity of air is being exhausted for proper airborne contamination control. See Figure 9.4-2.

9.4.2.4 Testing and In spection Requirements

The performance of the HVAC systems servicing the reactor building can be verified while the systems are operating. The operability of standby equipment is determined by rotating the duty of the redundant components.

The performance of those portions of the exhaust system which serves the fuel pool can be verified by measuring air flow in the fuel pool exhaust system exhaust header. A damper located in the fuel pool exhaust header controls this air flow.

The flow indicator and damper controller are mounted on a local panel in the reactor building.

All system ductwork was balanced for air flows in accordance with proce dures of the AABC.

All system components were subject to preoperational testing to verify that the system

functions in accordance with the design require ments. All piping systems and components were subject to hydrostatic and/or pressure tests during erection. Primary containment purge

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.4-18 system isolation valves are normally closed and are leak tested in accordance with the Technical Specifications.

The sump vent exhaust system filter housings and filters were subject to both shop and field efficiency tests. The HEPA filters are given in-place DOP tests in accordance with ANSI N510-1980, "Testing of Nucl ear Air Cleaning System." Filt ers are tested at rated flow with an acceptance limit of less than 0.05% penetration at rated flow. Charcoal adsorber filters are leak tested in accord ance with ANSI N510-1980 to en sure that filter bypass is less than 0.05% of full flow.

9.4.2.5 Instrumentation Requirements

The following are the major instrumentation devi ces used for controlling the reactor building heating and ventilation system:

a. Isolation valve control

The isolation valves on the purge lines to the primary containment and the reactor building supply air intake and exhaust air disc harge ducts are controlled by individual OPEN-AUTO-CLOSE selector switches located in the main control room. The switch springs back to AUTO from either the OPEN or CLOSE position. With the selector switc h in the AUTO position, an isolation signal will deenergize the solenoid and clos e the isolation valve. The isolation valves are designed to fail in the closed position in the event of loss of control air.

b. Supply ventilati on unit fan control

Selector switches, with indicating lights, are provided in the main control room for the two supply ventilation unit fans.

A differential pressure switch, with probes on either side of the fan/damper assembly, will automatically start the standby fan in the event of low differential pressure which is indicative of fan failure. The differential pr essure switch will also annunciate an alarm in the main control room.

In the event of an isolation signal, both supply fans are stopped automatically via electric interlocks.

c. Main exhaust system fans

The two main exhaust fans are controlled by selector switches located in the main control room. When one fan is started, an associated solenoid valve is

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.4-19 energized causing the pneumatic, fail-clo sed damper on the fan discharge to receive control air and open. A differential pressure switch, with probes on either side of the fan/damper assembly, will start the standby fan in the event of low differential pressure which is indicative of fan or damper failure.

The reactor building is main tained at a negative pressu re of 0.25 in. w.g. with respect to atmosphere by au tomatically varying the pitc h of the main exhaust fan blades, thereby changing fan capacity, via a pneumatic controller. There is an electronic differential pressure sensor across each of the reactor building's four exterior walls. The lowest of the four signals is transmitted to a differential pressure controller which transmits a signal, converted to a pne umatic signal, to the pneumatic fan blade controller to maintain the lowest differential pressure at the selected setpoint.

In the event of an isolation signal, bot h exhaust fans are st opped via electric interlocks.

d. Primary containment purge control

When the primary containment is purged to the reactor building exhaust system, the following operations are performed:

The primary containment purge exhaus t isolation valves and the valve connecting the purge exhaust header to the reactor building exhaust system are opened via individual selector switches located in the main control room.

A pneumatic volume damper in the pur ge line to the exhaust system is controlled by an adjustable differential pressure controller. The pressure is sensed between the ductwork connecti ng the primary containment and the reactor building. The differential pressure controller is located in the main control room. The primary containment is depressurized, if required, by slowly lowering the setpoint of the differentia l pressure controller thus gradually opening the damper. Flow indication of purge exhaust is provided in the main control room so that depressurization can be performed at a controlled rate.

Once the containment is depr essurized, the isolation valves on the supply purge connection to the containment are opened via selector switches in the main control room. Pneumatic da mpers in the supply system , which divert supply air into the purge supply header, are used fo r purge rate control. When the valve in the supply line to the supply purge head er is opened, an associated solenoid valve is energized permitting the diverting damper to receive a control signal. The control signal is transmitted from an adjustable flow controller located in the main control room, which will modulat e the damper as required to maintain C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.4-20 the purge supply rate at the controller setpoint. Purge flow is adjustable up to 10,500 cfm.

Normally the primary containment dryw ell and suppression pool area are purged separately by opening the a ppropriate isolation valves. In the event of an isolation signal during purge, all isolation valves are automatically closed.

e. Sump vent exhaust system control Selector switches with running lights are provided for each sump vent exhaust unit. When a unit fan is started, the control circuit for the unit is energized causing the electric blast coil heater to start and a damper cont rol solenoid valve to energize. When ener gized, the solenoid valve pe rmits the pneumatic volume damper on the unit fan discharge to receive a control signal from a differential pressure controller set to maintain a flow of 1000 cfm through the unit.

The drain system main exhaust header from which the filter unit draws is maintained at a negative pre ssure of approximately 1.0 in. w.g. with respect to the reactor building atmosphere. A volume control damper, located between the exhaust header and reacto r building atmosphere, is modulated by a differential pressure controller set to mainta in that pressure differential.

In the event that the opera ting sump exhaust unit fails, a differential pressure switch with probes set across the unit filt ers will annunciate an alarm in the main control room, thus alerting the operator to start the standby unit. The differential pressure switch also deenergizes the units electric heater.

A thermocouple on the charco al filter will annunciate an alarm in the main control room on a temperature rise to 250°F.

f. Heating control

Supply air is heated in the supply ven tilation unit by a steam coil with pneumatic face and bypass dampers. The source of the heating steam (HS) is from the gland steam evaporator during turbin e generator operation and from the auxiliary boiler during shut down. The dampers are controlled by a temperature controller which senses s upply air temperature downstr eam of the fans. The setpoint of the temperature controller is adjustable. A temperature switch with a sensor in the outdoor air intake energi zes a solenoid valve, thus causing the steam shutoff valve in the feed line to the steam coil to open whenever the outdoor temperature drops below the setpoint.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.4-21 The steam booster heater in the supply duct to the refue ling floor is controlled in a similar manner to that of the main ventilation unit and is supplied HS from the same sources that supply the main un it. Face and bypass dampers on the booster heater are modulated by an ad justable, area mounted, temperature controller.

g. Evaporative cooler control

The supply ventilation unit evaporative cooler (air washer) section is controlled by temperature switches with sensors located in the outdoor air intake.

9.4.3 RADWASTE

BUILDING

9.4.3.1 Design Bases

The radwaste building houses the solid, liquid, and gaseous radwaste treatment systems. The HVAC systems serving the radwaste build ing are designed to satisfy the following requirements:

a. To provide fresh tempered ventilation ai r to the various spaces in the radwaste building in sufficient quantity to limit the temperature as specified in Table 3.11-1 , while providing at least three air changes per hour in potentially contaminated areas. Additionally, areas in which personnel will spend extended periods of time performing sensitive operations are air conditioned for personnel comfort;

b. The air distribution within the building provides for cont rolled air movement from areas of low radiation contamina tion potential to area s of progressively higher radiation cont amination potential;

c. To filter the exhaust air from the building before discharging it to the atmosphere to limit the release of airborne radioactive particulates;

d. To heat the supply air as required during the winter season to maintain minimum temperatures as specified in Table 3.11-1

and e. To maintain the lower level of the radw aste building at a negative pressure with respect to atmosphere to minimize the release of radioactivity.

9.4.3.2 System Description

The radwaste building HVAC system is shown in Figure 9.4-3. The main system is a push-pull heating and ventilating sy stem providing once-through air fl ow with no recirculation.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-028 9.4-22 In addition, individual air conditioning units are provided for all rooms which personnel will normally occupy for extended periods of tim

e. Equipment details are given in Table 9.4-3. Equipment seismic information is given in Table 3.2-1.

The main radwaste building supply air system consists of a suppl y ventilation unit and distribution ductwork. The suppl y ventilation unit consists of an insulated steel cabinet housing a prefilter, steam coil wi th face and bypass dampers, an eva porative cooling section, and two 100%-capacity centrifugal fans.

The prefilter is of the rene wable roll type, automatically progressed to maintain uniform pressure drop. The steam coil is of the nonfreeze type with automatic face and bypass dampers fo r temperature control. The ev aporative cooling section is of the capillary type with two full-capacity pum ps with common suction from the washer basin and common discharge into the sp ray header system. Makeup water to the evaporative cooler is supplied from the plant potable water system.

During normal plant operation the supply unit operates continuously, providing tempered air throughout the building via the supply duct distribution system. The radwaste building supply system has no safety functions and will be inoperative in the event of loss of offsite power.

The radwaste building exhaust system is composed of three 50%-capacity exhaust filter units.

Each exhaust unit consists of medium efficiency prefilters, HEPA filters, and a centrifugal fan in a sheet metal housing. The prefilters are of the cartridge type with a minimum ASHRAE Dust Spot Efficiency of 80%-85%. The HEPA filters are designed for removal of fine airborne particulates. These filters are of wa ter repellent and fire-resistant construction for operation at temperatures up to 300°F. The HEPA filter has a minimum efficiency of 99.97% on a 0.3 m DOP test. The HEPA filters are fabricated in accordance with MIL-F-51068C, MIL-STD-282, and carry a UL label certifying compliance with UL586.

Each exhaust unit fan is provided with an automatic air-operated inlet vane for flow control.

The inlet vanes are controlled by differential pressure controlle rs set to maintain the tank enclosures in the lower level of the radwaste building at a negative pressure w ith respect to atmosphere.

During normal plant operation two exhaust units operate, with the third unit in standby, drawing air from a common exhaust duct system and discharging the ai r above the radwaste building roof. All radwaste building exhaus t air is processed by the exhaust units and monitored by radiation detectors prior to discharge. Radioactiv e releases to the environment are discussed in Section 11.3.3. The radiation monitoring system is discussed in Section 11.5.2.2.1.7 with arrangement details shown in Figure 11.5-7. The exhaust duct system is arranged in such a manner that all exhaust is drawn from areas with the highest radioactive contamination potentia l thus inducing air flow from clean areas into the potentially contaminated zones. Where applicable, automatic volume dampers are provided in ventilation openings in walls betwee n clean and potentially contaminated areas. The volume dampers are

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.4-23 controlled by differential pressure controllers, with probes on eith er side of the vent opening, to maintain "hot zones" at a negative pressure with respect to "clean zones."

The radwaste control room, inst rument shop, radioche mical laboratory, sample room, counting room, and health physics area are provided with air conditioning systems to maintain suitable conditions for personnel comfort. All areas except the radwaste contro l room are serviced by once-through air conditioning systems. Th e radwaste control room, which has no contamination potential, has a partial recirculation air conditioning system.

The air conditioning for each area is provided by air handling units which draw air from the radwaste building supply air system, pass it thr ough filters, chilled wate r coils, and electric heaters, and discharge it into the rooms being conditioned.

Chilled water is supplied to the coils from the WCH.

The conditioned air leaves the rooms by exfiltration to the rest of the building or, as in case of radiochemistry laborator y, hot instrument shop and sample room, through exhaust hoods which discharge into the main exhaus t system. The exhaust hoods in the radiochemistry laboratory, hot instrument shop and sample room are provided with local filter units composed of prefilters, HEPA filters, and fans to remove radioactive particulates at the hoods prior to discharge into the exhaust system. These HEPA f ilters are similar to those in the main exhaust filter units.

Effluent from the radwaste building ventilation is continuously monitored for gaseous activity and continuously sampled for laboratory analysis of particulate radioactivity. The sampling and monitoring system is described in Section 11.5.2.2.3.

9.4.3.3 Safety Evaluation

The radwaste building HVAC system is not required for a safe shutdown of the plant; however, the following features ar e incorporated in its design to ensure system reliability and to minimize the uncontrolled release of airborne radioactive contaminants during normal plant operation:

a. The 50% standby capacity of the exhaus t filter system en sures full system capacity with any unit inoperative due to equipment failure or maintenance outage,

b. Redundant supply fans ensure supply system operation in the event of single fan failure,

c. All exhaust air is passed through HEPA filters prior to discharge thus minimizing the release of radioactive contaminants, and

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.4-24 d. The potentially contaminated area is maintained at a negative pressure and all air is monitored by radiation detectors prior to discharge to ensure against release of radioactive contaminants.

9.4.3.4 Testing and In spection Requirements

The performance of the HVAC systems servicing the radwaste bu ilding can be verified while the systems are operating.

The operability and performan ce of standby equipment is determined by rotating the duty of redundant components. Pre ssure, temperature, and flow instrumentation is provided as shown in Figure 9.4-3.

The HEPA filters in the radwaste building exhaust filter units ar e subject to both shop and field efficiency tests. On installation, and periodically thereafter, HEPA filters are given in-place DOP tests in accordance with AN SI N510-1980, "Testing of Nuclear Air Cleaning System."

Filters are field tested at rated flow with an acceptance limit of less than 0.05% penetration.

All system ductwork was balanced for air flows in accordance with th e procedures of the AABC. All system components were subject to preoperational testi ng to verify that the system functions in accordance with the design requirements.

9.4.4 RADWASTE

BUILDING CHILLED WATER SYSTEM

9.4.4.1 Design Bases

The WCH is designed to provide a reliable source of chilled water to the main control room air handling units, the cable spreading room air ha ndling units, the switc hgear area air handling units, and the air handling units serving the conditioned areas of the radwaste building (see Sections 9.4.1 and 9.4.3).

The temperature of the chilled water supplied to the air handling units is maintained between 44°F and 55

°.

9.4.4.2 System Description

The WCH is shown in Figure 9.4-4. It is a closed loop system incorporating two 100%-capacity pumps and two 100%-capacity centrifugal water chille rs. The chillers are arranged to operate independently with bypass and shuto ff valves provided for ease of maintenance. During norm al operation, either chiller (or both) operates to main tain the chilled water supplied to the air handling units between 44°F and 55°F. The two chilled water pumps are arranged in parallel with one normally operating and one in standby. The liquid chillers are each rated at 150 tons refr igeration capacity. Ea ch consists of a hermetic centrifugal compressor with linear capaci ty control from 10% to 100%

of rated unit capacity, and evaporator, water-cooled condens er, purge system, and automatic control center. Condenser C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 9.4-25 cooling water is supplied to the chillers from the plant service water system (see Section 9.2.1).

An air separator, expansion tank, chemical feed tank, and appropriate valves and piping

connections are provided at the common suction to the two pumps for system control. Makeup water to the system is supplied from the plant potable water system (see Section 9.2.4).

9.4.4.3 Safety Evaluation The WCH has no safety function; however, redundant circulating pumps a nd liquid chillers are incorporated in the design to ensure uninterrupted system operation during normal plant operation.

All components of the chilled water system are designed to Se ismic Category II requirements.

9.4.4.4 Testing and In spection Requirements

The performance of the WCH can be verified while the system is operating. The operability of standby equipment is determined by rotating the duty of redunda nt components. Pressure, temperature, and flow instrumentation is provided as shown in Figure 9.4-4.

Chillers, pumps, tanks, and piping we re subject to hydrostatic test s in the field after erection.

9.4.4.5 Instrumentation Requirements

Each of the two chilled water pum ps is controlled by a locally m ounted selector switch. When either pump is energized, the chiller control circu it is energized via electrical interlocks. If the operating chilled water pump fails, a pressure switch on the pum p discharge will start the standby pump after a delay and an alarm will be actuated in the main control room.

Each chiller has a temperature sensing element in the chilled wa ter supply line which controls the chiller capacity by modulating the guide vane position to ma intain the required temperature.

A single trouble alarm is actuated in the contro l room to alert operators upon chilled water low temperature, high condenser temperature, or high motor temperature.

9.4.5 OFFGAS

CHARCOAL ADSORBER VAULT REFRIGERATION SYSTEM

The offgas charcoal adsorber vault refrigeration system has been permanently deactivated in-place.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 9.4-26 9.4.6 TURBINE GENERATOR BUILDING

9.4.6.1 Design Bases

The turbine generator building is provided w ith HVAC systems desi gned to satisfy the following requirements.

a. Provide fresh, tempered, filtered ven tilating air to the vari ous spaces within the turbine generator building in sufficient quantity to limit the temperatures as specified in Table 3.11-1

b. Provide for controlled air movement from areas of potentially lower airborne radiation contamination to areas of progressively higher airborne radiation contamination potential. Th is distribution serves to limit airborne contaminants from migrating from potentially contam inated areas into clean areas. The ventilation systems operate on a once-through basis without recirculation;

c. Monitor exhaust air from the building for radioactive contaminants, prior to discharge, to ensure that the release of contaminants does not exceed the concentration limits defined in 10 CFR 20;

d. Maintain the turbine generator build ing potentially radioactive areas at a negative pressure with respect to atmosphere to minimize the release of radioactive contaminants;

e. Automatically provide combustion air for the auxiliary bo iler in the turbine generator building;

f. Provide ventilation air to the makeup water pump transformers in the turbine building, this portion of the ventilation system is to remain operable through a design basis tornado; and

g. Provide the sample room exhaust hood with a supply and exhaust filtration system and to provide tempered ventilation air to the sample room.

The components of the turbine generator building ventilation systems are classified Seismic Category II. The system fans are constructed and rated in accordance with the applicable AMCA standards.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-028 9.4-27 9.4.6.2 System Description

The HVAC systems of the turbine generator building are shown in Figure 9.4-6. The primary system is a "push-pull" heati ng and ventilating system consis ting of the following systems:

a. Main supply,

b. Main exhaust,

c. Auxiliary boiler room ventilation,

d. Transformer vault ventilation, and e. Sample room air conditioning.

Equipment details are given in Table 9.4-4. Equipment seismic information is given in Table 3.2-1.

9.4.6.2.1 Main Supply System

The turbine generator building supply air system consists of four supply ventilation units and associated distribution ductwork. The units are operated in pairs, with one pair discharging into a common supply duct system servicing the west side of the building and the other pair discharging into a common supp ly duct system servicing th e east side of the building.

Each of the four ventilation un its consists of a prefilter, a steam coil with face and bypass dampers, an evaporative cooli ng section, and a centrifugal fan enclosed in an insulated housing. The prefilters are of the renewable roll type, automatically progressed to maintain uniform pressure drop. The st eam coil is of the nonfreeze type with automatic face and bypass dampers used for temperature control. The evaporative cooling s ection is of the capillary air washer type with two full capacity pumps whic h have a common suction line from the washer basin and common discharge line into the washer spray healer system. Makeup water to the air washer is supplied from the plant potable water system.

The centrifugal fans of each ventilation unit have automatic inlet vanes for fan capacity control and are used to vary supply air flow to mainta in the turbine building at a set pressure with respect to outdoors. The fan inta ke ducts of each pa ir of ventilation units are joined by means of a manual damper so that a si ngle fan can draw air through bot h ventilation units in the event of failure of one fan. Automatic dampers are provided on the intake of each supply fan which can be controlled from the local panel.

9.4.6.2.2 Main Exhaust System

The main exhaust system consists of four roof-mounted centrifuga l fans which draw air from a central exhaust plenum. Three of the exhaust fans normally operate continuously with one fan in standby. During the winter, if the temper ature in the turbine ge nerator building can be maintained within the desi gn limit, two of the exhaus t fans can be operated.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 9.4-28 Almost all exhaust air is drawn from the shielded areas of the turb ine building, where the potential for airborne radioactiv e contamination is highest, thus inducing flow from the less contaminated areas through the sh ielded areas. The air in th e exhaust duct is monitored for radioactive contaminants by a reco rder in the main control room. Radioactive releases to the environment are discussed in Section 11.3.3.

In the event that supply air to the turbine generator building is reduced, as during a plant outage, only one or two exhaust fans may be ope rated. Motor-operated shutoff dampers are provided in the main branches of the exhaust duct system so th at exhaust can be stopped on an area-by-area basis. Automatic volume dampers ar e provided in the exhaus t system so that full exhaust flow can be drawn from the shielded equipment vaults on the lower level of the turbine building when the exhaust system is operating at reduced capacity. These vaults house equipment with higher contaminat ion potential such as the air ejectors and the offgas system hydrogen recombiners.

9.4.6.2.3 Auxiliary Boiler Room Ventilation System

The auxiliary boiler room, which is located in the lower level of the turbine generator building, is normally ventilated from the turbine generator building supply air system. This ventilation rate is sufficient when the boiler is not operating. However, when the boiler is operating and

drawing combustion air from the room, additional ventilation is supplied to the boiler room by a separate air handling unit. This air handling unit starts automatically, via electrical interlocks, when the boiler is started and draws 100% outdoor air through a weather louver, heats it as required, and di scharges it into the room. Part of this air is drawn by the boiler as combustion air, with the balance of the air leav ing the boiler room via relief dampers in the exterior wall of the boiler room.

9.4.6.2.4 Transformer Va ult Ventilation System

Two transformers, located in adjacent equipment vaults in the lower level of the turbine generator building, must remain operational in the event of a de sign basis tornado. The power feeds of the makeup water pumps , which are the source of ma keup water to the emergency spray ponds, are drawn from these transformers which can be fed from the emergency diesel generator buses. Two tube-axial fans are provided for the ventilation of th e vaults in the event of a tornado. The two fans exhaust air from the vaults with makeup air provided through ventilation openings in the vaults walls. Either fan has sufficient capacity to remove the heat generated by the transformers and both are powered from the emergency diesel generator buses. During normal plant operation, makeup air to the vaults is provided by the main turbine generator building supply air system wi th one vault ventilation fan operating and other in standby.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 9.4-29 9.4.6.2.5 Sample Room Air Conditioning System The turbine building sample room located on the lower level of the turbine generator building is provided with a sample hood exhaust filter system and a self-contained air conditioning system. The sample room hood is of the air curtain type with air supplied to the hood by a centrifugal fan which draws air fr om the corridor outside the sample room and from the room itself. Air is exhauste d from the hood by a filter unit.

This unit is composed of a medium efficiency prefilter, a HE PA filter, and an exhaust fan in a sheet metal housing. The filter unit discharges into the main turbine building exhaust system.

9.4.6.3 Safety Evaluation

The transformer vault ventilation system is designed to operate in the event of a design basis tornado. Two full-capacity fans powered from the emergency diesel buses are provided to ensure system operation in the event of a single active component failure. The vaults, which house the ventilation fans as well as the transformers, are designe d to withstand the effects of the design basis tornado.

The following features are incorporated in the design of the tu rbine generator building HVAC system to ensure system reliability and to control air movement from the potentially contaminated areas:

a. Standby exhaust fan capacity is provided (three of four fans operating) to ensure full system capacity in the event of a single fan failure,

b. The supply system ventila tion units are designed with cr oss ties between fans to minimize the effect of a fan failure on system capacity,

c. The building air is monitored by radiation monitoring system sampling from the exhaust ductwork, and

d. Exhaust air is drawn from within the shielded areas of the turbine generator building thus inducing air flow from "cle an" areas to areas of potential airborne contamination.

The radiological considera tions of normal system operation are evaluated in Chapters 11 and 12.

9.4.6.4 Testing and In spection Requirements

The performance of the HVAC systems serving the turbine genera tor building can be verified while the systems are operating. The operability and performance of standby equipment is

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 9.4-30 determined by rotating the duty of redundant components. Pre ssure, temperature, and flow instrumentation is prov i ded, as shown in Figu r e 9.4-6 , to monitor system performance.

All system ductwork was balanced for air flows in accordance with th e procedures of the AABC. All system components were subject to preoperational testi ng to verify that the system functions in accordance with the design requirements.

9.4.6.5 Instrumentation Requirements

Control devices for the turbine generator bu ilding HVAC system are mounted on racks in

unshielded areas of the building. The following major instrumentation devices are used for the turbine generator building HVAC system:

a. Main supply system: The inlet vortex damper on each supply fan is automatically controlled by a differential pressure controller and transmitter, which sense the pressure inside the tu rbine building relative to outside the building. The differential pressure controller can mainta in a differential pressure from approximately -0.

25-in. w.g. to +0.25-in. w.g.

b. Auxiliary boiler room ventilation system:

The unit is automatic ally started, via electrical interlocks, when the boiler is started.

c. Transformer vault ventilation system

The two transformer vault fans are controlled by locally mounted selector switches. One unit normally operates continuously with the second unit in st andby. In the event of operating fan failure, a pressure switch on the fan disc harge will annunciate an alarm and start the standby fan.

9.4.7 EMERGENCY

DIESEL GENERATOR BUILDING

9.4.7.1 Design Bases

Each of the three diesel generator rooms is serviced by a separate HVAC system. With the exception of the fuel oil day tank room and the oil pump room exhaust fa ns, the function of the three systems is to main tain suitable temperatures within the rooms for equipment operation. The exhaust fans provided in each of the three oil pump rooms and in the three day tank rooms prevent the buildup of oil fumes.

All three HVAC systems operate automatically to maintain ambient temperature below equipment operability limits during all emergency modes of operation (see Table 3.11-1) for the various locations in the dies el generator building. Electric heaters are designed to maintain diesel generator rooms above the minimum required for re liable equipment operation (down to the design low of 0°F) and may be supplemented as necessary by other means (down to the

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-020,09-028 9.4-31 extreme low of -27°F). The high-pressure core spray (HPCS) diesel generator batteries will be maintained at a te mperature of 60°F or greater by the HVAC system and supplemented as necessary by portable heaters or other means.

Since an independent and separate diesel generator HVAC system serves each diesel generator, a failure in one system will not effect the operational function of the other systems. The HVAC systems are housed in separate rooms in the Seismic Categor y I diesel generator building. The means of protecting system vent s and louvers from missi les is discussed in Section 3.5.

All system components (except the electric unit h eaters, the intake filte rs, the day tank room exhaust fans, and the fuel oil pump room exhaust fans) are Seismic Category I. The system fans are constructed and ra ted in accordance with app licable AMCA standards.

9.4.7.2 System Description

The three similar heating and ventilating systems serving the diesel generator rooms are shown in Figure 9.4-7.

Each room has a main "push-pull" ventilation sy stem and an exhaust system for each oil day tank room. Each diesel fuel pump room is served by its own independent ventilation system.

Equipment details are given in Table 9.4-5. Equipment seismic information is given in Table 3.2-1.

Each "push-pull" ventilation syst em is composed of two air ha ndling units, an exhaust fan, associated ductwork, and controls. The two air handling units share a fr esh air intake plenum and a common intake air filter bank. Each air handling unit has a water cooling coil (with a

bypass damper) and a centrifugal fan in a sheet metal housing. The exha ust fan is a direct drive, vaneaxial fan.

Normally, the smaller of the two air handling units operates continuously to maintain proper temperatures in the diesel generator room. Heating is provided by an electric blast coil heater located in the air handling unit discharge duct.

Ambient temperature c ontrol is provided by temperature regulated proporti onal dampers on the air handling unit intake which mix outside air and recirculated room air.

When a diesel generator is started, the larg er air handling unit and the main exhaust fan automatically start, and SW is supplied to the water cooling co ils in both air handling units.

The exhaust fan ductwork is arranged such that the exhaust ai r can be discharged outside (via the pipe area) or recirculated through the two air handling units in any proportion from 0% to 100% to control supply air temperature. The water cooling coils provi de additional cooling during high outdoor air temperatures.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 9.4-32 The fuel oil day tank exhaust fans and the fuel oil pump room exhaust fans operate to purge oil fumes from the rooms. An outside air intake louver is provided in each oil pump room. The oil pump exhaust fan discharges to the atmosphere. The explosion-proof electric unit heaters in the pump rooms will maintain the temperature above 50°F.

9.4.7.3 Safety Evaluation

Components of the three emergenc y diesel generators are servi ced by separate and independent HVAC systems. Each HVAC sy stem is powered from a Class 1E bus which is supplied by the diesel generator it serves. A ll components of each system ar e located within the equipment room it serves and are therefore protected from all external missiles.

All HVAC components required to ensure emergency diesel generator operation are Seismic Category I and Quality Class I.

The HVAC system is started au tomatically, via electr ic interlocks, whenever the associated diesel generator is started. Fa ilure of part of any of the three systems will only effect the diesel served and will not impair the operational function of the remaining two systems. These systems are designed to operate in the event of a LOCA coincide nt with loss of offsite power.

9.4.7.4 Testing and In spection Requirements

The diesel generators are normally on standby w ith the ventilation system equipment accessible for out-of-service inspection. A ll system ductwork was balanced for air flows in accordance with the procedures of AABC.

All system components were s ubject to preoperational testing to verify that the system functions in accordance with the design requi rements. All piping system components were subject to hydrostatic tests during erection.

The performance of the heating and ventilati on system components are verified while the system is in operation. The performance of standby components are verified during Technical Specifications required testing of the diesel generators.

Temperature and pressure instrumentation are provided as shown in Figure 9.4-7.

9.4.7.5 Instrumentation Requirements

The following is a discussion of the instrumentation provided for each of the three emergency diesel generator r oom HVAC systems.

The smaller air handling unit in each room is controlled by a local rack-mounted ON-OFF switch. In the ON position, th e fan operates and the control circuits of all associated equipment are energized causing the following:

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 9.4-33 a. Electric duct heater cont rol circuit is energized perm itting the heater to operate, to satisfy the electric heating thermostat. (Heaters can only operate when the differential pressure switch across the air handling un it fan indicates air flow);

b. Fresh air/recirculation air damper control circuit permitting the electronic temperature controller to proportion air to maintain a proper supply air temperature (100% outdoor air supplied when outdoor temperat ure is above the damper motor controller setpoint). A sensor located in the supply air duct upstream of the heating unit signals the temperature of the supply air to the damper control; and

c. A motor-operated face and bypass damper on the air handling unit water cooling coil is controlled by a temperature switch lo cated in the fresh air intake. When the fresh air temperature falls below the damper motor controller setpoint, the air is bypassed over the cooling coil to prevent the supply air temperature from dropping below the setp oint temperature.

The larger air handling unit fan and the main exhaust fan are contro lled by separate local rack-mounted ON-OFF-AUTO switches. Both unit switches are normally in the AUTO position and are started automatically, via electric interlocks, when ever the associated diesel generator is started. The control circuits for the dampers associated with the larger air handling unit are energized when the fan is started and operated in the same manner as those of the smaller air handling unit.

The oil day tank room and fuel oil pump room exhaust fans are all controlled by local rack mounted ON-OFF switches. The explosion proof electric unit heaters are controlled by local ON-OFF switches. In the ON position, the heaters are cycled on and off by electric room thermostats.

Differential pressure switches are installed across all fans servi ng the diesel generator rooms.

In the event of low di fferential pressure, an alarm is a nnunciated in the ma in control room and/or on local panels. Temperat ure sensors in the diesel gene rator rooms and in the exhaust ducts also annunciate alarms in the event of abnormally high or low temperatures (see Figure 9.4-7

).

9.4.8 DIESEL

GENERATOR AREA CABLE COOLING SYSTEM

9.4.8.1 Design Bases The critical electric cabling which runs between the emergency diesel generators and the main control room and critical switchgear room is routed in corridors adjacent to the diesel generator building and in corridor s between the reactor building and radwaste building. These corridors are normally ventilated by the turbin e building and radwaste building ventilation

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 9.4-34 systems with normal ambient temperatures below 115°F (except for corridor C-121 below 104°F); however, an emergency cab le cooling ventilation system is provided to ensure that ambient temperatures in the corridors do not exceed the ambient environmental temperatures for which the cables are rated, in the event of loss of offsite power. During an extreme winter outside temperature of -27°F, the incoming air is heated to a minimu m temperature of 35°F.

The ventilation system is comprised of two i ndependent and separate systems which cool Division 1 and Division 2 cable ar eas. A failure in one system will not effect the operational functions of the other cooling system.

The fresh air intake opening is located in the south exterior wall and is shielded by a concrete barrier which would preclude entry by a missile such as generated by a tornado. Division 1 exhaust fan discharges air into a cable chase which is not open to the atmosphere. The air is removed by the radwaste building exhaust system.

All components in the system are Seismic Category I, Quality Cl ass I. The system fans are constructed and rate d in accordance with AMCA standards.

9.4.8.2 System Description

The cable cooling system is shown in Figure 9.4-7. The system is composed of one exhaust

fan powered from the Division 1 emergency power bus and one s upply air handling unit powered from the Division 2 emergency power bus.

The exhaust fan, which is normally in standby, is started automatically when the Division 1 diesel generator is started. The operation of this standby exhaust fan opens the outdoor air bypass damper when outdoor air temperature is above 40°F and when the Division 2 supply fan is not running.

The Division 2 air handling unit is composed of a 30-kW electric blast coil heater and a

centrifugal fan in a sheet metal hous ing. It is normally in standb

y. When the Division 2 diesel generator is started, th is air handling unit is au tomatically started thro ugh interlocks. When this air handling unit is started, a damper in the supply air duct from the turbine building HVAC system (which is used for normal ventilati on of the corridors) is automatically closed. This air handling unit then supplies tempered air to the corridors in which Division 2 cable is routed.

9.4.8.3 Safety Evaluation All components of the cable cooling system are powered from their respective emergency diesel buses. All HVAC components in this system are Seismic Cate gory I, Quality Class I.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 9.4-35 The capacity of the Division 1 a nd Division 2 components of the HVAC system is sufficient to maintain safe ambient conditions in the event of failure of either power division. The components powered from the two divisions ar e independent and phys ically separated to preclude the possibility of any single failure damaging both systems.

9.4.8.4 Inspection and Testing Requirements

All components of the cable cooling system which are normally in st andby are accessible for out-of-service inspection. All system ductwork was balanced for air flows in accordance with the procedures of AABC. All system components were subject to preoperational testing to verify that the system functions in accordance with the design requirements.

The performance of the ventilati on system components are verified while the system is in operation. Temperature and pressure inst rumentation is provided, as shown in Figure 9.4-7 , to monitor system performance.

9.4.8.5 Instrumentation Requirements

The Division 1 standby exhaust fan and the Division 2 air ha ndling unit are controlled by separate, local, ON-AUTO-OFF switches (spring back from OFF to AUTO). With the control switches in the AUTO position, the Division 1 standby exhaust fan starts automatically, via electric interlocks, when the Divisi on 1 diesel genera tor is started.

The Division 2 air handling unit control switch is normally in the AUTO position. It

automatically starts and its control circuit energizes, via electric interlocks, when the Division 2 diesel generator is started. When energized, the air handling unit's control circuits perform the following function:

The inlet mixing damper and electric blast coil heaters are controlled by electric thermostats located in the corridor. When corridor temperature is above 70°F, the damper is positioned for 100% outdoor air. When the temperatur e drops below a nominal 70°F, the damper is positioned for minimum outdoor air. When th e temperature drops below a nominal 62°F and 54°F, respectively, the two stages of electric heater are energized.

Failure of air flow through the air handling unit fan after the fan motor is energized will trip a differential pressure switch that deenergizes the air handling unit control circuits and activates an alarm. Also, the filter bank in the air handling units is provided with a differential pressure switch which activates an alar m on high differential pressure.

The operation of the Division 1 standby exhaust fan opens the outdoor air bypass damper when outdoor air temperature is above 40°F and when the Division 2 supply fan is not running.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-007 9.4-36 9.4.9 REACTOR BUILDING EMERGENCY COOLING SYSTEMS

9.4.9.1 Design Bases

All equipment located within the Seismic Category I reactor building which requires a controlled environment to opera te, and which must operate in the event of a LOCA, is enclosed in individual equipment rooms. These rooms are normally heated and ventilated by the reactor building HVAC system (see Section 9.4.2); however, under emergency conditions, the rooms are automatically cooled by recirculation of room air th rough their respective reactor building emergency cooling system. Ambien t temperatures of th e following rooms are maintained below equipment operability limits during all emerge ncy modes of operation. See Table 3.11-1.

HPCS pump room

Division 1 Low-pressure core spray (LPCS) pump room Division 2 Reactor core isola tion cooling (RCIC) pump room Division 1 Residual heat removal (RHR) pump room Division 2 RHR pump rooms (2 pump rooms)

Division 1 Motor control cen ter (MCC) room (el. 522)

Division 2 MCC room (el. 522)

Division 1 dc MCC room (el. 471)

Division 1 H2 recombiner MCC room (el. 572)

Division 2 H2 recombiner MCC room (el. 572)

Fuel pool cooling (FPC) pump room

The critical MCC and FPC pump ro oms emergency cooling fans auto start and the rooms are isolated from the reactor build ing HVAC system on an F, A, or Z signal. Although the FPC pump room is isolated from the reactor building HVAC system, the room temperature can be maintained below equipment opera bility limits with the reactor building HVAC dampers open. The ECCS and RCIC pump rooms emergency cooling fans auto start when their respective pump starts. SW is provided to the emerge ncy cooling coils as described in Section 9.4.9.5.

All components of the reactor building emergency cooling system are designed Quality Class I, Seismic Category I and are powered from the same diesel generator bus as the equipment being served. Since each separate cooling system services redundant emergency equipment systems, a failure of one cooling system will not effect the operationa l function of the other cooling systems or the safe shutdown of the reactor. The means of protecting the system vents and louvers from missiles is discussed in Section 3.5.

All ductwork connected to the fan coil units in this system is designe d to Seismic Category I requirements. The system fans are constructed and rated in accordance with the applicable AMCA standards. The water cooling coils are designed and code stamped in accordance with the requirements of the ASME Code Section III, Class 3.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-007 9.4-37 9.4.9.2 System Description

The reactor building emergency cooling system is shown in Figure 9.4-2. Each of the rooms housing critical equipment is provided with an individual air ha ndling unit (two units in the FPC pump room) which is fully en closed within the room. Each air handling unit is comprised of a direct drive centrifugal or vane axial fan and a water cooling coil in a sheet metal housing.

Water is supplied to the water coils by the SW system (see Section 9.2.7). During normal operation, these air handli ng units are in standby.

All units recirculate the air with in the room they serve, removing the heat generated in the room via the water coil, to maintain temperatures below the design limits. Equipment details are on Table 9.4-2. Equipment seismic information is given in Table 3.2-1.

9.4.9.3 Safety Evaluation

Each of the emergency equipment rooms in the reactor building is provided with a separate, independent cooling system, all components of wh ich are located within the room serviced. Each cooling system is powered from a Class 1E bus of the same division as the equipment it serves and is designed to withstand the effects of an SSE.

A failure of one c ooling system will not affect the operational function of any other system or the safe shutdown of the reactor.

With the exception of the FPC pump room the air handling units serving the pump rooms are interlocked electrically with th e pumps they serve in such a manner that they start when the pump is started. The air ha ndling units serving the FPC pump room and critical MCC rooms are started by any of the F, A, or Z isolation signals.

The FPC pump room air handling units also start on loss of offsite power.

9.4.9.4 Testing and In spection Requirements

All components of the reactor building emergency cooling system are normally in standby and are accessible for out-of-service inspection.

All piping systems were subject to hydrostatic tests during manufact ure and erection.

The cooling system was subject to preoperational testing to verify that the system functions in accordance with the design require ments, and performance is verified periodically by testing during unit operation.

9.4.9.5 Instrumentation Requirements

With the exception of the FPC pump room units, the air handling units serving the pump rooms are controlled identically. Each is electrically interlocked with the pump it serves to

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-007 9.4-38 operate when the pump operates. Running lights for the fa ns are located on the MCCs. The SW system supplies the air hand ling unit water coil when the pump is started. A local manual switch is provided in each pump room for testing the air handling unit fan.

The controls for each of the air handling units serving the FPC pump room and the critical MCC rooms are identical. An ON-AUTO-OFF switch (spring back from OFF to AUTO) is provided for each fan unit in the main control ro om. Normally all switches are in the AUTO position and the fan units are in standby. In the AUTO mode, any of the three isolation signals (F, A, Z) will cause the following op erations, via electric interlocks.

a. Air handling unit fans start,

b. Solenoid valves associated with the ai r-operated dampers in the reactor building ventilation system supply air ducts to the critical MCC and FPC pump rooms are deenergized, thus isolating these rooms from the balance of the reactor building.

For the F isolation signal, the Division 1 and 2 SW systems are auto started as described in Section 7.3.1.1.6 and provide cooling water to these unit cooling coils. For the A isolation signal, the Division 2 SW system is auto started via RCIC as described in Section 7.3.1.1.6 and the Division 1 SW system ca n be manually started as needed from the control room. For the Z isolation signal, the Division 1 and 2 SW systems can be manually started as needed from the control room.

In addition, the FPC pump room fan units will start on loss of offsite power when the control switch is in the AUTO position.

Room temperature indicators, al arms and/or HVAC related alarms are provided in the main control room for each of the equipment rooms as shown on Figure 9.4-2 or 9.2-12 , as applicable. The alarms will annunciate in the event that temperatures exceed the design limit.

In addition, the leak detection system provides temperature indi cation and alarms in the main control room for Division 1 RHR A pump room, Division 2 RHR B pump room and the RCIC pump room as shown on Figure 7.6-1. 9.4.10 STANDBY SERVICE WATER PUMP HOUSE

9.4.10.1 Design Bases

The SW pump house HVAC systems are designed to remove the heat generated by operation of the SW pumps and the HPCS service wate r pump and to limit the temperature in the two pump houses as specified in Table 3.11-1. The ventilation systems are designed as ESF systems and are powered from a Class 1E bus of the same division as th e pumps being served.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 9.4-39 Since each HVAC system as sociated with the redundant SW pumps is separate and independent, a failure in one syst em will not affect the operational function of the other system.

The means of protecting the system vents and louvers from missi les is discussed in Section

3.5. Heating

of the pump houses is provided by electric unit heaters (four in each building) sized to heat the building spaces to maintain temperatures as specified in Table 3.11-1. The unit heaters are powered from the emergency diesel-generator buses, but are classified Seismic Category II. If the pumps are required to operate following a seismic event, the heat generated from the pump motor is sufficient to maintain the buildi ng above freezing. The SW pumphouse outside air suppl y fans and motors are classified Seismic Category IM. These fans and motors are not required to ope rate to cool the safety relate d equipment in the pump houses under accident conditions. The remaining components, ductwork, and piping in the SW pump house ventilation systems are cl assified Seis mic Category I.

The fan coil units and the supply fans are c onstructed and rated in accordance with the applicable AMCA standards. The water coo ling coils are designed and code stamped in accordance with the requirements of ASME Code Section III, Class 3.

9.4.10.2 System Description

The SW pumps are located in pump houses adjacent to the emergency spray ponds. The loop A SW pump and the HPCS service water pump share one pump house and the loop B SW pump is in a second pump hous

e. The HVAC systems serv ing the two pump houses are depicted in Figure 9.4-7. Each system consists of a fan coil unit composed of a sheet metal cabinet containing a direct-drive centrifugal fan and a water cooling coil, and a separate centrifugal supply fan with inlet mixing dampers.

Each fan coil unit is interlocked electrically with the associated SW pump it serves in such a manner that the unit fan starts and recirculates room air over the water cooling coil when the pump starts. Water is supplied to the unit coil from the main supply header of the SW pump. The fan coil units are normally in standby and operate only when the pump is started.

The supply ventilation fans in bot h pump rooms operate automatica lly when required. A room thermostat in the pump house starts the fan on a temperature rise to 80°F or above and stops the fan on a temperature drop to 60°F or below. When the outdoor temperature is above 40°F, 100% outdoor air is supp lied to the room. When the outdoor temperature falls below 40°F, a temperature indicating switch, with its se nsor in the outdoor ai r intake duct, positions the motor-operated intake mi xing damper to a position for 440 0 cfm recirculation air and 600 cfm outdoor air. Air is exhausted from the pump houses through relie f dampers at a rate between 600 cfm and 5000 cfm depending on the intake of outside air.

There are four 10-kW electric unit heaters in each pump house. Each heater is controlled by a separate wall-mounted thermostat which starts the heater on a drop in room temperature to

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 9.4-40 below 40°F. Equipment details are given in Table 9.4-6. Equipment seismic information is given in Table 3.2-1. 9.4.10.3 Safety Evaluation

The following safety features are incorporated in the design of the heating and ventilating systems serving the SW pump houses:

a. Each of SW pump rooms 1A and 1B is serviced by a se parate 100%-capacity fan coil unit powered from the same Class 1E bus as the pump room being serviced,

b. The HPCS service water pump is located in SW pump house 1A,

c. Both fan coil units and the HPCS supply fan start when the pump serviced starts,

d. In the event of loss of offsite power, the ventilation equipment is powered from the emergency diesel generator buses, and

e. All ESF components of th e system are designed to ope rate in the event of an SSE.

Since each HVAC system as sociated with the redundant SW pumps is separate and independent, a failure in one system will not effect the ope rational function of the other system.

9.4.10.4 Testing and In spection Requirements

The performance of normally operating componen ts of the SW pump house ventilating systems can be verified while operating. All sta ndby equipment is accessible for out-of-service inspection. All system ductwork was balanced for air flows in accordance with the procedures of AABC. All system co mponents were subject to preoperational testing to verify that the system functions in accordance with the design requirements. All piping systems and components were subject to hydro static tests during erection.

9.4.10.5 Instrumentation Requirements

In addition to the instrument ation describe d in Section 9.4.10.3 the following devices are provided:

a. Each fan coil unit fan is provided with a local rack-mounted ON-AUTO-OFF switch (spring back to AUTO) which is used to test fan operation, C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 9.4-41 b. Each supply air fan is provided with a local rack-mounted ON-AUTO-OFF switch (spring back from OFF to AUTO) which is used to test fan operation,

c. Differential pressure switches are provided across each fan coil unit fan and each supply fan which annunciate an alarm in the main control room in the event of low differential pressure when the fan is given a start signal, and

d. Temperature switches in each pump room annunciate alarms in the main control room in the event of high or low temperatures.

9.4.11 PRIMARY CONTAINMENT

9.4.11.1 Design Bases

The primary containment is divided into two se parate volumes: the dr ywell which houses the reactor vessel and auxiliaries, and the wetwell which housed th e suppression pool. There is no heat producing equipment in the wetwell; ther efore, a wetwell air cooling system is not required. The drywell, which has a high internal heat load, is provided with recirculation cooling system designed to m eet the following requirements:

a. To limit during normal plant operation, the average air temperature within the primary containment drywell to Table 3.11-1 limits,

b. To provide suitable working temperatur es inside the primary containment for personnel during shutdown a nd refueling operations, and

c. To remove the additional heat released in the event of a reactor scram, and to limit temperatures in the neutron monito ring cable area beneath the reactor to Table 3.11-1 limits.

The primary containment cooling system is not required for safe shutdown of the reactor. Essential equipment located in primary containm ent that is required for safe shutdown of the reactor is designed to function without the containment cooling system in operation.

All components of the primary containment cooling system are classified Seismic Category I. System fans are constructed and rated in accordance with applicable AMCA codes. The water cooling coils are designed and c ode stamped in accordance with the requirements of the ASME Code Section III, Class 3.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 9.4-42 9.4.11.2 System Description Equipment seismic information is given in Table 3.2-1. The equipment details are described in Table 9.4-2. The primary containment cooling system is shown in Figure 9.4-8. Cooling of the drywell is provided by five fan coil units which recirculate containment air through water cooling coils for heat removal. Seven recirculation fans and two head area return fans are also provided at various locations in the drywell to provide additional air turbulence to prevent pockets of hot air from developing. Each of the five fan coil units consists of two vaneaxial fans, both of which operate at th e same time, and a water cooling coil in a sheet metal housing.

Provisions have been made to in stall filters in the un its, for coil protection, while the units are operated during plant constructi on; however, no filters are required during normal plant operation. Water is supplied to the unit cooling coils from the reactor building closed cooling water system as de scribed in Section

9.2.2. Three

of the fan coil units are located low in the drywell and two of the units are located at a higher level.

Each of the three lower level fan coil units are provided with tw o vaneaxial fans one of which discharges directly into the general drywell volume, and the other di scharges, via ductwork, into the neutron monitoring cable area beneath the reactor and general drywell volume.

The two upper level fan coil units are each pr ovided with one 30,000-cfm capacity fan and one 10,000-cfm capacity fa

n. The larger fans di scharge directly into the upper volume area.

The smaller fans discharge into the containm ent head area above th e refueling bellows. Two vaneaxial head area return fans draw air from the containment head area and discharge it

below the refueling bellows. Return air to the upper fan coil units is drawn from immediately below the refueling bellows. During normal operation, one or bot h of the head area return fans will be running.

Three recirculating fans are located at lower level a nd four fans are located at upper level in the drywell to provide air circulation. During normal operation, up to three lower level fans

and up to four upper level fans are operating.

In the event of a reactor scram, both of the h ead area return fans are started (unless already operating).

9.4.11.3 Safety Evaluation

The primary containment cooling system does not have to operate, in the event of LOCA, to ensure the safe shutdown of the reactor. Head area re turn fans (CRA-FN-4A and CRA-FN-4B) will operate post-LOCA to help en sure hydrogen mixing.

Design of the system incorporates the following featur es to ensure safe operation.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 9.4-43 a. All fan coil units, recirculation fans, and head area return fans are designed to withstand the effects of an SSE and ar e powered from the emergency diesel buses; b. All containment fans have been evaluated for the normal environmental conditions in containment to ensure long term reliable operation; and

c. Those fans used for hydrogen mixing following a LOCA are identified and qualified to Class 1E require ments. Additionally, these fans are powered from Class 1E sources.

9.4.11.4 Testing and In spection Requirements

All components of the primary containment cooli ng system were subject to shop and field tests prior to plant operation. Syst em performance is verified dur ing reactor operation when the equipment is operating at design conditions.

All system ductwork was balanced for air flows in accordance with th e procedures of the AABC. All system components were subject to preoperational testi ng to verify that the system functions in accordance with the design requirements.

9.4.11.5 Instrumentation Requirements

The primary containment fan coil units, recirculation fans, and c ooling water isolation valves are each controlled by individual selector switche s with indication lights, mounted in the main control room.

Temperature sensors ar e located throughout the primary containment as indicated in Figure 9.4-8. These sensors continuously monitor ambient conditions inside the containment and the performance of the cooling system.

9.4.12 MAKEUP WATER PUMP HOUSE

9.4.12.1 Design Bases The heating and ventilating system provided in the makeup water pump house is designed to maintain temperatures within the structure between 50°F and 104°F to ensure suitable conditions for equipment operation. In the ev ent of the hypothesized dewatering of the SW spray ponds due to a tornado, the makeup water pumps may be operated to refill the spray ponds (see Sections 9.2.7 and 10.4.5). Since the ventilation system must be operated to ensure an acceptable environment for the makeup pump motors, the system is designed with redundant equipm ent to ensure that a single

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 9.4-44 component failure will not interf ere with the operation function of the system.

In the event a tornado causes the loss of offsite power, the system is powered from the emergency diesel generator buses. The fresh air intake and exhaust air openings are located in the east exterior wall and are shielded by a concre te barrier wall which precludes entry of a missile such as a tornado-generated missile.

The makeup water pumps and auxili aries are not required to opera te in the event of an SSE; therefore, all components of the heating and ventilating system serving the pump house are designed to Seismic Category II requi rements as defined in Section 3.2. The system fans are constructed and rated in accordance with applicable AMCA standards.

9.4.12.2 System Description

The makeup water pump house heating a nd ventilating system is shown in Figure 9.4-7. It consists of two full-capacity ai r handling units and two battery hood exhaust fans which service the electric equipment area, and two full-capacity fan coil units and two electric space heaters which service the pump area. E quipment details are given in Table 9.4-7. Equipment seismic information is given in Table 3.2-1.

The two air handling units serving the electric equipment area consist of an insulated sheet metal cabinet housing a replaceable roughing filter, a two-stage electric blast coil heater, a water cooling coil, and a centrif ugal fan. One of the two air handling units operates at all times to maintain design temperat ures in the electric equipmen t area. The second unit is in standby and starts in the event that the operating unit fails.

The air handling units draw air from the outside atmosphere through intake louvers. The air is discharged, via ductwork, into the electric equipment area from which it flows into the pump room. It is then released eith er to the outside atmosphere, via relief dampers, or is partially recirculated back through the unit. Motor-operated dampers on the unit intake ducts are so arranged that the unit can draw 100% outdoor air or recirculate air drawn from the pump area back through the unit. The da mper motor is controlled by a te mperature switch which senses outdoor temperature. The damper will be pos itioned for 100% outdoor air when the outside temperature is between 50°F and 70°F.

The fan coil units servicing the pump area consist of a centrifugal fan, a water cooling coil, and a roughing filter in a sheet metal housing. The units recirc ulate air in the pump room only and are interlocked electrically with the makeup water pumps to start when the pumps start.

Each unit has sufficient capacity to maintain design conditio ns with two of the makeup water pumps operating. In the event all three pumps are operating, th e second fan coil is automatically started.

Water is supplied to the electr ical area air handling units and the pump area fan coil units from the discharge header of the makeup water pumps. A fail-clos ed, motor-operated valve is

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 9.4-45 provided in the water supply line to each unit. The valve control circuit is energized when its respective fan is energized. With the control circuit energized, the valve is controlled by a thermostat, located in the area serviced, wh ich opens the valve on a temperature rise to 90°F or above.

The electric blast coil heaters in the electrical equipment area air hand ling units are controlled by separate two-stage room th ermostats. Two electric unit heaters provide supplementary heating to the pump area. A temperature switch starts the heaters at the desired temperature setpoint.

Two spark resistant, battery hood exhaust fans both operate continuously to exhaust any combustible gases generated from the batteries to the atmosphere.

9.4.12.3 Safety Evaluation

The makeup water pumps are requi red to supply water to the SW spray ponds in the event a design basis tornado empties the ponds of their coolant (see Section 9.2.7). The pump house HVAC system is required to ensure that the operation of the makeup pu mps is not diminished by extremes in temperature.

The HVAC system pr ovided in the makeup water pump house incorporates the following safety features to ensure that a single component failure will not prevent the system from perfor ming its operational function:

a. Two full-capacity air handling units are provided for the electrical equipment area and two full-capacity fan coil units are provided for the makeup pump area; therefore, failure of any one unit will not effect system operation,

b. The redundant HVAC equipment is power ed from different divisions of the emergency diesel generator buses; therefore, failure of any one bus will affect only one train of ven tilating equipment, and

c. All heating and ventilating equipment is located within th e pump house where it is protected from tornado missiles.

9.4.12.4 Testing and In spection Requirements The performance of the heating and ventilating system servicing the makeup water pump house can be verified while the systems are operating. The operability of standby equipment is determined by rotating the duty of redundant systems.

All systems ductwork was balanced for air flows in accordance with the procedures of AABC and all piping systems and components were subject to hydros tatic tests during erection.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 9.4-46 9.4.12.5 Instrumentation Requirements The following major instrumentation devices are used in the control and monitoring of the makeup water pump house heating and ventilating system in addition to those described in Section 9.4.12.2: a. Three position selector switches (ON-OFF-AUTO) are provided locally for each of the two pump area fan coil units and ar e normally in the AUTO position. In this position, the starting of the makeup pump, powered from the same bus as the fan coil unit, will start the fan coil unit and energize its control circuit via an electric interlock. Th e second makeup pump start will not start the second fan coil unit. Both fan coil units operate onl y when all three pumps are operating.

Differential pressure switch es across the fan of each unit will annunciate an alarm and start the standby unit in the event of a low diffe rential pressure;

b. Three position selector switches (ON-OFF-STANDBY) are provided locally for each of the two electrical equipment area air handling units. One unit is normally operating with the second unit in standby. Differential pressure switches across each fan, annunciate an alarm and start the standby unit in the event of low differential pre ssure across the operating fan;

c. Each of the two battery hood exhaust fans are provided with local selector switches and a differential pressure switch across th e fan which annunciates a local alarm in the event of fan failure; and

d. Temperature switches located in the electrical equipment and pump areas annunciate alarms in the main control room on high or low ambient temperature conditions.

9.4.13 SERVICE BUILDING

9.4.13.1 Design Bases

The service building HVAC syst em is designed to provide a controlled environment for personnel comfort within the building. The h eating and air conditioni ng system is of a conventional design.

9.4.13.2 System Description

The service building HVAC system is a "push-pull," multizone system using chilled water for cooling and hot wa ter for heating.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 9.4-47 9.4.13.3 Safety Evaluation The service building HVAC syst em has no safety function.

Malfunction or failure of the HVAC system will not impair norma l or emergency plant operations.

There are no potential sources of radioactive contaminants, with the exception of tritium (see Section 9.4.16.3), nor safety related equipmen t in the service building.

9.4.14 WATER TREATMENT AREA AND MACHINE SHOP 9.4.14.1 Design Bases

The HVAC system serving the wa ter treatment area and machine shop, both of which are in the service building, is designed to remove noxious fumes genera ted in the area served and to provide tempered air for personne l comfort. The HVAC system is of a conven tional design. The system is sized to provide three air changes per hour in the areas served.

9.4.14.2 System Description

The system is basically a once-through system with the facility for partial recirculation of air

from the machine shop to reduce heating requirements during the winter months.

Air is supplied to the machine shop and water treatment area by a centr al supply air system composed of a ventilation unit and distribution ductwork.

Exhaust fans are provided to exhaust air from the machine shop and water treatment area.

9.4.14.3 Safety Evaluation

The heating and ventilating system serving th e water treatment area and machine shop has no safety function. Malf unction or failure of the system will not effect reactor operation nor cause the release of radioactive materials.

There are no potential sources of radioactive contaminants, with the exception of tritium (see Section 9.4.16.3) nor safety-related equipment in the areas serviced.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 9.4-48 9.4.15 CIRCULATING WATER PUMP HOUSE

9.4.15.1 Design Bases

The HVAC system serving the circulating pum p house is designed to satisfy the following criteria:

a. Limit the maximum temperature in the circulating pump area to within 10°F of the ambient temperature when the pumps are operating,

b. Maintain the pump area at a minimum temperature of 50°F when the pumps are not operating,

c. Maintain the electrical equipment r oom and control room pressurized, with respect to the adjacent rooms, to prevent the ingre ss of noxious fumes, and maintain the room temperature between 65°F and 85°F,

d. Exhaust air from the halogen storage tank room, halogen injection pump room, and acid pump room to prevent the potential buildup of noxious fumes, and heat the halogen injection pump room to 65°F for personnel comfort during room occupancy, and

e. Continuously exhaust air to atmosphere from the diesel oil storage room to prevent the potential buildup of combustible fumes.

9.4.15.2 System Description

The HVAC systems serving the circulating water pump house are shown in Figure 9.4-7.

The pump area is ventilated by six 33,000-cfm cap acity roof exhaust fans, each of which is controlled by a manual switch.

When the exhaust fans are on, air is induced into the pump area through weather louvers in the exterior walls of the structure. Heat is provided in the pump area by twelve electric unit heaters, each controlled by separate room thermostats.

The electrical equipment room a nd control room are serviced by an air-cooled, roof-mounted air conditioning unit. The unit has a built-in, el ectric blast coil heater, a roughing filter bank, an integral air-cooled condenser , and a centrifugal fan. The un it supplies air to the two rooms, drawing outdoor air to pressurize the rooms and recirculating ai r from the electric equipment room. The pressurizing outside air supplied to the rooms is relieved to the pump area via relief dampers in the partition wall be tween the rooms and the pump area.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 9.4-49 A fiberglass roof ventilator con tinuously exhausts air from th e halogen injection pump room, and the acid pump room. The halogen storage tank room is served by a heat pump unit, which recirculates air from this area.

If needed, the fiberglass roof ve ntilator can be used to exhaust air from the halogen storage tank room by opening a volume damper in the exhaust duct.

Makeup air is induced into the halogen injection pump room and the acid pump room through relief dampers from the pump area. An electric unit heater is provided in the halogen injection pump room for personnel comfort when the area is occupied. A motor-operated damper in the branch exhaust duct from the halogen inject ion pump room is controlled by a thermostat located in the halogen storage tank room in such a manner that air exhaust from the halogen injection pump room can be reduced when the outdoor temperature (as sensed in the halogen storage tank room) falls below a predetermined setpoint to minimize the halogen injection pump room heati ng requirements.

An explosion proof wall fan conti nuously exhausts air from the di esel oil tank storage room to the atmosphere to prevent the buildup of combustib le gases. Air is in troduced into the room from the pump area through a vent opening pr otected by a 3-hr rated fire damper. Air exhausted from the room is also vented through a 3-hr rated fi re damper. Equipment details are given in Table 9.4-8. Equipment seismic in formation is given in Table 3.2-1.

9.4.15.3 Safety Evaluation

The HVAC system serving the circulating water pump house has no safety function. Malfunction or failure of the system will not impair normal or emergency operation.

9.4.15.4 Testing and In spection Requirements

All system ductwork was balanced for air flows in accordance with the pr ocedures of AABC.

All system components were subjected to preoperational testing to verify that the system functions in accordance with the design requirements.

9.4.15.5 Instrumentation Requirements

Each electric unit heater is put in service by a local hand switch and is controlled by its own thermostat.

Each fan is put in service by an OFF-ON switch.

The electrical equipment room air conditioning unit is put in service by a local ON-OFF switch. When ON, the fan runs continuously and the unit control circuit is energized permitting a one-stage cooling, two-stage heating thermostat to cycle the unit compressor or the two-stage electric blast coil heater, as required.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 9.4-50 Pressure switches with sensors in the exhaust duct of the chlo rine exhaust fan and the supply duct of the air conditioning unit fan will annunciate alarms in the even t of a fan failure.

9.4.16 PLANT HEATING STEAM SYSTEM

9.4.16.1 Design Basis

The plant HS system, in conjunction with heating coils, humidifiers, and steam unit heaters in the plant HVAC systems, discussed previously in this section, is designed to:

a. Maintain a minimum temperature of 50°F in the reactor, radwaste, and turbine generator buildings during cold weather conditions,

b. Maintain a minimum temperature of 65°F on the refueling floor of the reactor building and in the service building m achine shop and water treatment areas,

c. Provide humidified air to the radwaste control room, radiochemistry laboratory, counting room, and hot instrument shop, and

d. Generate, through a hot water heat exchanger, 180°F water for perimeter heating of service building offices and laboratories.

The plant HS system is designed for 50 psig service. The source of the HS is the gland steam evaporator during turbine generator operation and the auxiliary boiler during plant shutdown.

The heating condensate (HCO) system is designed as a gravity return to the auxiliary boiler condensate return tank (CO-TK-1) in the turbine generator building.

9.4.16.2 System Description

The plant HS system is depicted in Figure 9.4-9. The system originates from four pressure reducing stations (two in the turbine generator building and one each in the reactor and radwaste buildings). Steam at 200 psig pressure is supplied to th ese pressure reducing stations from either the auxiliary boiler or the gland steam evaporator. At the pressure reducing stations, the steam pressure is reduced to 50 psig and this steam is fed to the heating coils, humidifiers, steam unit heaters, and hot water heat exchanger.

The condensate from the HS syst em is returned to the auxili ary boiler condensate return tank located in the auxiliary boiler room of the turbine generator building. Condensate from the reactor building, turbine generato r building, and upper level of the service building is returned to the auxiliary condensate return tank by gravity.

In the radwaste building and lower level of the service building, the condensate returns are below the level of the auxiliary condensate return tank. A condensate pump-set is, therefore, C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 9.4-51 provided in each of these areas to pump the condensate to the return tank. Equipment details are given in Table 9.4-9. Equipment seismic in formation is given in Table 3.2-1. 9.4.16.3 Safety Evaluation

The auxiliary boiler and associated steam systems were origina lly intended to be free from radiological contaminati on. During operation, however, the system has become contaminated with tritium. Possible sources of the tritium activity are tube le aks in the feedwa ter heaters or the steam evaporator. As is discussed in Section 11.1.3 , all tritium produced in the reactor is eventually released to the environs. Tritium may be released from the heating system in vapor and gaseous form or in auxilia ry boiler water blowdown. The boiler blowdown tank drains to a turbine building sump (see Section 9.3.3.2.3.1) which is directed to radwaste processing.

The tritium contamination in the auxiliary boiler and associated steam systems is monitored and efforts are made to minimize the levels of activity. The tritium does not necessitate changes in system design or operation and will not cause significant radiological impacts.

The plant HS system has no safety function. Any rupture in HS or HCO piping does not impair safe reactor shutdown as discussed in Sections 3.6.1.15.3 and 3.6.1.18.3.6. All system piping except that in the reactor building is Seismic Category II. All system piping in the reactor building is Seismic Category I.

9.4.16.4 Testing and In spection Requirements

The performance of plant HS system was verifi ed by tests prior to startup and is verified during system operation. All components in th e system were tested and inspected at the manufacturers plant for conformance with specifications. After installation, the major components were checked and the system hydrostatically tested to ensure leaktightness.

9.4.16.5 Instrumentation Requirements

Adequate instrumentation is provided to monitor and control operation of the system.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-028 9.4-53 Table 9.4-1

Main Control Room/Cable Spreadi ng Room/Critical Switchgear Area Major Components of HVAC Systems

a. Exhaust fans 1. Tag number WEA-FN-51 WEA-FN-52 WEA-FN-53A WEA-FN-53B 2. Number of fans 1 1 1 1 3. Area served Main control room Cable spreading room Battery room 1 Battery room 2 4. Fan type and style Centrifugal SWSI Centrifugal SWSI Centrifugal SWSI Centrifugal SWSI 5. Drive Direct Direct Direct Direct 6. Capacity 750 acfm 1000 scfm 2400 scfm 900 scfm

7. Total static pressure (in., w.g.) 1.4 1.75 1.64 1.5 8. Motor type Open drip-proof self-ventilated with Class B insulation with a maximum temperature rise of 50° C above a 50°C ambient 9. Motor (hp) 1 1 2 1

b. Air handling units 1. Tag number WMA-AH-51A WMA-AH-51B WMA-AH-52A WMA-AH-52B WMA-AH-53A

WMA-AH-53B 2. Number of units 2 (1 standby) 2 (1 standby) 2 3. Area served Main control room Cable spreading room Critical switchgear 4. Air flow per unit (acfm) 21,000 9500 28,100 (22,800 for WMA-AH-53B)

5. Total static pressure (in., w.g.) 5.50 4.27 6.13

6. Sensible cooling capacity (Btu/hr) per

unit a) Normal operation 600,000 235,000 370,000 (320,000 for WMA-AH-53B) b) Emergency operation 285,000 160,000 370,000 (320,000 for WMA-AH-53B) 7. Water supply source a) Normal operation Chilled water Chilled water Chilled water or Plant service water b) Emergency operation CCH or standby service water Standby service water Standby service water

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-028 9.4-54 Table 9.4-1

Main Control Room/Cable Spreadi ng Room/Critical Switchgear Area Major Components of HVAC Systems (Continued)

c. Main control room emergency filter units

1. Tag number WMA-FU-54A, WMA-FU-54B

2. Number of units 2 3. Capacity (scfm) 1000 4. Fan Centrifugal direct drive

5. Total static press (in., w.g.)

9.6 6. Prefilters a) Type Replaceable media b) Efficiency Min 80-85% (atmospheric dust spot ASHRAE Standard 52-68) c) Initial pressure drop 0.5 in. maximum 7. HEPA filters a) Type High efficiency b) Media Glass media, water repellent and fire resistant c) Efficiency (DOP 0.3

µ particle size) 99.97% d) Initial pressure drop (in., w.g.)

1.0 maximum

8. Charcoal adsorber a) Media Activated carbon b) Test efficiency @ 70% RH 97.5% (methyl iodide)

d. Emergency chiller units

1. Tag number CCH-CR-1A, CCH-CR-1B 2. Number of units 2 (1 each division) 3. Peak operating load 50 tons

4. Motor (kW) 126 5. Chilled water (gpm) 145 6. Chilled water supply temperature 44°F

7. Condenser cooling source Standby service water

e. Emergency chiller pumps

1. Tag number CCH-P-1A, CCH-P-1B 2. Number of units 2 (1 each division) 3. Capacity 145 gpm at 56 ft total head 4. Motor (hp) 7.5

5. Units served CCH-CR-1A, CCH-CR-1B

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-028 9.4-55 Table 9.4-2

Reactor Building and Primary Containment Areas Major Components of HVAC Systems

a. Heating and ventilating unit (evaporative air washer)

1. Tag number ROA-HV-1 2. Number of units 1 3. Air flow (scfm) 80,000 4. Steam flow (lb/hr) 6800 5. Supply air temperature a) Summer 72°F DB (evaporative cooling) b) Winter 50°F 6. Fan type Vaneaxial V-belt drive (200 hp motor)

7. Number of fans 2 (1 standby) 8. Fan total pressure (in., w.g.)

9.3 b. Reactor building exhaust fans

1. Tag number REA-FN-1A REA-FN-1B 2. Number of fans 2 (1 standby)

3. Fan type Vaneaxial

4. Drive Direct (200 hp motor)

5. Capacity 80,000 - 105,000

6. Total (scfm) pressure (in., w.g.) 9.43 @ 105,000 scfm 7. Motor type TEAO with Class "RH" insulation

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-007 9.4-56 Table 9.4-2

Reactor Building and Primary Containment Areas Major Components of HVAC Systems (Continued)

c. Reactor building emergency fan coil units (part 1 of 2) 1. Tag number RRA-FC-1 RRA-FC-2 RRA-FC-3 RRA-FC-19 RRA-FC-20

RRA-FC-4

RRA-FC-5

RRA-FC-6 2. Number of units 3 2 1 1 1

3. Air flow per unit (acfm) 5208 10,000 15,625 9375 3125

4. Sensible cooling capacity (Btu/hr) per unit 165,000 134,000 500,000 280,000 60,000

5. Total static pressure (in.,w.g.) 1.34 0.5 1.64 1.46 1.53 6. Area served RHR pump rooms FPC pump room HPCS pump

room LPCS pump

room RCIC pump

room 7. Water supply service Standby service water Standby service water Standby service water Standby service water Standby service water c. Reactor building emergency fan coil units (part 2 of 2) 1. Tag number RRA-FC-10 RRA-FC-11 RRA-FC-12 RRA-FC-13 RRA-FC-14

2. Number of units 2 1 2 3. Air flow per unit (acfm) 5730 6500 4100

4. Sensible cooling capacity (Btu/hr) per unit 71,280 85,000 53,900

5. Total static pressure (in.,w.g.) 0.5 1.41 1.39

6. Area served MCC rooms Division 1, dc-MCC-room H 2 recombiner rooms 7. Water supply service Standby service water Standby service water Standby service water

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-028 9.4-57 Table 9.4-2

Reactor Building and Primary Containment Areas Major Components of HVAC Systems (Continued)

d. Reactor building steam tunnel fan coil units

1. Tag number RRA-FC-8 RRA-FC-9 RRA-FC-21

2. Number of units 3 3. Air flow per unit (scfm) 4250 (RRA-FC-8) 4000 (RRA-FC-9)

3000 (RRA-FC-21) 4. Sensible cooling capacity (Btuh) per unit 97,200

5. Total static pressure (in., w.g.)

1.53 6. Area served Steam tunnel 7. Water supply source Plant service water

e. Vehicle air lock (railroad bay) HVAC

1. Tag number RRA-AH-7 RRA-AH-8 RRA-EUH-1 RRA-EUH-2 RRA-EUH-3 RRA-EUH-4 RRA-EUH-5 RRA-EUH-6 RRA-EUH-7 RRA-EUH-8 2. Number of units 1 1 8 3. Fan type Centrifugal DWDI Centrifugal Axial

4. Drive (per unit) V-belt (3 hp motor)

Belts (20 hp motor) Direct drive (1/3 hp motor) 5. Nominal air flow 5000 scfm 32,500 cfm 1635 cfm 6. Sensible cooling capacity at nominal air flow (MBtu/hr) 42.5 None None 7. Electric heating capacity per unit (kw) 15 None 38.4

8. Total static pressure (in., w.g.) 1.68 N/A N/A

9. Area served Vehicle air lock (railroad bay) Vehicle air lock (railroad bay)

R106 Door Vehicle air lock (railroad bay )

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-028 9.4-58 Table 9.4-2

Reactor Building and Primary Containment Areas Major Components of HVAC Systems (Continued)

f. Sump vent filter units

1. Tag number REA-FU-2A REA-FU-2B

2. Number of units 2 (1 standby)

3. Capacity (scfm) 1000 4. Fan Centrifugal direct drive

5. Total static pressure (in., w.g.)

9.6 6. Prefilters a) Type Replaceable media b) Efficiency 80-85% minimum (atmosphere dust spot ASHRAE Standard 52-68) c) Initial pressure drop (in., w.g.) 0.5

7. Heater a) Type Electric resistance (3 stage) b) Quantity 1 c) Capacity 21 kW total 8. HEPA filters a) Type High efficiency b) Media Glass media, water repellent and fire resistant c) Efficiency (DOP 0.3

µ particle size) 99.97% d) Initial pressure drop (in., w.g.) Maximum 1.0 9. Charcoal adsorber a) Media Activated carbon b) Test efficiency @ 70% RH 99.9% elemental iodine 99% methyl iodide 10. Moisture separator a) Type Impingement b) Initial pressure drop (in., w.g.) Maximum 1.0

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-028 9.4-59 Table 9.4-2

Reactor Building and Primary Containment Areas Major Components of HVAC Systems (Continued)

g. Sample room fume hood exhaust filter unit

1. Tag number REA-FU-15

2. Number of units 1 3. Capacity (scfm) 850 4. Type of fan Centrifugal direct drive (3 hp motor)

5. Prefilter Replaceable viscous impingement type 6. HEPA filter Glass media, water repellent and fire resistant 7. HEPA filter efficiency (DOP 0.3

µ particle size) 99.97% 8. Total static pressure (in., w.g.) 4.5

h. Primary containment fan coil units

1. Tag number CRA-FC-1A CRA-FC-1B CRA-FC-1C

CRA-FC-2A CRA-FC-2B 2. Number of units 3 (1 standby) 2 (1 standby) 3. Fan type Vaneaxial (2 per unit) Vaneaxial (2 per unit)

4. Fan motor type TEAO with Class RN insulation TEAO with Class RN insulation 5. Nominal air flow (scfm) 40,000 40,000 6. Cooling capacity (sensible) at nominal air flow (Btu/hr) 1,140,000 1,140,000 7. Water supply source Reactor building closed cooling system Reactor building closed cooling system

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-036,09-028 9.4-60 Table 9.4-2

Reactor Building and Primary Containment Areas Major Components of HVAC Systems (Continued)

i. Primary containment head area return fans

1. Tag number CRA-FN-4A CRA-FN-4B

2. Number of fans 2 (1 standby)

3. Fan type Vaneaxial

4. Drive Direct 5. Capacity (acfm) 5000 6. Total pressure (in., w.g.)

1.68 (normal condition) 5.23 (accident condition)

7. Motor type TEAO with Class RN insulation

j. Primary containment recirculating fans

1. Tag number CRA-FN-3A, CRA-FN-3B, CRA-FN-3C, CRA-FN-5A, CRA-FN-5B, CRA-FN-5C, CRA-FN-5D

2. Number of fans 7 (2 standby)

3. Fan type Vaneaxial

4. Drive Direct 5. Capacity (acfm) 20,000 6. Total pressure (in., w.g.)

0.83 7. Motor type TEAO with Class RN insulation

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-028 9.4-61 Table 9.4-3 Radwaste Building Area Major Components of HVAC Systems

a. Heating and ventilating unit (evaporative air washer)

1. Tag number WOA-HV-1 2. Number of units 1 3. Maximum air flow (acfm) 91,304 (82,430 cfm normal)

4. Steam flow (lb/hr) 6980 5. Supply air temperature a) Summer 72°F DB (evaporative cooling) b) Winter 50°F (minimum)

6. Fan type Centrifugal SWSI V-belt drive (150 hp motor) 7. Number of fans 2 (1 standby) 8. Total static pressure (in., w.g.)

5.43 b. Air handling units 1. Tag number WOA-AH-3 WOA-AH-4 WOA-AH-5 WMA-AH-6 WOA-AH-9 2. Number of units 1 1 1 1 1 3. Drive V-belt (1 hp motor)

V-belt (5 hp motor)

V- belt (2 hp motor)

V-belt (5 hp motor)

V-belt (1.5 hp motor) 4. Nominal air flow (scfm) 1475 6900 3000 7000 2000

5. Cooling capacity (sensible and

latent) at

nominal air

flow (MBtu/hr) 57 255 155 204 76 6. Total static pressure (in., w.g.) 1.9 1.9 1.77 1.9 1.7 7. Area served Health physics area Radio chem

lab Hot instrument

shop Radwaste control room Counting room 8. Cooling water supply source Chilled water Chilled water Chilled water Chilled water Chilled water

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-028 9.4-62 Table 9.4-3 Radwaste Building Area Major Components of HVAC Systems (Continued)

c. Exhaust filter units

1. Tag number WEA-FU-1A, WEA-FU-1B, WEA-FU-1C 2. Number of units 3 3. Maximum capacity (scfm) 42,000 4. Fan type Centrifugal direct drive (75 hp motor) 5. Total static pressure (in., w.g.) 7.6

6. Prefilters a) Type Replaceable media b) Efficiency Minimum 80-85% atmospheric dust spot ASHRAE Standard 52-68 c) Initial pressure drop (in., w.g.) 0.5 maximum

7. HEPA filters a) Type High efficiency b) Media Glass media, water repellent and fire resistant c) Efficiency (DOP 0.3

µ particle size) 99.97% d) Initial pressure drop (in., w.g.) Maximum 1.0

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-028 9.4-63 Table 9.4-3 Radwaste Building Area Major Components of HVAC Systems (Continued)

d. Fume hood filter units

1. Tag number WEA-FU-2A WEA-FU-2B WEA-FU-2C WEA-FU-4 WEA-FU-5 WEA-FU-6 2. Number of units 3 2 1 3. Capacity (scfm) 1850 1850 3700 4. Fan type Centrifugal direct drive Centrifugal direct drive Centrifugal direct

drive 5. Prefilter Replaceable viscous impingement type Replaceable viscous

impingement type Replaceable viscous

impingement type 6. HEPA filter Glass media, water repellent and fire resistant Glass media, water repellent and fire resistant Glass media, water repellent and fire resistant

7. HEPA filter efficiency (DOP 0.3

µ particle size) 99.97% 99.97% 99.97%

8. Total static pressure (in., w.g.) 4.5 4.5 4.5 9. Area served Fume hood radio chem labs Fume hood instrument

shop Fume hood sample

room e. Water chillers

1. Tag number WCH-CR-51A, WCH-CR-51B 2. Number of units 2 (1 standby) 3. Nominal capacity 150 tons (145 kW compressor motor) 4. Chilled water (gpm) 360

5. Chilled water supply temperature 44°F 6. Condenser cooling water source Plant service water 7. Area served Main control room, cable spreading room, health physics area, radwaste control, radio chem lab, counting room, and hot instrument shop

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-028 9.4-64 Table 9.4-3 Radwaste Building Area Major Components of HVAC Systems (Continued)

f. Offgas charcoal adsorber vault air handling units (spared in place)

1. Tag number WRA-AH-7A, WRA-AH-8A, WRA-AH-7B, WRA-AH-8B 2. Number of units 4 (2 standby)

3. Drive V-belt (15 hp motor) 4. Nominal air flow (acfm) 15,000 5. Total cooling capacity at -55°F brine (R-11) temp (MBtu/hr) 155 6. Nominal brine flow (gpm) 150 7. External static pressure (in., w.g.) 1.5

g. Brine chillers (spared in place)

1. Tag number WRE-CR-7A, WRE-CR-7B

2. Number of units 2 (1 standby)

3. Brine R-11

4. Brine flow range (gpm) 265

5. Brine inlet and outlet temperature -51°F and -55°F

6. Nominal capacity (MBtu/hr) 175 7. Condenser cooling water source Plant service water or chilled water (both physically disconnected from brine chillers) 8. Area served Offgas charcoal adsorber vault

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-028 9.4-65 Table 9.4-4

Turbine Generator Building Major Components of HVAC Systems

a. Heating and ventilating unit (evaporative air washer)

1. Tag number TOA-HV-1A TOA-HV-1B TOA-HV-2A

TOA-HV-2B

2. Number of units 2 2 3. Nominal air flow (scfm) 80,000 97,000 4. Steam flow (lb/hr) 4530 5570 5. Supply air temperature a) Summer 72°F DB (evaporative cooling) 72°F DB (evaporative

cooling) b) Winter 50°F 50°F 6. Fan type Ce ntrifugal V-belt (125 hp motor)

Centrifugal V-belt (125 hp motor)

7. Number of fans 2 (1 for each unit) 2 (1 for each unit) 8. Total static pressure (in., w.g.) 5.8 4.8 b. Exhaust fans

1. Tag number TEA-FN-1A, TEA-FN-1B, TEA-FN-1C, TEA-FN-1D

2. Number of fa ns 4 (one standby)

3. Fan type Ce ntrifugal SWSI

4. Drive Direct (200 hp motor)

5. Capacity (acfm) 117,000 6. Total static pressure (in., w.g.) 7.0

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-028 9.4-66 Table 9.4-4

Turbine Generator Building Major Components of HVAC Systems (Continued)

c. Boiler room air handling unit

1. Tag number TOA-AH-51

2. Number of units 1 3. Drive V-belt (15 hp motor)

4. Nominal air flow (scfm) 15,000 5. Steam flow (lb/hr) 980 6. Total static pressure (in., w.g.)

2.1 7. Area served Boiler room

d. Local exhaust fans

1. Tag number TEA-FN-52 TEA-FN-53 TEA-FN-2 TEA-FN-3A TEA-FN-3B 2. Number of units 1 1 1 2 3. Drive Direct (2 hp motor) Direct (3/4

hp motor)

Direct (1/4

hp motor)

Direct (2 hp

motor) 4. Nominal air flow (scfm) 1350 810 400 9000

5. Total static pressure (in., w.g.) 4 1/2 5/8 3/4 1/4 6. Area served Fume hood sample room Sample room Toilet Transformer room C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-028 9.4-67 Table 9.4-5

Diesel Generator Building Areas Major Components of HVAC Systems

a. Diesel generator building air handling units

1. Tag number DMA-AH-11 DMA-AH-21

DMA-AH-31 DMA-AH-12

DMA-AH-22

DMA-AH-32

DMA-AH-51 2. Number of units 3 3 1

3. Design air flow per unit (scfm) 36,000 15,000 a 6000 4. Sensible cooling capacity (Btu/hr) 506,000 210,000

5. Total static pressure (in., w.g.) 3.57 4.37 2.57

6. Water supply source Standby service water Standby service

water 7. Area served Electrical equipment and generator area (normally standby)

Electrical equipment

and generator area (normally standby) Corridor area

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-020,09-028 9.4-68 Table 9.4-5

Diesel Generator Building Areas Major Components of HVAC Systems (Continued)

b. Diesel generato r building exhaust fans

1. Tag number DEA-FN-11 DEA-FN-21 DEA-FN-31 DEA-FN-52 DEA-FN-12 DEA-FN-22 DEA-FN-32 DEA-FN-13 DEA-FN-23 DEA-FN-33 2. Number of fans 3 1 3 3 3. Fan type Vaneaxial Propeller Centrifugal SWSI Centrifugal SWSI 4. Drive Direct (50 hp motor)

Direct (1.0 hp motor)

Direct (0.5 hp motor)

Direct (0.5 hp motor)

5. Design air flow (scfm) 51,000 a 2,000 350 300

6. Total pressure (in., w.g.) 4.03 1.07 1.25 0.5 7. Motor type Totally enclosed air over (TEAO) class F insulated

windings, rated for a 50°C ambient temperature Open drip proof self ventilated

type class B

insulated windings with a maximum temperature rise

of 50°C above ambient Explosion proof type

with class B

insulated

windings 8. Area served Diesel generator area Corridor

area Day tank rooms Oil pump rooms a Actual air flow of the units is higher and is shown in Figure 9.4-7.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-028 9.4-69 Table 9.4-6

Standby Service Water Pump House Areas Major Components of HVAC Systems

a. Standby service water pump house supply fans

1. Tag number POA-FN-2A POA-FN-2B

2. Number of fans 2 3. Fan type and st yle Centrifugal SWSI 4. Drive Direct (2 hp motor)

5. Capacity (scfm) 5000 6. Total static pressure (in., w.g.) 1.0

b. Standby service water pump house fan coil units

1. Tag number PRA-FC-1A PRA-FC-1B 2. Number of units 2

3. Nominal air flow (scfm) per unit 17,000

4. Sensible cooling capacity (Btu/hr) per unit 404,000 5. Total static pressure (in., w.g.) 1.71

6. Water supply source Standby service water

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-028 9.4-70 Table 9.4-7

Makeup Water Pump H ouse Major Components of HVAC Systems

a. Air handling and fan coil units

1. Tag number PMA-AH-81A PMA-AH-81B PRA-FC-91A PRA-FC-91B

2. Number of units 2 2 3. Air flow per un it (acfm) 8000 16,500 4. Sensible coo ling capacity per unit (Btu/hr) 221,000 474,000 5. Total static pressure (in., w.g.) 2.57 2.19 6. Fan motor (hp) 7 1/2 10 7. Area served Electrical equipment area Pump area

8. Water supply service Cooling tower makeup water Cooling tower

makeup water

b. Exhaust fans 1. Tag number PEA-FN-81A PEA-FN-81B PEA-FN-82 2. Number of units 2 1 3. Air flow per unit (scfm) 150 100

4. Total static pressure (in., w.g.) 0.625 0.375 5. Area served Battery hood exhaust Toilet room 6. Motor (hp) 1/6 1/8

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-028 9.4-71 Table 9.4-8

Circulating Water Pump House Major Components of HVAC Systems

a. Ventilators

1. Tag number PEA-FN-51 PEA-FN-52 PEA-RVT-11 PEA-RVT-12 PEA-RVT-13 PEA-RVT-14 PEA-RVT-15 PEA-RVT-16

2. Number of units 1 1 6 3. Rated air flow, (scfm) 4800 250 33,000 4. Total static pressure (in., w.g.) 0.5 1/8 1/8

5. Fan (hp) 2 1/12 7-1/2

b. Air conditioning unit

1. Tag number PMA-AC-51

2. Number of units 1

3. Rated air flow (scfm) 2000 4. Cooling capacity (Btu/hr) 52,000 5. Heating capacity (Btu/hr) 66,000

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORTDecember 2009LDCN-09-028 9.4-72 Table 9.4-9 Plant Heating Steam System

a. Unit heaters

1. Tag number TRA-SUH-1 TRA-SUH-2 SRA-SUH-1 WRA-SUH-1 WRA-SUH-3 WRA-SUH-4 WRA-SUH-5

2. Number of units 1 1 1 1 3 3. Heating capacity (mbh) 1410 708 915 1265 204 4. Entering air temperature (°F) 50 50 65 65 65 5. Nominal flow (cfm) 25,000 12,000 15,400 25,000 2250 6. Steam flow (lb/hr) 1540 775 1000 1490 224 7. Area served Turbine building railroad door Turbine building

truck door Machine shop

truck door Radwaste building

truck door Radwaste building mechanical equipment room and demineralizer removal area

b. Steam humidifiers

1. Tag number WOA-HU-4 WOA-HU-5 WMA-HU-6 WOA-HU-9 2. Capacity (lb/hr) 200 95 35 75

3. Area served Radio chem lab Hot instrument shop Radwaste control room Counting room

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORTDecember 2009LDCN-09-028 9.4-73 Table 9.4-9 Plant Heating Steam System (Continued)

c. Condensate pumps 1. Tag number SHCO-CU-1 WHCO-CU-1

2. Unit capacity 15,000 80,000 3. Pumpset capacity (gpm) 22.5 120 4. Discharge pressure (psig) 20 25 5. Area served Service building Radwaste building

d. Heat transfer package 1. Tag number SHHW-HX-51

2. Heating capacity (mbh) 1800 3. Steam flow (lb/hr) 1980 4. Water flow (gpm) 190

5. Inlet water temperature 161°F 6. Outlet water temperature 180°F 7. Area served Service building

e. Reheat Coil 1. Tag number ROA-HC-2 2. Air flow (cfm) 33,200 3. Steam flow (lb/hr) 2560 4. Air temperature in 50°F

5. Air temperature out 115°F 6. Area served Reactor building

Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.4-1.1 101 M548-1HVAC for Control and Switchgear Rooms Radwaste BuildingRev.FigureDraw. No.Amendment 61December 2011 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.4-1.2 6 M548-2HVAC for Control and Switchgear Rooms Radwaste BuildingRev.FigureDraw. No.Amendment 61December 2011 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.4-2.1 71 M545-1HVAC Systems - Reactor BuildingRev.FigureDraw. No.Amendment 61December 2011 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 09.4-2-2 12 M545-2HVAC Systems - Reactor BuildingRev.FigureDraw. No.Amendment 60December 2009 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 09.4-2.3 25 M545-3HVAC Systems - Reactor BuildingRev.FigureDraw. No.Amendment 60December 2009 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.4-3.1 58 M549-1HVAC Systems - Radwaste BuildingRev.FigureDraw. No.Amendment 61December 2011 Amendment 59December 2007 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.4-3.2 2 M549-2HVAC Systems - Radwaste BuildingRev.FigureDraw. No.

Amendment 59December 2007 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.4-4.1 36 M550-1HVAC Chilled Water System - Radwaste BuildingRev.FigureDraw. No.

Amendment 59December 2007 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.4-6.1 54 M546-1HVAC Systems - Turbine Generator BuildingRev.FigureDraw. No.

Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 09.4-7 61 M551HVAC - CW, SW, and MUW Pump Houses and Diesel Generator BuildingRev.FigureDraw. No.Amendment 60December 2009 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.4-8.1 86 M543-1Flow Diagram - Reactor Building Primary Containment Cooling and Purging SystemRev.FigureDraw. No.Amendment 61December 2011 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.4-8.2 11 M543-2Flow Diagram - Reactor Building Primary Containment Cooling and Purging SystemRev.FigureDraw. No.Amendment 61December 2011 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.4-8.3 4 M543-3Flow Diagram - Reactor Building Primary Containment Cooling and Purging SystemRev.FigureDraw. No.Amendment 61December 2011 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.4-9.1 39 M514-1Heating Steam System - All BuildingsRev.FigureDraw. No.Amendment 61December 2011 Amendment 59December 2007 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.4-9.2 5 M514-2Heating Steam System - All BuildingsRev.FigureDraw. No.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 LDC N-0 2-0 0 0 9.5-1 9.5 OTHER AUXILIARY SYSTEMS

9.5.1 FIRE PROTECTION SYSTEM

Appendix F , Fire Protection Evaluation, contains information on the plant fire protection systems.

9.5.2 COMMUNICATIONS

SYSTEMS

9.5.2.1 Design Basis Columbia Generating Station is provided with the follo wing communication systems:

a. Public telephone access,

b. Private branch exchange (PBX),

c. Sound-powered telephone system,

d. Public address and bu ilding-wide alarm system,

e. Radio communication systems, and

f. Telephone link to Bonneville Power Administration (BPA) Dittmer Control Center.

The public telephone network provides for connection of PBX phones to outside lines.

The PBX provides intraplant communications, co mmunication to BPA, a nd access to the public address and radio paging systems.

The sound-powered telephone system provides a backup to the PBX and can be used as a

supplementary communications circ uit to aid in the testing and maintenance of plant process systems.

The public address system provides a way of c ontacting personnel in the various buildings of the plant and locations of the site that might be inaccessible using other means of communication. The building-wide alarm syst em alerts (via the public address system speakers) operating personnel to fire hazards and other trouble cond itions for which plant management finds it necessar y to alert plant personnel.

The radio communications system provides a co mmunications link for security and emergency communications to local law enforcement agencies and emergency control centers. In C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 LDC N-0 0-0 5 5 9.5-2 addition, the radio communications system is us ed for communications w ith personnel involved in maintenance and security in and around the plant complex by means of hand-held portable radio units, mobile radio units, and paging receivers. The te lephone link to BPA provides a direct communication link to the BPA Dittmer Control Center.

Post-fire safe shutdown operator actions that require emer gency communication with the control room or remote shutdown room rely on the PBX phone system. The post-fire safe shutdown communication system mu st remain operational during a loss of offsite power. Fire brigade activities use handheld-to-handheld radio communication and may communicate with the control room using a PBX phone.

All equipment, components, raceways, and support systems located in the main control room are designed and anchored such that they will not cause loss of function of nearby safety-related equipment as a result of the safe shutdown earthquake.

The same criteria is followed wherever such equipment is located close to safety-rel ated equipment.

9.5.2.2 System Description

9.5.2.2.1 Public Telephone Access

This system consists of inte rconnections to the public telephone network as described in the following:

a. Individual direct lines with inward a nd outward dialing access to several plant locations:

1. Plant Manager's office,

2. Main control room,

3. Security central alarm station (CAS),

4. Remote shutdown room, and

5. Security secondary alarm station (SAS).

b. Provisions for extension of individual lines with direct inward and outward dialing access to other plant locations:

1. Shift Manager's office and

2. Radwaste control room.

c. Trunks to the PBX provide inward and outward dialing access to various plant locations.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-07-020 9.5-3 9.5.2.2.2 Private Branch Exchange

This system consists of an all electronic, stored program, computer controlled telephone switching system with integral redundant computers. Telephones are strategically located

throughout the plant complex. The system re ceives power from an uninterruptible power supply (UPS) system for reliable operation, c onsisting of a battery E-B0-PBX and battery chargers E-C0-PBX/1 and E-C0-PBX/

2, located in building 25 (PAAP).

The PBX provides complete inte rcommunication at all times between any two telephones. Connections are established by means of a dial on each telephone. Automatic dial tone, busy tone, and ringing current are provided.

Telephone communications boxes (CBs) are provided throughout the plant. The CBs in offices have a jack for plugging in a desk-type telephone. These CBs are ex tensions of the PBX exchange.

Most CBs in the operating and work areas ha ve a PBX telephone jack and a sound-powered telephone jack. Instrument a nd control panels in the cont rol room have PBX and sound-powered telephone jacks installed in the panels. Operating and work areas use portable

telephones plugged into the jacks.

A telephone number is assigned to one CB or jack of a compatible group and the ot hers in the group are wired in parallel to it.

Head sets may be plugged into telephones in the operating a nd work areas when hands-free communication is required. Approved wireless communication devices, such as cordless handsets, may be used in the plant main control room and radw aste control room. To support outage activities approved wirele ss communication devices can be used in the drywell and approved reactor building elevations. Administrative control of wireless devices based on the location, operating frequency, fi eld strength, minimum distance of an exclusion zone, visual postings, and end user trai ning are established to preclude the effects of EMI.

Dedicated communication links provide access to BPA. The PBX provides the following special features:

a. Conference - up to eight telephones can be connected into a conference network to facilitate maintena nce, testing, and ma nagement activities.

b. Paging - the Public Address System can be accessed from any PBX telephone extension.

c. Ringdown - the PBX can be programmed to provide ringdown (hotline) service for selected telephones. The service provides automatic ringing without dialing.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 9.5-4 9.5.2.2.3 Sound-Powe red Telephone System The sound-powered telephone system consists of jack access in the communications boxes and panels and connecting wiring. The system is divided into eight circuits. Each circuit serves a different area of the plant.

All wiring is routed to a terminal box located in the communications equipment room. The term inal box is equipped with jumpers for interconnecting the circuits. During normal ope ration the jumpers are connected to form a single bus so that all sound-powe red jacks are connected in para llel. Each circuit can be isolated at the terminal box if shorts or grounds occur.

Portable sound-powered telephones are plugged into the jacks to complete a communications link.

9.5.2.2.4 Public Address and Building-Wide Alarm Systems

9.5.2.2.4.1 Public Address System. The public address system is designed to provide area-wide announcements throughout the plant by means of loudspeaker s located in various areas. Audio power to the speakers is provided by electronics located in equipment racks in the communications equipment room, which is locate d in the radwaste and control building. Power to the electronics is provided from the UPS bus.

The speakers in ea ch of the buildings or zones are conn ected in two separa te circuit loops.

This provides an alternative pa th for partial communications should one loop be damaged.

Paging microphones are located in various areas and are used for "all zone" pages and have priority over telephone-accessed paging.

Most telephone instruments in the PBX can access the paging system by dialing the appropriate access number. Telephone pages can be directed to a specific building/zone or to all zones.

The connection between the PBX and the paging system is provided through special paging adapters which interface the two systems.

Critical electronic equipment is redundant and is automatically switched on line as required. Failure of the amplifier systems is alarmed in the control room as well as locally at the public address racks.

9.5.2.2.4.2 Building/

Zone-Audio Alarm System. The audio alarm system consists of a multitone generator with redundant backup which is located with in the public address (PA) system racks. The audio genera tors are capable of producing fi ve distinct audio tones which are amplified by the PA system. The redundant generator is automa tically switched online should the primary unit fail and an alarm is activated. The units are fed from the UPS system along with the PA system. One tone is us ed to alert personne l, followed by specific announcements. The audio alarm system has priority over PBX paging and "microphone actuated paging," and will override either of these functions.

The main control room is C OLUMBIA G ENERATING S TATION Amendment58 F INAL S AFETY A NALYSIS R EPORT December2005 9.5-5 capable of activating all tones th at ar e used. The alert tone may be activated from either the Plant Manager's office or the remo te shutdown room or the TSC.

9.5.2.2.5 Radio Comm unications System

The radio communications system for Columbia Generating Station is integrated into the Energy Northwest radio network, which provides communications for all Energy Northwest facilities in th e Hanford area.

The system is compri sed of the following:

a. Multiple (at least three) in-plant repeater stations for operations and security within the plant, b. Radio base stations,

c. Portable handheld radios, and

d. Mobile radios in stalled in vehicles.

In-Plant Repeaters

The in-plant repeaters are used to provide radio communications with in the plant using a distributed antenna system. The units are phys ically located in the radwaste building and separated from other co mmunications systems.

Base Station Equipment

The base stations are duplicate units with the same frequencie s on two channels and the same output power. Two 2-way communi cations channels are provided. One channel is used for operations and maintenance a nd the other for security.

One base station is located in building 62 a nd controlled by the CAS radio console. The antenna is also lo cated on building 62.

The second base station is located in the radwas te building HVAC room

5. It is controlled from the SAS and the TSC. This station is capa ble of being used as a backup to the Security Base in building 62. The SAS may also prev ent the CAS from controlling this station if required in an emergency. Two radio receivers are also provided. They receive on the two security channels. One is located in building 62 with sp eaker and controls on the CAS console, and the other is located in the radwaste building with controls on the SAS console. These receivers allow secur ity to monitor both channe ls at once if required.

C OLUMBIA G ENERATING S TATION Amendment58 F INAL S AFETY A NALYSIS R EPORT December2005 9.5-6 All equipment other than mobiles or handheld portables are powered by a UPS bus for reliable operation.

9.5.2.2.6 Telephone Link fo r Connection to the BPA Dittmer Control Center

This circuit consists of telephones located in the main c ontrol and remote shutdown rooms of the radwaste and control building, which are di rectly connected to th e Columbia Generating Station/BPA communications equipment. Th ese phones automatically ring to the Dittmer Control Center of BPA when they are used.

9.5.2.3 Inspection and Testing Requirements

All communication cable conductors are tested for continuity and insulation resistance before connection to the various communi cation apparatus. A functional test on all communications systems is made after installation.

The functional test on the total installed radio communications system included a complete test of all system functions such as operation from base station, remote cont rol units, and two-way communication between offsite and onsite stations. Tests were made using both the maintenance and security system frequencies. Tests were performed to ensure that no harmful interference results between this equipment, the repeater st ation on Rattlesnake Mountain, and control room equipment. Duri ng preoperational and pos toperational surveilla nce testing, solid state electronics in some areas exhibited spurious response to ra dio transmission. These areas have been flagged by signs s howing "Prohibited Area For Use of Radios" and are put under administrative control.

9.5.2.4 Capability During Postulated Accident and Anticipated Transients

Table 9.5-1 shows the strategic work areas and the type of comm unications available between these locations and the control room, remote shutdown room, and outside Columbia Generating Station.

9.5.2.4.1 Protective Measures

The plant communications system provides assu rance that a reliable communications link is available from the strategic plant work stati ons to the control room. The following design factors contribute toward this goal:

a. The PBX is built with pl ug-in modular circuit boards so that defective boards can be quickly replaced,

b. The PBX has redundant pro cessing units with automatic switching from a failed computer to the redundant computer, C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 9.5-7 c. The sound-powered telephone system b acks up the PBX system in case of a major failure of the PBX,

d. Several telephones are connected direc tly to the public telephone network to operate independently of th e PBX. These include telephones in the main control room, remote shutdown room, building 62, and the Plant General Manager's office, and

e. During a loss of offsite power, the PEC/PAAP diesel generator ensures the PBX system operation beyond the 8-hr capaci ty of the PBX communication system battery, E-B0-PBX.

9.5.2.4.2 Severing of Lines or Trunks

The failure of a single line or trunk cannot prevent communications from critical plant locations. Alternate facilities are available as follows:

a. Radios do not require offsite lines a nd can be used in lieu of telephones if telephone lines are not available,

b. Portable radios can be used if the ante nna leads to fixed stations are inoperable, and

c. Tie lines connect the PBX to the BPA communications system. The lines provides telephone communications from Columbia Generating Station to

telephones in BPA facilities. These co mmunications are independent of the public telephone network.

9.5.2.4.3 High Noise

The communication system design addresses high noise area problems as follows:

a. Telephones may be installed in sound dampening booths with either or both noise-canceling handsets and amplified receivers, and

b. Public address system speakers have volume controls which are adjusted according to the ambient noise level.

9.5.2.4.4 Post-Fire Safe Shutdown

Fire induced failures of the PBX communication system are evaluated for the various fire areas as part of the post-fire safe shutdown analysis. See F.2.6.2 for more detail.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-06-012 9.5-8 9.5.3 PLANT LIGHTING SYSTEM

9.5.3.1 Design Bases

a. Lighting intensities are designed to provide indoor and outdoor illumination consistent with the Illumination Engi neering Society recommendations (July 1974), and to meet or exceed OSHA requirements,

b. Light sources are selected with consideration for environmental conditions and ease of maintenance, c. Fluorescent or high-intensity discharge (HID) sources are not used inside primary containment. On ly incandescent sources ar e used inside primary containment, d. The normal lighting supplied from either the main generator or offsite power is designed to provide adequate lighting for normal plan t operation and associated plant access routes, and for control and maintenance of equipment, and

e. The emergency lighting supplied from onsite power is designed to provide adequate lighting for safe shutdown of the plant and for associated plant access routes during the full spectrum of accidents, transients and special events.

9.5.3.2 System Description

The plant lighting system consists of four sy stems: normal ac lighting, normal-emergency ac lighting, emergency dc lighting, and battery-p ack emergency lighting. For the location of various fixed emergency lighti ng systems that support safe shutdown of the plant, see Table 9.5-2. See Figures F.6-18 , F.6-19 and F.6-20 for locations of portable lantern use.

9.5.3.2.1 Normal Alternating Current (ac) Lighting Systems

This system consists of two systems (A and B) which are energized fr om the plant non-safety-related 480-V ac auxiliary system motor control centers directly from th ree-phase 480-V ac, or through 208/120-V ac dry type lighting transf ormers and local ar ea lighting panels.

Fluorescent, incandescent, and HID sources are used for the normal ac lighting system.

9.5.3.2.2 Normal-Emerge ncy ac Lighting Systems

Normal-emergency ac lighting is provided for safe and orderly shutdown during the loss of normal ac power. This system is energized from the safety-r elated 480-V ac motor-control centers through isolation devices to non-Class 1E three-phase, four-wire 208/120-V ac dry type lighting transformers that feed lighting panels.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-009 9.5-9 This lighting system consists of two systems (D ivisions 1 and 2). Each system has ac lighting energized from critical buses that are connected both to and back ed by their associated diesel generators (DG1 & 2).

This ac lighting comprises approxi mately 15% of the normal plant lighting load and consists of fluorescent, sodium (outdoor), and incandescent sources. The ac li ghting fixtures in the main control room are designed and su pported as Seismic Category I.

Some of the ac lighting system within the main control room can be manually transferred from its critical bus to an inverter supplied source if the associated battery charger is in service.

9.5.3.2.3 Emergency Direct Cu rrent (dc) Lighting Systems

Emergency dc lighting is provided in the main control room, the access route to the remote shutdown room, and in the remote shutdown room. The emergency dc lighting system is necessary for conducting shutdown operations during the coping period of a Station Blackout event.

This lighting consists of two systems (Divisions 1 and 2). Each system is energized from a 125-V dc plant emergency battery system.

The remote shutdown ro om and access route lighting is continuously lit, and the control room dc lighting is energized on a loss of power to the normal-emergency lighting in the main control room. This dc lighting load consists of incandescent light sources.

9.5.3.2.4 Battery-Pack Emergency Lighting Systems

Battery-pack emergency lighting consists of 1.5-hr and Appendix R lighting units. The lamps automatically energize on loss of the respective normal and/or no rmal-emergency ac lighting.

Emergency battery units in Seismic Category I areas of the radwaste building are Seismic Category I mounted and supported.

The 1.5-hr lighting consists of Emergency Battery Units (EBUs) and Emergency Reserve Ballast (ERB) units in areas without EBUs, installed for th e safety of opera ting personnel to provide lighting for egress routes.

Appendix R 8-hr emergenc y battery packs and portable 8-hr lanterns are available for use to ensure operators can perform pos t-fire safe shutdown manual actions outside the control room, for a fire in specific plant area s, concurrent with a LOOP.

At a minimum, 8-hr emergency battery pack lighting is installe d at locations where operators perform long-term plant control and shutdown actions. Th e portable 8-hr lanterns are used for short-duration operator actions.

There are four stations provided in various plant locations with five portable lanterns available at each location.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-009 9.5-10 9.5.3.3 Safety Evaluation

Normal and normal-emergency lighting is locate d throughout the plant to provide necessary lighting during normal plant operation and during a power system failure. In the Reactor Building, during accident conditions, the normal lighting is tripped by an accident signal (FAZ) to reduce building heat load.

In areas that are serviced by HID lighting so urces, 10% of the fixtures are provided with tungsten halogen standby lamps. These standby lamps will be immediately energized after an

interruption of the power source and will provide partial illumination until the HID lamps restrike.

The normal and normal-emergency lighting provide no safety-rel ated function and thus are non-safety related.

9.5.4 DIESEL

GENERATOR FUEL OIL STORAGE AND TRANSFER SYSTEM

9.5.4.1 Design Bases

a. The onsite storage capacity of each subsystem provide s for continuous operation of each diesel generator for at least 7 days while satisfying post-LOCA maximum load demands;

b. The design of the system conforms to Regulatory Guide 1.137. The equipment within the system conforms to the a pplicable codes and standards of ASME, ASTM, ANSI, DEMA, IEEE, API, and NFPA;

c. The system piping off of the engine skid is constructed to ASME Section III, Class 3 and Seismic Category I requirem ents except for the portions of the piping associated with the filter/polisher system. Except for the diesel oil storage tanks and portions of the filter/polisher system, all portions of the system, including the fuel oil day tanks, are protected from tornado missiles by enclosure in tornado-hardened, Seismic Category I structures. The diesel oil storage tanks and associated piping (except for exposed fill, vent, and filter/polisher piping as described in Section 3.5) are buried for tornado protection and to maximize c ontainment of postulated o il spills. The system is not subject to flooding since the site is not subject to flooding. The piping on the engine is constructed using ANSI B31.1 as a guide and is Seismic Category I;

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-07-047 9.5-11 d. The diesel fuel storage system is augmented by the installation of a filter/polisher unit located in a separate filter/po lisher building east of the storage tanks. This unit is non-safety-relat ed. It is used to clean bulk fuel of contaminants and fuel degradation products formed due to long storage. It is also used to polish fuel o il delivered by fuel supply ta nkers, if necessary, prior to delivery to the storage tanks. The filter vessels are constructed to ASME Section VIII, Division I, and the piping components to ANSI B31.1 Power Piping Code.

e. The Quality Class II Auxiliary boiler storage tank is used as an additional storage tank for the emergency diesel gene rator fuel oil. This tank was cleaned before putting it into diesel fuel storage service and is maintained to the same cleanliness requirements as th e other Class I fuel oil tanks. The diesel fuel oil stored in this tank is surveyed to th e same requirements as the other diesel storage tank's fuel oil. This stor age tank and its connective piping are not required to be Seismic Category I, since this additi onal qualified fuel oil is non-safety-related.

9.5.4.2 System Description

The fuel oil storage and transfer system consis ts of separate , independent di esel oil supply systems serving each of the two tandem diesel generators (DG 1 and DG 2) and the high-pressure core spray (HPCS) diesel engine generator (DG 3). Each of these systems consists of a fuel oil storage tank, a transfer pump, a da y tank, interconnecting pi ping and valves, and associated instruments and controls.

The system diagram is shown in Figure 9.5-1.

In each supply system, a transfer pump powered from a safety-related motor control center (MCC) takes suction from the dies el oil storage tank and discharges to an associated diesel generator fuel oil day tank to maintain the fuel oil level within the da y tank. The transfer pump is sized to provide a flow several times the maximum engine cons umption rate (of 110%

load) and is automatically controlled by level switches activated by day tank fuel level. The fuel stored in each diesel ge nerator 60,000-gal storage tank (50,000 gal for the HPCS diesel generator) plus the fuel maintained in each 3000-gal day tank is adequate for 7 days of operation.

Each transfer pump supplies a separate day tank for each diesel generator. There is a pipe interconnecting the diesel generator 1 and 2 transfer lines with normally locked closed valves.

If a rupture occurs in the inte rconnecting cross line when in use, this line will be manually isolated and thereby the fuel oil supply will not be interrupted between any storage tank and its associated day tank.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-06-002 9.5-12 The volume of the day tanks permits 8.5 hr of engine operation of the associated diesel generator without resupply to the day tank. Th is arrangement allows the transfer pump start signal to occur prior to the day tank low level alarm. In the event the transfer pump does not start, there remains at least 3.

5 more hours of operation after lo w level alarms are actuated to take required corrective action. These operating times are for Division 1 and Division 2 with the diesel generators operating at 100% of th e continuous rated lo ad of 4400 kW. The maximum operating time for Division 3 (HPCS) will be substantially longe r since the day tanks are the same size. The single HPCS engine ca rrying 2600 kW will use s ubstantially less fuel than two engines carrying a load of 4400 kW.

At normal high oil level a switch will shut off the pump. If the fuel oil level goes above the normal high level, a high level switch is activated and sends an alarm signal. If the pump does not shut off, the day tank overflow line will retu rn all pump flow to the storage tank of the same diesel generator.

The fuel oil storage syst em has provisions to fill the storage tanks, transfer fu el from one tank to another, and filter/polish the fu el in the tanks or fuel being de livered to the tanks. The fill header is designed to allow de livery of fuel directly to a storage tank or run through the filter/polisher unit prior to deliv ering it to the tank from the fu el tanker. Th e piping boundary between the ASME III fill lines a nd header and the Quality Class II filter/polisher are at ASME III, 3 normally clos ed butterfly valves.

The filter/polisher, when used to polish the fu el stored in the tanks , removes oil from the bottom of the tank through a suction sparger that extends along the length of the tank. This oil is filtered and polished removing water and particulates. Then the filter/polisher returns the oil to the tank through the fill line.

The filter/polisher can be used to transfer fuel from one tank to another with or without filtering and polishing. The filter/polisher is equipped with a record ing flow meter so the operator knows the quantity of fuel that has passed through the filter/polisher.

The filter/polisher suction line from the storage tank flanged nozzle to the filter/polisher suction header isolation valves is ANSI B 31.1 Quality Class 2, Se ismic Category I. The sparger in the tank is ANSI B31.1, Quality Class 2, Seismic Cate gory I, and the fill line returning oil to the tank is ANSI B31.1, Quality Class 2, Seismic Category II. The suction and fill line for the auxiliary boiler storage tank is ANSI B31.1, Qua lity Class 2, Seismic Category II.

Operation of the fuel storage tank transfer pump is controlled manually when fuel is being transferred through the interconnecting line from storage tank A to day tank B or from storage

tank B to day tank A. High level annunciation of the day tank fuel level will provide a warning of overfilli ng the day tank.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-020 9.5-13 The fuel oil supply from the day tanks to each diesel engine cons ists of two systems. Either system is capable of supplying fuel oil to the engine. For diesel genera tor 1 and 2 each system contains a fuel supply line stra iner, fuel oil pump, duplex filte r, pressure gauge, relief and check valves, and separate fuel return lines to the day tanks. The HPCS diesel generator fuel oil system is slightly different, see Section 8.3.1.1.7.2.6 for a description of the HPCS diesel generator fuel oil system.

One of the fuel supply pumps is mechanically driven by the engine and is normally used during engine operation. The other supply pump is driven by a 120-V dc motor and is used to fill the fuel oil system and fuel header prior to initial operation and after maintenance has been performed on system piping a nd components. The dc-motor-driven pump is running during engine operation in the event fuel supply th rough the engine-driven pump system fails.

The fuel pumps are located 2.3 ft higher than th e suction pipes inside th e day tank. The fuel pumps are designed to operate at this negative suction pressure.

The fuel oil supply and return piping is not exposed to ignition sources such as open flames or hot surfaces. The transfer lines between the storage and day tanks are buried.

The fuel oil day tank is located in a separate ventilated room wh ich is sized to contain the full contents of the tank should a l eak develop. For discussion of fire protection see Section 9.5.1.

The tanks are filled through individual lines from a common fill header th at has a strainer and locked fill connection. Each tank also has an individual vent with a flame arrestor.

The fill and vent lines terminate at 2.33 ft a nd 8.0 ft, respectively, a bove plant grade, which prevents direct seepage of any gr ound water into the storage tanks.

In the event of fill line damage due to a missi le, the pump-out connecti on which is protected by a metal enclosure, located at ground level, ma y be used for the fuel oil filling operation.

Prior to filling the storage tanks, the day tank w ill be confirmed to be full. This will allow sufficient time for sediment to settle before oil from the storage tanks is transferred to the day tank.

Diesel fuel oil conforming with the requirements of the Technical Specifications is provided for operation of the diesel generato rs. This grade of diesel fu el complies with the engine manufacturer's requirements and is available from local distribution sources as discussed in Section 9.5.4.3.

Equipment design characteristics for the fuel oil supply system are shown in Table 9.5-3.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-09-038,10-031 9.5-14 9.5.4.3 Safety Evaluation

The entire diesel oil supply syst em is located within the confines of a Seismic Category I building except for the buried storage tank and the fill, vent, and filter/polisher piping. Each system oil storage tank transfer pump, day tank, and dies el generator set is physically separated within separa te concrete enclosures designed to protect against missiles, in compliance with Section 3.5, and to provide fire protection. No high- or moderate-energy piping is present in the diesel generator buildin g which could present a po tential hazard to the operational function of these systems. The oil storage tanks are buried for protection so that storage tank failure is completely contained within the soil at a level below any building

penetration or access opening. Each storage and day tank is provided with a vent directly to the outside atmosphere. In addition, the enclosures are provided with exhaust ventilation to the outside atmosphere to ensure that diesel fuel vapors are mainta ined well below the combustible limit. The enclosures are auto matically monitored by temperatur e detectors which initiate the preaction sprinkler system in the event of fire (Appendix F

). All storage and day tank vents are equipped with flame arrestor devices.

Although a single failure may result in loss of fuel to one diesel genera tor, the other diesel

generator can provide sufficient capacity for emergency conditions , including safe shutdown of the reactor (see Section 8.3) coincident with lo ss of offsite power.

The fuel level in each storage tank is maintained above the le vel at which the fuel in the storage tank plus the fuel in the corresponding day tank is adequate fo r 7 days of continuous operation at 4400 kW for diesel generators 1 and 2, and 26 30 kW for diesel generator 3 (HPCS).

The minimum site storage of 7 da ys is considered adequate ti me for obtaining additional fuel oil, if required. The auxiliary boiler storage tank is available as an additional source for qualified diesel fuel oil.

On an expedited basis fuel can also be available at the site from a remote source within 12 to 24 hr.

Materials for the fuel oil s upply system are provided in Table 9.5-3.

For corrosion protection, the exterior surfaces of the buried piping and components are coated with coal tar enamel. Application of coa tings are in strict accordance with AWWA Standard C203.

The buried components of the fuel oil system ar e all at a relatively uni form temperature and not subject to condensation phenomena. The periodic sampling of the diesel fuel storage tank bottom below the transfer pump will be performe d to determine if any water has accumulated and if so corrective action will be taken.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 9.5-15 A fuel oil filter and strainer is provided on each fuel line to each engine to eliminate passage of particles, 12 or larger in size to the engine injectors.

The overflow lines from the day tank to the stor age tank run underground south of the diesel generator building. Diesel oil pipe lines extending under the diesel generator building do not

receive full protection from the exterior rectifier-anode system because of the electrical shielding effect of the ground grid and foundati on reinforcing and structur al steel. Since the earth area under the diesel generator building is sheltered and hence relatively much drier than the earth exterior to this bu ilding, no additional cathodic protection system is provided or required.

9.5.4.4 Testing and In spection Requirements

System components are in spected and cleaned prio r to installation. In struments are calibrated during testing and automatic controls are tested for actuati on at the proper setpoints. Alarm functions are checked for operab ility and limits during plant preope rational testing. Automatic actuation of system components is tested pe riodically in accordan ce with the Technical Specifications and plant procedures. The system is operated and tested initially with regard to flow paths, flow capacity, and mechanical operability (see Chapter 14

).

Operability of the di esel fuel oil system is ensured by

a. At least once per 12 years: Draining each fuel oil storage tank, removing the accumulated sediment, and cleaning the tank using a sodium hypochlorite or equivalent solution, and

b. At least once per 10 years: Performing a pressure test of those portions of the diesel fuel oil system de signed to Section III, Subsect ion ND of the ASME Code in accordance with ASME Code Section XI, Article IWD-5000.

New and stored fuel oil is sampled and tested in accordance with the Technical Specifications.

9.5.4.5 Instrumentation Requirements

Each diesel oil storage tank is provided with local level i ndicators and high and low level switches which actuate alarm annunc iators in the main control r oom. In addition, there are system fault annunciators in th e main control room for each di esel oil storage tank that will alert the operator of a malfunction of the level in strumentation for a given storage tank. Each day tank is provided with two level switches and a float swit ch which perform the following functions:

a. Start and stop the transfer pump to maintain level in the day tank, b. Actuate an alarm in the main control room on low level, and

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 9.5-16 c. Actuate an alarm in the main control room on high level.

The fuel oil filter/polisher is equipped with a flow totalizer to aid in maintaining minimum fuel quantity in each storage tank when transferring from tank to tank. The filter/polisher is equipped with instrumentation and a level switch in a sump to alarm in the main control room if a malfunction occurs or fuel oil leakage occurs in the filter/polisher building. The filter/polisher control panel contains a bargraph/digital level indicator for each diesel oil storage tank. Each bargraph drives a visual and audible annunciator that alerts the operator of a high level condition in that tank. An outside alarm is provided at the filter/polisher facility to alert the operator if any storag e tank has a high level condition.

A remote shutdown switch is provided in the control room for the operator to shut down the f ilter/polisher. An emergency shutdown switch is also located just outside of the door of the filter/polisher building.

Each transfer pump discharge line is provided with local pressure indicators. The system maintains the proper supply of diesel oil in each day tank by means of the level switches in the day tanks which signal the corresponding pump motor starters to automatically start and stop the transfer pumps.

Each diesel generator local control panel is provided with a control switch for control of its respective transfer pump.

Local indication of differential pressure is provided across the duplex filters of the fuel oil

supply lines to the diesel engines of diesel genera tors 1 and 2.

9.5.5 DIESEL

GENERATOR COOLING WATER SYSTEM

9.5.5.1 Design Bases

a. The diesel generator cooling water sy stem is designed to provide full load cooling for the diesel generator engines while they are operating and to maintain each engine at an acceptable star ting temperature under standby conditions;

b. The piping system associated with the diesel cooling water system was constructed to the guidelines of ANSI B31.1. Seismic classification of the system is Seismic Category I as discussed in Section 3.2.4. The diesel cooling water heat exchangers for diesel generators 1A and 1B are designed and built in accordance with ASME Section III, Class 3 and TEMA Standards, Class C; diesel generator 1C (HPCS) cooling water heat exch anger is designed and built in accordance with ASME Section VIII, TEMA Standards, Class C; and

c. The reliability of the diesel cooling water system is achieved by providing separate cooling water systems for each diesel generator.

Thus, failure of a single component in one cooling water sy stem would not effect the operation of

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-09-038 9.5-17 the other diesel generator. In addition the systems are housed separately in a

Seismic Category I structur e containing no high- or moderate-energy piping which could present a potential hazard to the operational function of these systems.

9.5.5.2 System Description

Each diesel generator is serviced by an independ ent cooling system. These systems are located in separate rooms associated with their respective diesel generator. Ea ch engine cooling water system is a closed water circuit which recirculates treated water (see Section 9.2.3) for engine cooling (see Figures 9.5-1 and 9.5-3). The treated water is circulated through the water-jacketed components of the engine to remove heat from the engine parts.

This jacket water heat is rejected through a sh ell and tube heat exchanger to the standby service water (SW) system (see Section 9.2.7). Location of the system is shown in Figure 9.5-2.

The forced circulation of cooli ng water through the engine, lube oil cooler, heat exchanger, and heat exchanger bypass circu it is maintained by two engine-driven pumps. The separate bypass piping flow paths are provided to bypass the heat exchanger at low engine outlet temperatures and to heat the jacket water system during sta ndby. The heat exchanger bypass flow and temperature is automatically regulated by a three-way self-contained thermostatic valve. This valve is set to ma intain the engine outlet water temperature at about 180°F. This thermostatic valve outlet opens to the heat exchanger when the engine jacket water temperature reaches 165°F and is full open to the heat exchanger at 180°F. A high temperature alarm annunciates at 200°F (195°F for HP CS diesel engine). A high temperature shutdown switch is provided to shut down th e engine when coolant temperatur e reaches 208°F (205°F for HPCS diesel engine) during test conditions.

The heat exchangers are designed for 110% co ntinuous rating of each di esel generator using 95°F service water to the heat exchanger. Th e maximum service water temperature is always well below 95°F. This provides a 10% margin in the size of the heat exchangers.

The only time the diesel generators are run at 110% of their conti nuous rating is during required surveillance testing.

This is usually for 2 hr once a year, and always less than 10 hr/year, compared to the 2 hr/day as allowed by the engine manufacturer's rating.

The diesel generator heat exchangers are designed to perform as given in Table 9.5-4. An expansion tank is mounted on the diesel engine skid, lo cated above the cooling water circulating pump suction. The expansion tank is provi ded with a pressure cap that maintains pressure on the cooling water system (7 psi) an d prevents loss of water due to evaporation.

The expansion tank is provided with a level sight glass which is mounted on the front with instructions that indicate mini mum water level. An alarm is provided in the control room to

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 9.5-18 annunciate in case of low wa ter level. Makeup water is normally supplied by the demineralized water system, but a Seismic Category I, Safety Cl ass 3, makeup water line from the SW system is provided as an alternate supply to the expansion tank.

Diesel generator unit reliability, including the fu nctions required of the circulating water pump and expansion tank were demonstrated prior to installation (qualification and shop performance tests). Periodic testing and mainte nance ensure continued reliability.

During shutdown periods, an el ectric immersion heater is provided for standby heating. The engine can thus be kept in constant readiness for an immediate start. The heating unit is mounted at the bottom of the accessory rack to heat the engine cooling water which circulates by thermosyphon action to the lube oil cooler, engine, and turbocharger after coolers.

A thermostat sensing water temperature controls the heating elements. The water temperature is controlled to maintain a lube oil temperature of approximately 120 F as described in Section 9.5.7. The auxiliary motor-driven oil pump circulates lube oil through the lube oil cooler to pick up heat during st andby conditions and then returns the warmed oil to the engine sump (see Section 9.5.7). Low oil temperature alarm is prov ided to ensure that the immersion heater is operating prop erly (see Sections 9.5.7.2 and 8.3.1.1.7.2.3

).

To ensure that all components and piping are in itially filled with water, a demineralized water supply is temporarily connected to the 1.25-in. fill-drain connection located on the engine base at the cooling water pum p end. Filling the cooling water sy stem from the bottom up allows entrapped air to be vented to the expansi on tank. (The HPCS engine has a permanent connection.)

The engine cooling water return pipe (between engine block a nd temperature regulating valve) is slightly higher than the top of the expans ion tank. However, as may be seen from Figures 9.5-1 and 9.5-3, during system operation any entrappe d air will be prope rly relieved to the expansion tank through the provided vent lines due to the differential pressures involved.

A 500-gal reservoir tank is provided in the c ooling water system of each diesel engine associated with diesel generators 1 and 2 to permit operation of the engine for the time required to receive SW cooling. The HPCS diesel engine is desi gned to permit operation without cooling for a time equivalent to that required to bring the cooling equipment into service with energy from the HPCS diesel generator.

In accordance with the manufacturer's maintenance instructi ons, a corrosion inhibitor is added to the demineralized fill water to preclude corrosion and organic fouling in the diesel engine cooling water system. Since the entire system is enclosed in th e diesel genera tor building and maintained in a warm condition with immersion heaters, antifreeze compounds are not needed.

Cooling system materials of c onstruction include cast irons, carbon steel, rubber, and bronze. Corrosion inhibitors can be used effectively with these materials.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 9.5-19 Demineralized water and a corrosion inhibitor are in conformance with the engine manufacturer's recommendations.

9.5.5.3 Safety Evaluation

The diesel generator cooling water system meets the single failure criterion in that if a failure in the system prevents the operation of its associated diesel generator, the remaining diesel generators will not be affected.

This redundancy is accomplishe d by segregating the power s upplies to engineered safety features into three mutually exclusive divisions, each provided with a diesel generator and associated cooling water system. The cooling water systems associated with a particular diesel generator are cooled from an independent SW system.

In the event of the loss of offsite power, the SW pumps wh ich supply cooling wa ter to the heat exchangers begin operation within a safe margin of the point at which the diesel generators 1 or 2 would require the cooling capability of the heat exchangers. See Section 9.2.7 for evaluation of this system.

The high temperature shutdown switches are locked out of th e safety circuit during the automatic (emergency) operational m ode of the diesel generators to ensure the availability of the emergency power from each generator.

Evaluation of the diesel gene rator operation under light load conditions is presented in Sections 8.3.1.1.7.1.11 and 8.3.1.1.7.2.11.

9.5.5.4 Testing and In spection Requirements

The system is operated and tested initially with regard to flow path, flow capacity, and mechanical operability (see Section 14.2).

To ensure continued integrity of the diesel generator cooling water system, scheduled testing of equipment is performed as part of the overall engine performance checks at regular intervals in accordance with the Technical Specifications. Periodic inspections are conducted no less than once every 30 months in accordan ce with plant procedures prep ared in conjunction with the manufacturer's recommendations.

Instrumentation is provided to monitor cooling water temperature and pr essure, expansion tank level, and to alarm high water jacket temperature. These instrume nts receive periodic calibration and inspection to verify their accuracy.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 9.5-20 The water in the cooling water system is periodically analyzed and treated, as necessary, to maintain the desired quality.

9.5.6 STARTING

AIR SYSTEM

9.5.6.1 Design Bases

a. Each emergency diesel generator, including the HPCS diesel generator, is provided with separa te, independent starting air systems;

b. Each starting air system on diesel genera tor 1 or 2 has suffici ent air receivers to provide for five diesel generator starts. The starting air system on each diesel generator consists of two completely redundant systems including two banks of

air receivers, separate pipi ng and valves, and one pair of air motors per engine which are both actuated on a start signal. The starting system on HPCS diesel generator 3 consists of two separate systems from separate air receivers through separate piping and c ontrol valves to a pair of air motors on each side of the engine. The air receivers have sufficient air capacity for three starts; and

c. The starting air piping off the engine skid is designed, fabricated, inspected, and erected in accordance with ANSI B31.1. The piping on the engine skid is

constructed using ANSI B31.1 as a guide. The sy stem is designed to Seismic Category I requirements. The air receivers associated with diesel generators 1 or 2 are designed and constructed in accordance with the requirements of ASME Section VIII (1973 Edition). The HPCS di esel generator ai r receivers were designed and constructed in accordance with the 1971 Edition of the same code.

9.5.6.2 System Description

The starting air system is shown schematically in Figure 9.5-1. The starting air systems for diesel generators 1 a nd 2 consist of two electric-motor-driven air compressors, eight air receivers, and associated piping and controls.

Control switches for the electr ic-motor-driven operation of th e air compressors on diesel generators 1 and 2 are on the local diesel engine control board. Thes e control switches permit on-auto-off operation. A selector switch permits selection of either co mpressor function as the primary pressurization compressor.

Pressure switches in either air receiver bank automatically start the selected compressor when the receiver pressure decays to 241 psig. If the selected compre ssor fails to operate or cannot hold system pressure, a separate low pressure alarm switch is provided for each bank of air receivers and is set to alarm at 238 psig on a local panel and in the main control room. When the receiver pressure decays to a lower pressure, the back-up air compressor starts.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-04-021 9.5-21 The HPCS starting air system has two separate air supply trains: one supplied by a diesel-driven compressor and the other by an electric-motor-driven compressor.

The compressor discharge piping is cross connect ed. Both air receivers charge if either compressor operates. A check valve on each receiver inlet isolates one train from the other.

The compressors are controlled automatically by pressure switches on their associated air receiver. The compressor's low pressure setpoint ensures that the compressor starts to maintain the air receiver pressure at a sufficient amount to star t the engine the required number of times.

The HPCS diesel-driven compressor may shut down prior to cl earing the low pressure alarm on the receiver in the electric-mo tor-driven compressor train.

The air receivers are equipped w ith safety/relief valves set at the receiver design pressure.

The major system components are located ad jacent to the diesel generator skid.

For each diesel generator (1 and 2), two separate air-cooled compressors discharge through common piping to two banks of four 32 ft 3 air receivers which are connected in parallel. Each bank of air receivers has the capability of a mi nimum of five engine starts. Each bank is connected through separate piping to a pair of air start motors on each engine.

The flow path is from the air receiver manifold, through an isolation valve, a pressure reducing valve, through piping to the engine, then a strainer, an air relay valve, and a lubricator to each pair of air starter motors. The starting air system on each engine consists of four air start motors. These air start motors drive a flywheel ring gear which turns the engine.

When a start signal is given, an air start solenoi d valve in each redundant system admits air to engage a pair of the air start motor pinions on each engine to the flywheel ring gear. When the pinions are engaged, air is admitt ed through an air control valve to a pair of air start motors. The other pair of start motors on the engine ar e simultaneously engage d by the redundant start solenoid valves. Engine cranki ng time is approximately 2 sec.

The starting air system is designed to provide a reliable method for automatically starting each diesel generator unit. The syst em design is such as to preclude fouling of its components.

Downstream of the air compressors , dryers are installe d to ensure the dewpoint of the air will be below the minimum room temperature if the system is filled during the worst ambient air conditions. The air dryers are drained periodically to remove oil and moisture. A filter is provided downstream of the dryer to ensure no desiccant or debris can enter the system. In

addition, "Y" type strainers are provided upstream of the starting air valves and motors.

These strainers are cleaned periodically to ensure that they are kept free from contaminants.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 9.5-22 The air starting system of the HPCS diesel generator is similar to that of diesel generators 1 and 2 described above except (a) the number of air receivers:

two, of approximately 36 ft 3 each, (b) the number of engine starts: three from the air receivers with the compressors locked out, (c) the redundant starting air compressor is diesel driven and automatically operates without manual actions, and (d) there are two pa rallel systems that each simultaneously operate a pair of air motors on the diesel generator for a start. The pa rallel systems prov ide the ability to produce a start even if one system sustains a pressure boundary failure or a failure of one starting air solenoi d valve to open.

See Section 8.3.1.1.7.2.5 for additional discussion of the HPCS diesel engine starting air system.

9.5.6.3 Safety Evaluation

Each diesel generator starting air system is capable of suppl ying a sufficient quantity of air from its associated air receivers to ensure a successful starting op eration of the diesel generator independent of normal plant power sources.

The starting air systems for each diesel generator unit are physically and electrically separated to ensure that no single failure can cause malfunction of both divisions of standby ac power.

The single failure criterion is satisfied and si gnificantly enhanced by ha ving redundant piping systems and mechanical equipment for diesel ge nerators 1A and 1B a nd duplicate (but not functionally redundant) pi ping systems and mechanical co mponents for the HPCS diesel generator. All three diesel generators also have redundant starting solenoid valves.

9.5.6.4 Testing and In spection Requirements

Preoperational testing was perfor med as described in Section 14.2. To ensure continued integrity of the diesel generator starting ai r system, scheduled testing of equipment is performed as part of th e overall engine performa nce checks at regular intervals in accordance with the Technical Specifications. Periodic insp ections are conducted no less than once every 30 months in accordance with plant procedures prepared in conjunction with the manufacturer's recommendations.

9.5.7 DIESEL

GENERATOR LUBRICATION SYSTEM 9.5.7.1 Design Bases

a. The diesel generator lubrication syst em is designed to provide sufficient lubrication for proper operation of its associated diesel generator under all loading conditions. The system is required to circulate th e lube oil to the diesel engine working surfaces and to remove excess heat generate d by friction during operation. The system provides oil at the engine surfaces at approximately C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 9.5-23 120°F during the anticipated long periods of standby duty by use of an electric immersion heater in the cooling water system; and

b. The piping system was constructed to the guidelines of ANSI B31.1 and Seismic Category I requirements.

9.5.7.2 System Description

The lubrication system for each diesel generator is shown in Figure 9.5-1. The lubrication system for each diesel engine is mount ed on its diesel generator skid (see Figure 9.5-4) and is a combination of three separate systems. These are the main lubricating oil systems, the piston cooling system, and the scavenging oil system. Each system has its own pump. The main lube oil pump and the piston cooling oil pump are in tande m housings and share a common drive shaft. All three pumps ar e positive displacement, helical gear type, mounted externally at the front of the engine, and are gear-driven from the engine.

The main lubricating pump supplies oil under pr essure to the various moving parts of the engine. The piston cooling pump supplies oil for the cooling of the pistons and lubrication of the piston pin bearing surfaces. After circulation through the engine part s, the lubricating oil flows back to the engine oil sump. The scavenging oil pump take s suction from the engine oil sump and pumps this oil through a filter and lube oil cooler to the strain er sump which supplies the main and piston lubricating pumps. The lube oil cooler is a shell and tube, water-cooled type capable of adequately coo ling the engine lube oil when operating at any load point within the engine generation load range. The diesel cooling water acts as the lube oil cooler heat sink.

The engine lubrication system including the lube oil cooler is fu rnished by the engine manufacturer, Electromotive Diesel, a division of General Motors. The diesel cooling water system, also furnished with the e ngine, removes the heat from the lube oil coolers. Heat from the diesel cooling water system is removed in the diesel generator cooling water heat exchangers. The characteristics of these heat exchangers are described in Section 9.5.5.

The lubrication system on each engine also has three small lube oil pumps to circulate oil through the engine main bearings and the turbocharger beari ngs to minimize wear when the engine starts.

The circulating lube oil pump con tinuously circulates lube oil thr ough the main lube oil system filter, the lube oil cooler (w here it is warmed to approximately 120°F) and then through the engine main bearings. This maintains the oil level in the engine just below the camshaft so the oil pressure will increase rapidly when the engine starts from a normal standby condition or on a hot restart.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 9.5-24 An ac-driven soak back pump continuously circulates lube oil through the turbocharger bearings to minimize wear when the engine star ts and to remove heat from the turbocharger following engine shutdown. A dc-driven soak back pump is a rranged in parallel with the ac soak back pump and will automatically start on loss of pressure downstr eam of the ac and dc pumps.

Instrumentation is provided that alarms locally and activates a "DG Trouble Alarm" in the control room on loss of pressure downstream of any of the pumps.

Abnormal lube oil pressures, temperatures, low sump level, and loss of pressure on the lube oil circulating pump disc harge are annunciated.

In the event of a high crankcas e pressure, annunciator and com puter alarms are provided to alert the operator. A manual shutdown will then be made for diesel gene rators 1 and 2. For

diesel generator 3 (HPCS) an automatic shutdown will occur unless an auto initiation signal is present. The hand-hole or top d eck covers, following a high crankcase pressure condition, will not be opened until the engine has been allowed to cool off. This will prevent ignition of oil vapors due to air admittance.

Suitable screens and/or filters in the engine lubrication oil fill pipes prevent entry of foreign material into the engine crankcase. Procedural control, operator traini ng, and careful labeling of fill ports to identify the sta ndard and grade of lubricant to be used ensures that the proper lubricant is used. Lubricant storage containers are similarly labeled to identify contents.

Sampling and testing of the lubricating oil to verify conformance to ASTM Standard 0975-74, Grade 2-D, is performed periodically.

The manufacturers recommendations on measures to be taken to maintain the required quality of the lubricating oil provided in the engine instruction manual are followed.

The following sensors and alarms are provided as described below (diese l generator 3 values are in parentheses):

Sensor Alarm Point Purpose Engine lube oil temperature low 115°F (100°F) To warn of loss of warm-up immersion heater Engine lube oil temperature high 240°F (230°F) To warn of loss of cooling

water system

Engine lube oil pressure low (alarm) 26 psig (25 psig)

To warn of low oil pressure

prior to engine damage

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-07-023 9.5-25 Sensor Alarm Point Purpose Engine lube oil pressure low (shut down) 21 psig (23 psig)

Shut down engine only during test mode to prevent

damage Shutdown engine lube oil

pressure low (alarm) 7 psig (7 psig) Warns of inadequate lube oil

pressure in standby mode.

Operator actions to be taken to resolve alarm conditions are provided in appropriate plant procedures.

See Sections 8.3.1.1.7.1.4 and 8.3.1.1.7.2.4 for additional discussion of the diesel generator lubrication system.

9.5.7.3 Safety Evaluation

The diesel generator lubrication system is capable of providing suffici ent lubrication under all loading conditions. Each engine oil sump is of adequate size to cont ain a supply of oil to support 7 days of c ontinuous operation.

The provision for a physically separate lubrication system for each diesel engine satisfies the requirements of the single failure criterion for complete inde pendence and redundance of the onsite power system by avoiding any commonality between diesel generato r units. All system equipment is housed inside a Seismic Category I structure containing no high- or moderate-energy piping.

The lubrication system has a low le vel alarm in the sump to warn of low oil level. The engines are periodically visually inspected for oil l eaks to guard against excessive oil leakage.

9.5.7.4 Testing and In spection Requirements

Preoperational testing was perfor med as described in Section 14.2. To ensure continued integrity of the diesel generator lubrication system, scheduled testing of equipment and lubrication oil quality is performed as part of the overall engine performance checks at regular intervals in accordance with the Technical Specifications. Periodic inspections are conducted no less than once every 30 months in accordance with pl ant procedures prepared in conjunction with the manufacturer's recommendations.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 9.5-26 9.5.8 DIESEL GENERATOR COMBUSTION AIR INTAKE AND EXHAUST SYSTEM

9.5.8.1 Design Bases

a. The diesel generator combustion air intake and exhaust syst em is designed to supply clean combustion air to each dies el engine and to exhaust combustion gases in a manner that will not effect the operational function of the diesel engines; b. Each diesel engine is provided with an independent combustion air intake train which filters and directs air from the ex terior of the diesel generator building (south side) to the engine turbocharger, and an independent exhaust train which silences and directs engine exhaust gases to the exterior of the diesel generator building (north side). Since each diesel generator unit has its own independent and separate intake and e xhaust train, the single failu re criterion is satisfied;

c. The air intake trains are designed to eliminate contaminating substances, such as dust and larger foreign object s, by filtering the air suppl y through a screened air intake louver, a prefilter, and then through an oil bath air cleaner for the HPCS engine and a cartridge type air cleaner for the engines associated with diesel generators 1 and 2; and

d. The combustion air intake and exhaust system is protected from externally generated missiles (i.e., tornado missiles) by enclosure in a Seismic Category I structure. The piping for the diesel engine intake and exhaust systems is Seismic Category I and is in accordance w ith the requirements of ANSI B31.1. The exhaust silencers, however, are not ANSI B31.1 material but are ASTM A569 which is suitable for the se rvice. The seismic and quality group classification of components in this system is provided in Section

3.2. Nondestructive

examination in accordance with ASME Section III, ND-5000 requirements, was performed on the welds in the combustion air intake and exhaust systems piping.

9.5.8.2 System Description

The combustion air intake and exhaust system is shown in Figure 9.5-1 and the location within the diesel generator building is shown in Figure 9.5-2.

The air intake trains associated with each diesel generator ar e housed in separate rooms and each is supplied air from the exterior of the diesel generator building (south side) through a screened air intake louver. Each engine air intake system consis ts of prefilters, an oil bath type air cleaner for the HPCS engine and cartridge type for the engines associated with diesel C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-09-038,10-017 9.5-27 generators 1 and 2, air turning box, the necessary piping, ductw ork, and flexib le connections to the inlet of the engine turboc harger and aftercooler. An in-line air intake silencer is also provided in the HPCS diesel engine air intake system.

Intake air is cooled in the aftercooler by the engine cooling water syst em to improve engine operational efficiency. The air intake capacity of each diesel engine associated with diesel generators 1 or 2, at 14.7 psi and ambient temperature, is 10,120 cfm. The air intake flow to the HPCS diesel engine at 14.7 psi and ambient te mperature is 10,200 cfm.

The exhaust trains associated with each diesel generator are also housed in separate rooms.

Each engine exhaust system consists of an e xhaust manifold, turbocharger, exhaust silencer, and the necessary piping and ductwork. Exhaust piping from the diesel-driven air compressor in the HPCS starting air system is connected to the correspondi ng diesel exhaust line upstream of the exhaust silencer.

Exhaust gases are discharged th rough the turbocharger from the exhaust manifold and are expelled through ductwork and an exhaust silencer to the exteri or of the diesel generator building (north side). The exhaust gas flow from each diesel engine associated with diesel generators 1 or 2 is 23,000 cfm at 770°F and from the HPCS di esel engine is 23,000 cfm at 735°F.

9.5.8.3 Safety Evaluation

The provision of a physically separated and i ndependent intake and exhaust train for each diesel generator unit satisfies the requirements of the single fa ilure criterion for complete independence between units.

Since each air intake and exhaust train is housed within a Seismi c Category I building, they are protected from externally generate d missiles. No high- or modera te-energy piping is present in the diesel generator building wh ich could present a potential hazard to the operational function of these systems.

The air intake trains are designed so that an ad equate supply of quality air is available to the diesel engines as required. To eliminate foreign objects which could restrict the supply of air to an engine, air is drawn into the system fr om the outside atmosphere through a screened intake louver and prefilter and th en through an oil bath cleaner for the HPCS engine and a cartridge type air cleaner for the engines associated with diesel generators 1 and 2. The postulated design worst-case dust storm is defined in Section 2.3. The postulated design basis ashfall event is defined as a 20 hour2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> event concurrent with a 2 ho ur loss of offsite power. The dust storm concern is negligible by comparison with the design worst-case ashfall event due to the smaller amount of particulate accumulation associated with th e dust storm. The results of an analysis of the oil bath and cartridge filters for the intake ai r system shows the adequacy of either of these filters in handling a severe dust storm event without affecting the diesel

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-017 9.5-28 generators performance. The results of an analysis for the cart ridge filters for the intake air systems shows the ad equacy of these filters in handling an ashfall ev ent without affecting the safe shutdown credited Division 1 and 2 diesel generato rs. Evaluation of a postulated release from the containment inerting nitrogen tank determined that such a release could stall all three diesels for approximately 8 minutes. This postulated release, how ever, would not result in any conditions that would lead to turbine tr ip or reactor scram.

Because the postulated release from the nitrogen tank does not result in a reactor scram or turbine trip, offsite power

remains available during the postulated nitroge n tank release, and reliance on the diesel generators is not required. Th is assumption is consistent with guidance in Standard Review Plan (SRP) Sections 3.6.1 and 3.6.2.

There is no appreciable effect on a diesel generator's ability to carry its required load as a result of barometric pressure drops which would effect the capacity of the combustion air intake train.

Recirculation of combustion products from the diesel exhaust to the air intake which could significantly effect the operation of the diesel engine is precl uded by the degree of horizontal and vertical separati on between the exhaust and air intake (see Figure 9.5-2

).

9.5.8.4 Inspection and Testing Requirements

The combustion air intake and exhaust system was checked for system leaks and blockage following initial installation and testing of the diesel engines (see Section 14.2).

Periodic cartridge replacement/cleaning of air intake filter units is performed at regular intervals and when the filter di fferential pressure observed during engine operation indicates the filter is dirty. Periodic inspection of the adapter and screen assemblies in the exhaust manifold is performed as needed. Routine testing of the diesel engines in accordance with the Technical Specifications verifies the integrity of the air intake and exhaust systems. Periodic inspections are conducted no less than once every 30 months in accordance with plant procedures prepared in conjunction w ith the manufacturer's recommendations.

9.5.9 PLANT

DECONTAMINATION FACILITY

9.5.9.1 Design Bases

The plant decontamination facility is designed to provide a central location for equipment decontamination in a relatively sa fe and efficient envi ronment. Safety of personnel has been carefully considered and built into all aspects of the operation.

Local filtration and increased capability, per recommendations of the Handbook of the American Conference of Governmental Industrial Hygienists, have been incorporated into the existing radwaste building ventilation system to minimize airborne contamin ation. Each of the facility components has been strategically located to provide a safe, efficient, and relativ ely quiet process from entry to C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 9.5-29 disassembly, cleaning, reassembly, and exit. Controls and indication ha ve been positioned in a central location to minimize setup time and thus reduce pers onnel exposure. The decontamination facility has been designed to Quality Cl ass II and Seismic Category II requirements.

9.5.9.2 System Description

The decontamination facility is located in the radwaste building on the 467-ft el. adjacent to the contaminated tool room (see Figure 9.5-5). This facility consists of the equipment necessary to enable decontamination of a vari ety of plant components, tools, and other portable equipment.

To minimize airborne contamination and process fumes in the decontamination room, cleaning tanks are provided with indivi dual exhaust hoods per recomme ndations of the Ventilation Handbook of Governmental Industrial Hygienists. The e xhaust system for each hood is complete with its own high-efficiency particulate air (HEPA) filt er unit and exhaust fan. The air from the exhaust fans is ducted to the radwaste building exhaust system (see Section 9.4.3.2 for system description).

Differential pressure switches are provided across the prefilters and the HEPA filters to annunciate in case of a di rty filter condition. The fan flow is controlled by the fan differential pressure controller which controls a modulating damper at the fan discharge. All the fans are located in the contaminated tool room to minimize noise levels in th e decontamination room during operation.

An automatic sprinkler system has been installe d for protection against the unlikely event of a fire. Curbs have been installe d at all doors to the decontamina tion room, the equipment hatch, and the contaminated tool room to contain fl ooding caused by a major spill or activation of the sprinkler system. A decontamination sink and an emergency eyewash shower station have been installed to permit immediate treatment in case of a splashing accident.

9.5.9.3 Safety Evaluation

The plant decontamination facility has no safety function. Malfuncti on or failure of the decontamination facility will not impair normal or emergency plant operations.

9.5.9.4 Testing and In spection Requirements On installation and periodically thereafter, HEPA filters are given in-place DOP tests in accordance with ANSI N 510-1980, Testing of Nuclear Air Cleaning Systems.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 9.5-30 9.5.10 REFERENCES

9.5-1 H. R. Clay, "Power Generation Control Complex Design Criteria and Safety Evaluation," General Electric NEDO-10466, Revision 1, September 1977.

9.5-2 Electro-Motive Divisi on, General Motors, LaGrange, IL, "Stationary Power Operating Manual," Volume 11, 2nd Edition, June 1971.

9.5-3 SER Licensing Condition No. 9. Lette r from G. D. Bouchey, Supply System, to A. Schwencer, NRC, da ted June 4, 1982 (G02-82507).

C OLUMBIA G ENERATING S TATION Amendment 55 F INAL S AFETY A NALYSIS R EPORT May 2001 9.5-31 Table 9.5-1

Strategic Work Area Communication

Strategic Area Building Location Communication RPS room Radwaste, el. 467 PBX, sound powered Local feed pump control station Turbine, el. 441 PBX, sound powered Hotwell level control station Turbine, el. 441 PBX, sound powered Nonvital 4160 switchgear Turbine, el. 471 PBX, sound powered, two-way radio Vital 4160 switchgear Radwaste, el. 467 PBX, sound powered Vital 4160 switchgear SM-8 Radwaste, el. 467 PBX, sound powered One way from outside to inside DG building corridor DG, el. 441 PBX, sound powered, two-way radio DG building DG, el. 441 PBX, sound powered, two-way radio SW building 1 SW building 1 PBX, sound powered, two-way radio SW building 2 SW building 2 PBX, sound powered

Circulating water pump house CW pump house PBX, s ound powered, two-way radio ECCS equipment Reactor, el. 420 and 441 PBX, sound powered RHR valve room #1 Reactor, el. 471 PBX, sound powered, two-way radio RHR valve room #2 Reactor, el. 471 PBX, sound powered, two-way radio Reactor closed cooling pumps Reactor, el. 549 PBX, sound power ed, two-way radio Hydrogen recombiner Reactor, el. 572 PBX, sound powered, two-way radio Central alarm station CAS PBX, sound powered, two-way radio

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 Table 9.5-2

Locations of Fixed Emergency Lighting LDCN-06-012 9.5-32 BatteryPacks Safe Shutdown Equipment Areas/Access Egress Normal-Emergency AC (DG Backed)

DC (125V Plant Battery

Backed) Life Safety (1.5-hr Batteries) (8-hr Batteries)

RADWASTE BUILDING Access route from the Control Room to RSD Room via stairwell A-7 YES YES EBU Remainder of stairwell A-7 for PFSS event mitigation YES EBU EBU Control Room YES (with Fluorescent Lighting Fixtures) YES EBU EBU RSD Room YES YES EBU ARSD Panel YES EBU SM-8 YES (with Fluorescent Lighting Fixtures)

EBU SM-7 YES Battery Chargers Div. 2 YES ERB RPS 1A YES EBU RPS 1B YES EBU 125 VDC Div. 1 Battery ERB 125 VDC Div. 2 Battery ERB Remainder of RW el. 467 ft for PFSS event mitigation YES EBU EBU RW el. 525 ft PFSS event

mitigation YES EBU, ERB Remainder of RW YES EBU, ERB, High Intensity

Discharge

Light C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 Table9.5-2

Locations of Fixed Emer gency Lighting (Continued)

LDCN-06-012 9.5-33 Battery-Packs Safe Shutdown Equipment Areas/Access Egress Normal-Emergency AC (DG Backed)

DC (125V Plant Battery

Backed) Life Safety (1.5-hr Batteries) (8-hr Batteries)

TURBINE BUILDING Corridor C121 at el. 441 ft Corridor C120 at el. 441 ft (remainder of the access route

from the Control Room to the DG Bldg. For PFSS event

mitigation)

SM-1, 2, 3; SH-5, 6 at el. 471 ft

Remainder of TG YES YES

YES

YES

EBU, High Intensity

Discharge

Light

EBU, ERB, High Intensity

Discharge

Light EBU EBU

DIESEL GEN. BUILDING Corridor D104 at el. 441 ft

Div. 1 DG

Div. 2 DG

HPCS DG, SM-4 YES

YES

YES

YES EBU

EBU

EBU

EBU SSW PUMPHOUSE Pumphouse 1A (Div. 1) Pumphouse 1B (Div. 2 & 3)

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 Table9.5-2

Locations of Fixed Emer gency Lighting (Continued)

LDCN-06-012 9.5-34 Battery-Packs Safe Shutdown Equipment Areas/Access Egress Normal-Emergency AC (DG Backed)

DC (125V Plant Battery

Backed) Life Safety (1.5-hr Batteries) (8-hr Batteries)

REACTOR BUILDING

Access route via stairwell A-6 for PFSS event mitigation

Fuel Pool at el. 606 ft

MCCs at el. 572 ft

SSW Valves at el. 548 ft

RWCU-V-32 at el. 501 ft

YES

YES

YES

YES

YES

EBU

EBU, High

Intensity

Discharge

Light

EBU

EBU

EBU

MCCs at el. 522 ft

YES EBU RHR-P-3 at el. 422 ft

YES EBU LPCS-P-2 at el. 422 ft

YES EBU Access Route from corridor

C402 at el. 501 ft via stairwell S3 to el. 522 ft

YES EBU Remainder of RB

YES EBU TECHNICAL SUPPORT CENTER EBU EBU Fixed Emergency Battery Units (Life Safety Units 1.5-hr, PFSS Units 8-hr) ERB Fixed "Emergency Remote Ballast" with 1.5-hr Battery

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 9.5-35 Table 9.5-3 Diesel Generator Fuel Oil Storage and Transfer System a Number 3 Capacity (gal) Diesel generator 1A or 1B 60,000 HPCS diesel generator 50,000 Type Horizontal - buried Shell material ASME SA-515 Grade 70 Shell thickness (in.) 3/4-15/16 Design temperature 150 Design pressure Atmosphe re plus static head Corrosion allowance (in.) 3/16 Code ASME Section III, Class 3, April 1973 Seismic Category I Diesel Oil Day Tank Number 3 Capacity (gal) 3000

Type Horizontal Shell material ASME SA-285 Grade C Shell thickness (in.) 3/8 Design pressure Atmosphe re plus static head Corrosion allowance (in.) 3/16 Code ASME Section III, Class 3, April 1973 Seismic Category I Diesel Oil Transfer Pumps Number 3 Type Vertical turbine

Rated speed (rpm) 3500 Rated capacity (gpm) 25 Total dynamic head (ft) 51 Code ASME Section III, Class 3, April 1973 Seismic Category I a Capacities are on a per component basis.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 9.5-36 Table 9.5-4 Diesel Generator Heat Exchanger Design and Performance Data Shell Side Tube Side Diesel generators 1A and 1B Fluid circulated Engine water Standby service water Number per engine 1 1 Flow (gpm) 1100 825 Temp in (°F) 190 95 Temp out (°F) 175.4 113.9 Fouling factor 0.0005 0.001 Heat load (Btu/hr) 7,800,000 a Design temperature (°F) 300 300 HPCS generators 1C Fluid circulated Engine water Standby service water Number per engine 1 1 Flow (gpm) 1100 910 Temp in (°F) 187 95 Temp out (°F) 170 118 Fouling factor 0.0005 0.00185 Heat load (Btu/hr) 8,872,000 a Design temperature (°F) 250 200 a This heat rejection value is ba sed on 110% of the c ontinuous rating of the diesel generator.

Amendment 60December 2009 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 42 M512-1 09.5-1.1 Diesel Oil and Miscellaneous SystemsRev.FigureDraw. No.

Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 9.5-1.2 34 M512-2 Diesel Oil and Miscellaneous SystemsRev.FigureDraw. No.Amendment 61December 2011 Form No. 960690ai Columbia Generating StationFinal Safety Analysis ReportRev.FigureDraw. No.9.5-1.3 36 M512-3 Diesel Oil and Miscellaneous Systems Amendment 61December 2011 Amendment 60December 2009 Form No. 960690ai Columbia Generating StationFinal Safety Analysis Report 09.5-1.4 10 M512-4 Diesel Oil and Miscellaneous SystemsRev.FigureDraw. No.

Figure Not Available For Public Viewing 30psi Connection Connection Supporting Systems - Lubricating Oil System andEngine Cooling Water System with ImmersionHeater System - Turbocharged Units 950021.37 9.5-3 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.EngineScav. PumpWaterTankW.T.ETSETA Oil StrainerTankTo Turbo IHS Heat ExchangerRaw WaterTurbo Filter Lube Oil FilterWater Convection Flow - Immersion HeaterWater and Lube Oil - Operating Flow Water Lubricating Oil DC 3 gpm Soak Back Pump Normally Secured AC 3 gpm Soak Back Pump Normally OperatingWater Fill Oil Cooler 75 psi Oil Pan Sump Drain 6 gpm Oil CircPump A.C.W.T.O.T.O.T.1" 1" 1"By-Pass TypeTemperature Control Valve LOTRaw Water Immersion Heater After Coolers Columbia Generating StationFinal Safety Analysis Report Normal Engine Operation Lube Oil System -Turbocharged Engines 950021.38 9.5-4 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Turbo Drive Gears Oil Cooler BypassValve Main Filter Strainer Oil PanTo Sump Load Regulator Camshaft Main Lube Oil ManifoldRocker Arms

& Shaft Piston Cooling Oil Manifold Piston Cooling Oil Pump Strainer Housing Piston Cooling Oil PipeTurbo Filter Hot Oil Detector (Shutdown)

Low Oil Pressure Shutdown DeviceTurbo ImpellerTurbine Planet Gears No. 2 Idler Gear StubshaftTurbo Housing Compressor &

Thrust Bearing Oil Pres.

Gauge No. 1 Idler Gear Stubshaft CheckValve Liner& PistonLow Water & Crankcase

Pressure Detector Scavenging Lube Oil Pump Main Lube Oil Pump Hot Oil Detector (Sensor)Oil Pressure Relief ValveGov.Gov. Drive Gear

StubshaftValve Bridge CrankshaftTo Sump A A Section A-A Columbia Generating StationFinal Safety Analysis Report

Draw. No.

Rev.Figure Schematic Diagram - Standby Lube OilCirculating and Keep Full System 920843.92 9.5-6 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Lube-Oil Filter Lube Oil Cooler EngineTo Circulating Oil Pump Alarm Pressure Switch

Pickup at 20 psi

Dropout at 15 psi 6 GPM Pump Pump "Out" Plug From Scav. Oil PumpTurbo Soak

Back Oil Filter A.C. Soak Back

Pump 3 GPM D.C. Soak Back Pump 3 GPM 1" IPS 75 psi Relief Check 30 psi Relief Check 1" IPS 1" IPS Strainer 1" IPS 1 1/2" IPS Engine Sump Lube Oil Strainer Box Main Bearing Pressure Pump

Outlet Elbow 1/2" OD Steel TubeTo Turbo 1" OD Steel Tube

• • • 1 1/4" IPS 1/2" IPS Swing Check Valve.060" OrificeCamshaft C Weight

Housing1/2" OD Steel Tubes Siphon Break (Connect to Side Outlet of Tee onOil Filter Vent Line

at Engine as Shown)Vent 1/8" Orifice* Indicates Sight Glass (Vertical Height Critical) 5/8" OD Steel Tube on 20-645E4 4" IPS Prime Plug Pump "In" Plug 1" IPSTo Turbo Soak

Back Oil Pump

Alarm Switch

and Switch

and Gauge.

Gauge 0-100 psi

Switch Pickup

at 10 psi Dropout at 6 psi Columbia Generating StationFinal Safety Analysis Report